ML19309C285

Amend 55,to License DPR-20,revising Tech Specs Contained in App A,Changing Tech Specs to Conform W/Operating Limits of 10CFR50,App G, Fracture Toughness Requirements

ML19309C285
Person / Time
Site:	Palisades
Issue date:	03/06/1980
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML19309C286	List: ML19309C285 ML19309C289 ML19309C292
References
NUDOCS 8004080405
Download: ML19309C285 (13)

Text

{{#Wiki_filter:* [L]/ pson f( y[g UNITED STATES g NUCLEAR REGULATORY COMMISSION E WASHINGTON, D. C. 20655 \\..... CONSUMERS POWER COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 55 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 July 2,1979, as supplemented November 6,1979, 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 confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is resonable 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 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. / [ ggg40BO

. 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 Appendix A, as revised through Amendment No. 55, 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 COMMISSION f g; 3 wM 3[ [Il tC u... -

i. (t. Dennis L. Ziemann, Chief

/ Operating Reactors Branch #2 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 6,1980 a L

[ ATTACHMENT TO LICENSE AMENDMENT NO. 55 PROVISIONAL OPERATING LICENSE N0. DPR-20 DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the following pages and by inserting the enclosed pages. The revised pages contain the captioned amendment number and marginal lines indicating the area of change. e REMOVE INSERT 4 l 3-4 3-4 3-5 3-5 3-6 3-6 3-7 3-7 i 3-8 3-8 3-9 3-9 (Figure 3-1) i 3-10 3-10 (Figure 3-2) 3-11 3-11 (Figure 3-3) 3-12 3-12 3-13 3-13 3-14 through 3-16

• Next page is 3-17 (Section 3.1.4).

I i I ~. - -

4 3.1 EIMARY COOLANT SYSTE'i (Cont 'd) i 3.1.2 Heatup and Cooldown-Rates The primary coolant pressure and the system heatup and cooldovn rates shall be limited in accordance with Figure 3-1, Figure 3-2 and as fol-I lows: a) Allevable combinations of pressure and temperature for any heatup rate shall be below and to the right of the limit lines as shown s on Figure 3-1. "he average heatup rate shall not exceed 100 F/h h in any one-hour time period, b) Allowable ecmbinations of pressure and temperature for any cool-down rate shall be belov and to the right of the limit lines as 4 lt shown on Figure 3-2. The average cooldevn rate shall not exceed 9 l 100 F/h in any one-heur time period, t c) Allowable c:=binations of pressure and te=perature for inservice 4 testing from heatup are as shown in Figure 3-3 Those curves in-elude allowances for the temperature change rates noted above. 1 Interpolation between li=it lines for other than the noted temper-ature change rates is per=itted in 3.1.22, b or c. d) The average heatup and cooldovn rates for the pressurizer chall not exceed 200 F/h in any one-hour time period. e) Before the radiation exposure of the reactor vessel exceeds the. er.- r posure for which the figures apply, Figures 3-1, 3-2 and 3-3 shall be updated in accordance with the following criteria and procedure: (1) US :Tuclear Regulatory Co= mission Regulatory Guide 1 99 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test i

data, i

i a If measure =ents on the irradiated specimens show increase above this curve, a new curve shall be ccnstructed such that it is above and to the left of all applicable data points. (2) Before the end cf the integr<tted pcVer period for which Figures 3-1, 3-2 cnd 3-3 apply, the limit lines en the figures shall be updated for a new integrated power period. The total in-tegrat,r.rt rc teter t:ternal power fren start-u > to the end af th" ~ new power pericd chtll be convertec to 'tn + ulvtle*'t i nte ' "%. Cast neutren expecure (E 11 !!eV). Such a conversion shall le made consistent wit'h the desi:.ctr'f evaluatitu af *he initial f 3h D D 3 mm -m AmendmentJ,f,55 e J . g. 4 ao =

e t l 3 1.2 Heatup and Cooldcun Rates (Cont'd) l ~(2) (Cont'd) surveillance program capsule which was= removed at the begin-t ning of the Cycle 3 For purposes of deter =inin; fluence at 18 the reactor vessel beltline until a fluence of 6.75 x 10 nyt l is realized at the inner vessel vall at the beltline region, 19 the'following basis is established: 3.6h x 10 nyt cal-e culated at the reactor vessel beltline for 2540 m t i h0 years at a 80", load factor. This conversion has resulted 1 in a correlation of 1.227 x 10 nyt per 1 mwd. g (3) The limit lines in Figures 3-1 through 3-3 shall be moved parallel to the temperature axis in the direction of in-creasing temperature a distance asscelated with the RTg ^ increase during the period since the curves vere last cen-structed. The RT increase vill be based upon surveillance g program testing of the specimens in the initial surveillance capsule. Bacis 4 All components in the primary coolant syste= are designed to withstcnd the effects of cyclic loads due to primary. system temperature and pres-(1) sure changes. These cyclic loads are introduced by nor 21 uni load transients, reactor trips'and start-up an1 shutdown operation. During unit start-up and shutdown, the rates of tenperature and pres-sure changes are limited. A =aximum plant heatup and cooldcun rate of 100 F per hour is consistent with the design nt:=ber of cycles and sat-isfies stres:: li=its for cyclic operation.( 2) i The reactor vessel plate and material opposite the core has been pur - chased to a specified Charpy V-:Tctch test result of 33 ft-lb or greater at an NDTT of +10 F or less. The testing of base line specimens associ-( ated with the reactor curveillance progros indicates that the vessel l weld has the highest RT of plate, veld and HAZ specimens g at the fluence to which the Figures 3-1, 3-2 and 3-3 apply. (h,6)The unirradiated HT o has been determined to be 0 F. g ( I' An RT of 0 F in used as an unirradiated value to which. g irradiation effects are added. In addition, this plate has been 100", volumetrically inspected by ultrasonic test ustrig both Amendment No. j/,[. 55 3_5

