ML20134G254
| ML20134G254 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/21/1985 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20134G257 | List: |
| References | |
| NUDOCS 8508230194 | |
| Download: ML20134G254 (11) | |
Text
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UNITED STATES
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O NUCLEAR REGULATORY COMMISSION
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- CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 89 License No. DPR-20 1.
The Nuclear Regulatory Commission (the. Commission) has found that:
A.
The application for amendment by Consumers Power Company (the licensee) dated June 14, 1985 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,
.the provisions of the Act, and the rules and regulations of the 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 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.
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B. of Provisional Operating License No. DPR-20 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 89, 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY CO P ISSION 1
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s Joh A. Zwolinski, Chief Oper ing Reactors Branch #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: August 21, 1985
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ATTACHMENT TO LICENSE AMENDMENT NO. 89 PROVISIONAL OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3-5 3-5 thru thru 3-12 3-12 e
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a
3.1.2 Heatup and Cooldown Rates (Contd)
(2)
(Contd) surveillance program capsule which was removed at the beginning of the Cycle 3.
For purposes of determining fluence at the reactor vessel beltline until a fluence of 1.3 x 1019nyt is realized at the inner vessel wall at the beltline region, the 19 following basis is established:. S 9 x 10 nyt calculated at the reactor vessel beltline for 2540 W for 40 years at an 80%
load factor. This conversion has resulted in a correlation of 1.989 x 1012nyt per 1 Wd.
g (3) The limit lines in Figures 3-1, 3-2 and 3-3 are based on the requirements of Reference 9, Paragraphs IV.A.2 and IV.A 3.
These lines reflect a preservice hydrostatic test pressure of 2400 psig and a vessel flange material reference temperature of 60*F.
Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature a'nd pres-sure changes.
These cyclic loads are introduced by normal unit load transients, reactor trips and. start-up and shutdown operation. During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A maximum plant heatup and cooldown rate of 100*F per hour is consistent with the design number of cycles and satisfies stress limits for cyclic operation.(
The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-Notch test result of'30 ft-lb or greater at an NDTT of + 10*F or less. The vessel weld has the highest RT f plate, weld and HAZ materials at the fluence to which the ET Figures 3-1, 3-2 and 3-3 apply.(10) The unirradiated RT has been ET determined to be -56*F.(
An RT f -56*F is used as an unirrad-ET isted value to which irradiation effects are added.
In addition, the plate has been 100% volumetrically inspected by ultrasonic test using 3-5 Amendment No. 27, 4, 55, 89
9 4
3.1.2 Heatup and Cooldown Rates (Contd)
Basis (Contd) both-longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate design code requirements and specific
(}
0 component function and has a maximum NDTT of +40 F.
i As a result of fast neutron irradiation in the region of the core, there will be an increase in the RT with operation. The techniques l
used to predict the integrated fast neutron (E > 1 MeV) fluxes of the reactor vessel are described in Section 3.3.2.6 of the FSAR and also in Amendment 13,Section II, to the FSAR.
Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift from a sample can be applied to the adjacent'section of the reactor vessel for later stages in plant The life equivalent to the difference in calculated flux magnitude.
maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron (E > 1 MeV) exposure of the reactor vessel is computed to be 0
5.9 x 10 nyt for 40 years' operation at 2540 MW and 80% load t
factor. The predicted RT shift for the base metal has been NDT predicted based upon surveillance data and the appropriate US NRC Regulatory Guide. ( ) To compensate for any increase in the RT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.
Reference 7 provides a procedure for obtaining the allowable loadings This for ferritic pressure-retaining materials in Class I components.
procedure is based on the principles of linear elastic fracture mechan-f ics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical values. The stress 1
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3-6 Amendment No.11/ M/ II, 89 o~
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l 3.1.2 Heatup and Cooldown Rates (Contd)
Basis (Contd) intensity factor computed ( } is a function of RTg, operating tempera-I ture, and vessel wall temperature gradients.
j Pressure-temperature limit calculational procedures for the reactor-coolant. pressure boundary are defined in Reference 8 based upon Refer-ence 7.
The limit lines of Figures 3-1 through 3-3 consider a 54 psi.
pressure allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline.
In addition, for
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calculational purposes, 5'F and 30 psi were taker as measurement error i
allowances for temperature and pressure, respectively. By Reference 7, i
reactor vessel vall locations at 1/4 and 3/4 thickness are ifmiting.
It is at these locations that the crack propagation associated with the j
hypothetical flaw must be arrested. At these locations, fluence attenu-ation and thermal gradients have been evaluated. During cooldown, the 4
4 1/4 thickness location is always more limiting in that the RT i'
1 NDT l
higher than that at the 3/4 thickness location and thermal gradient j
stresses are tensile there. During heatup, either the 1/4 thickness or 4
3/4 thickness location may be limiting depending upon heatup rate.
Figures 3-1 through 3-3 define stress limitations only from a fracture mechanic's point of view.
Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other inherent plant charac-teristics may limit the heatup and cooldown rates which can be achieved.
Pump parameters and pressurizer heating capacity tends to restrict both normal heacup and cooldown rates to less than 60*F per hour.
The revised pressure-temperature limits are applicable to reactor
~
19 vessel inner wall fluences of up to 1.3 x 10 nyt.. The application 1
of appropriate fluence attenuation factors (Reference 10) at the 1/4 i
and 3/4 thickness locations results in RTND'1, shif ts of 223*F and 170*F, respectively, for the limiting weld material. The criticality condition i
i 3-7 Amendment No. 27, M, 55, 89
3.1.2 Heatup and Cooldown Rates (Contd)
Basis (Contd) which defines a temperature below which the core cannot be made critical i
(strictly based upon fracture mechanics' considerations) is 3'52*F.
The l
most limiting wall location is at 1/4 thickness. The minimum criticality temperature, 352*F is the minimum permissible temperature for the inser-t l
vice system hydrostatic pressure test. That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.
1 The restriction of heatup and cooldown rates to 100*F/h and the maint-enance of a pressure-temperature relationship under the heatup, 4
~
cooldown and inservice test curves of Figures 3-1, 3-2 and 3-3, respec-tively, ensures that the requirements of References 6, 7, 8 and 9 are met.
The core operational limit applies only when the reactor is critical.
The criticality temperature is determined per Reference 8 and the core operational curves adhere to the requirements of Reference 9.
The inservice test curves incorporate allowances for the thermal gradients associated with the heatup curve used to attain inservice test pressure.
These curves differ from heatup curves only,with respect to margin for primary membrane stress.( } For heatup rates less than 60*F/h, the hypothetical O'F/h (isothermal heatup) at the 1/4 T location is con-trolling and heatup curves converge. Cooldown curves cross for various cooldown~ rates, thus a composite curve is drawn. Due to the shifts in NDT, NDTT requirements associated with nonreactor vessel materials RT are, for all practical purposes, no longer limiting.
References-(1) FSAR, Section 4.2.2.
(2) ASME Boiler and Pressure Vessel Code,Section III. A-2000.
(3) Battelle Columbus Laboratories Report, " Palisades Pressure Vessel Irradiation Capsula Program: Unirradiated Mechanical Properties,"
August 25, 1977.
(4) Battelle Columbus Laboratories Report, " Palisades Nuclear Plant Reactor Vessel Surveillance Program: Capsule A-240," March 13, 1979, submitted to the NRC by Consumers Power Company letter dated July 2, 1979.
3-8 Amendment No. 27, 41, 55, 89
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FIGURE 3-3
3.1.2 Heatup and Cooldown Rates (Contd) p References (Contd)
(5)
FSAR, Section 4.2.4.
(6)
US Nuclear Regulatory Commission, Aegulator Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor-Vessel Materials," July, 1975.
(7)
ASME Boiler and Pressure Vessel Code,Section III, Appendix G,
" Protection Against Non-Ductile Failure," 1974 Edition.
(8)
US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, " Pressure-Temperature Limits."
(9) 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements,"
May 31, 1983.
(10) US Nuclear Regulatory Commission, Regulatory Guide 1.99 Draf t Revision 2, April, 1984.
(11) Combustion Engineering Report CEN-189, December,1981.
3.1.3 Minimum Conditions for 'riticality C
a)
Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525'F.
b)
In no case shall the reactor be made critical if the primary coolant temperature is below 352*F.
c)
When the primary coolant temperature is below the minimum tempera-ture specified in "a" above, the reactor shall be suberitical by an amount equal to or greater than the potential reactivity inser-tion due to depressurization.
d)
No more than one control rod at a time shall be exercised or with-drawn until after a steam bubble and normal water level are estab-lished in the pressurizer.
e)
Primary coolant boron concentration shall not be reduced until after a steam bubble and normal water' level are established in the pressurizer.
Basis At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly negative at operating temperatures with all control rods withdrawn.( }
However, the uncer-tainty of the calculation is such that it is possible that a slightly positive coefficient could exist.
Amendment No. J/, AK, )),89 3-12
-