= 1 ( [

3. l'. 2 Heatup and Cooldown Rates (Cont'd) f Basis (Cont'd)

' longitudinal and' shear wave methods. The remaining material in the 3 reactor vessel, and other primary coolant system ccmponents, meets the appropriate design code requirements and specific component i f ' function and has a maximum NDTP of + h0 F. I i i As a result of fast neutron irradiation in the region of the core, t f there vill be an increase in the RT vith operation. The techniques 4 l used to predict the integrated fast neutron (E> l MeV) fluxes of the l reactor vessel are described in Section.3'.3.2.6 of the FSAR and also I in Amendment 13, Section II, to the FSAR. Since the neutron spectra and the flux measured at the samples and i reactor vessel inside radius should be nearly identical, the measured transition shift for a sample can be applied to the adjacent section L of the reactor vessel for later stages in plant life equivalent to the 1 difference in calculated flux magnitude. The maximum exposure of the 1 reactor vessel vill be obtained from the measured sample exposure by 7 application of the calculated azimuthal neutron flux variation. "he maximum integrated fast neutron (E ) 1 MeV) exposure of the reactor l Vessel is computed to be 3.64 x 10 nyt for 40 years' operation at [ 19 25b0 m and 80", load factor. The predicted RTg :hift for the g base metal has been predicted based upon surveillance data and the i appropriate US URC Regulatory Guide. The actual shift in RTg vill be established periodically during plant operation by testing of reactor vessel material samples which are irradiated cu=ulatively f l by securing them near the inside vall of the reactor vessel as described i in Section L.5.3 and Figure k-11 of the FSAR. I'o ecmpensate for any increase in the RT caused by irradiation,. limits on the pressure-terf>- erature relationship are periodically changed to stay within the stress limits during heatup and cooldown. Reference 7 provides a procedure for obtaining the allowable leadings t for ferritic pressure-retaining materials in Class I components. This 4 procedure is based ~on the principles of linear elastic fracture mech-I anien. and. involves a stress intensity factor prediction whict is t - lower bound of statie, dynamic and crack arrest critical valuco. The AmendmentNo.'.g.55 3-6 e u y ,p-- 4-,y.i*..,-yas y e- ,as-3, q- .-w++ e g e.,s- .* i--. ng--wr----h-=rnwe==' ' -

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3.1.2 Heatup and Cooldown Rates (Cont'd) Basis (Cont'd) stress intensity factor computed is a function of RTg, operating temperature, and vessel vall temperature gradients. Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference 7. The limit lines of Figures 3-1 through 3-3 consider a 5h psi pressure allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline. In addition, for calculational purposes, 5 F and 30 psi were taken as measurement error allowances for te=perature and pressure, respect-ively. By Reference 7, reactor vessel vall locationa at 1/4 and 3/h thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flav must be arrested. At these locations, fluence attenuation and thermal gradients have been evaluated. During cooldown, the 1/4 thickness location is always more limiting in that the RT is higher than that at the 3/4 thickness location and g themal gradient stresses are tensile there. During heatup, either the 1/h thickness or 3/4-thickness location may be limiting depending upon heatup rate. Figures 3-1 through 3-3 define stress limitations only frem a fracture mechanic's point of view. Other considerations may be more restrictive with re.ipect to pressure-temperature limits. For normal operation, other inherent plant charac-teristics may limit the heatup and cooldown rates which can be achieved. Fump parameters ami pressurizer heating capacity tends to restrict both nor=al heatup and cooldovn rates to less than 60 F per hour. The revised pressure-temperature limits are applicable to reactor 1 vessel inner vall fluences of up to 6.75 x 10 nyt or approximately 6 5 5 x 10 mwd of thermal reactor power. The application of appro-priate fluence attenuation factors at the 1/h and 3/4 thickness loca-10 1 tions'results in fluences of 3.8 x 10 nyt and.95 x 10 avt, respectively. From Reference 6, these fluences are extrapolated to RT shifts of 150 F and 75 F, respectively.. for the limiting veld g / '3-7 Anendment No. J, f4', 55

3.1.2 Heatup and Cooldown Rates (Cont'd) Basis (Cont'd) material. The criticality condition which defines a temperature below which the core cannot be made critical (strictly based upon fracture meenanics' considerations) is 286 F. The most { limiting vall location is at 1/4 thickness. The minimum criti-cality temperature 286 F is the minimum permissible temperature l for the insertice system hydrostatic pressure test. That temperature is calculated based upon 2100 psig operation pressure. The restriction of heatup and cooldown rates to 100 ?/h and the maint-enance of a pressure-te=perature relationship to the right of the heat-up, cooldown and inservice test curves of Figures 3-1, 3-2, and 3-3, respectively, ensures that the requirements of References 6, T, 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 '"h e insertice test curves incorporate allevances for the thermal gradients sssociated with the heatup curve used to attain inservice test pres-sure. These curves differ frcm heatup curves only vith respect te margin for primary membrane stress.( ) For heatup rates less than o0 F/h, the hypothetical 0 F/h (isothermal heatup) at the 1/LT iccation is controlling and heatup curves converge. Cooldown curves cross for various cooldown rates, thus a composite curve is drawn. Due

• .o tne shifts in RT.

,, IIDTT requirements associated vi',h nonreactor A vessel =aterials are, for all practical purposes, no longer limi*.ing. 3.1.2 Hentun and coolclown Pates (cent'd) 3eferences (1) FSiR, Section L.2.2 (2) ASME Boiler and Pressure Vessel Code, Section II!, IT L15 (3) Battelle Columbus laboratories Report, "Falisade.: Pressure 7essel Irradiation Capsule ?rogram: Unirradiated "echanical Properties," August 25, 1977 (L) Battelle Columbus Laboratories Report, "Pa l imules Nuclear Plan t Ifeac t or Vestiel Survelllance Program: Capsule A-240", March 11, 19/9 r.ubm i t t ed t o the NRC by Consumers Power Compant letter daleil .f u l y 1979 3-8 . f, f(, 55 Amendment No.

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3.1.2 Heetup and cooldown Rates (Cont'd) References (Cont'd) ( 5) FSAR, Section L.2.h (6) US Nuclear Regulatory Commission, Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," July, 1975 (T) ASME Boiler and Pressure Vessel Code, Section III, Appendix G, " Protection Against Non-Ductile Failure," 19Th, Edition. (8) US Atomic Energy Cocsission Standard Review Plan, Directorate of Licensing, Section 5.3.2, " Pressure-Temperature Limits." (9) 10 CPR Part 50, Appendix G, " Fracture Toughness Requirements," August 31, 1973. 3.1.3 Minimum conditions for criticality a) Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is belov 525 ?. b) In no case shall the reactor be made critical if the pri=ary ecolant temperature is below 286 F. c) When the primary coolant te=perature is below the minimum temper-ature specified in "a" above, the reactor shall be suberitical by an amount equal to or greater than the potential reactivity in-sertion 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 vater level are esta-blished in the pressurizer. el Primary coolant boron concentration shill not be reduced until af ter a stea= bubble and normal vater level are established in the pressurizer. Baats At the beginning of life of the initial fuel cycle, the moderator tenper-ature coefficient is expected to be slightly negative at operating temper-atures with sll control rods withdrawn. However. the uncertainty of the calculation is cuch that it is possible that a slightly positive coefficient could exist. Anendment No. J7', d, 55 3-12

3.1.3 Minimum conditions for criticality (cont'd) Basis (cont'd) The moderator coefficient at lower temperatures will be less negative or more positive than at operating temperature.( ' It is, there-fore, prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating te=perature ( E525 F). Assuming the most pessimistic rods out moderator coefficient, the maxi =um potential reactivity insertion that could result from depressurir.ing the coolant from 2100 psia to saturation pressure at 525 ? is 0.1" A p. During physics tests, special operating precautions vill be taken. In addition, the strong negative Doppler coefficient (3} and the small integratedA p would limit the magnitude of a power excursien resulting from a reduction of moderator density. '~he require =ent that the reactor is not to be made critical below 286 F provides increased assurance that the proper relationship betseen primar/' coolant pressure and temperature vill be maintained relative to the RT IIDT of the primary coolant system pressure boundary material. Heatup to this temperature vill be accomplished by operating the primary coolant pumps. If the shutdown sargin required by Specification 3.10.1 is maintained, there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure. '!or=al water level is established in the pressurizer prior to the vithdrawal of control rods or the dilution of boren so as to preclude the possible overpressurization of a solid primary ecolant system. References (1) FSAR, Table 3-2 (2) FSAR, Table 3-o (3) FSAR, Table'3-3 Inert naan is Lil) Amendment No. 27. d. cn ,,,}}