ML18044A208
| ML18044A208 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/06/1979 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18044A207 | List: |
| References | |
| NUDOCS 7911080352 | |
| Download: ML18044A208 (12) | |
Text
ATTACHMENT .
Page Changes for Proposed Technical Specifications Change 7911 oso .;~52-
'**rV
3 .1 : PRIMARY COOLANT SYSTEM . (Cont 'd) 3.1.2 Heatun and Cooldown Rates The primary coolant pressure and the system heatup and cooldown rates shall be limited in accordance with Figure 3-1, Figure 3-2 and as fol-lows:
a). Allowable combinations of presst.lre and temperature for any heatup
~ate shall be below and to the right of the limit li::ies as shown on Figure*3-l: The average heatup rate shall not exceed l00°F/h in any one-hour time period.
b) Allowable combinations of pressure and temperatt:.re for any cool-down rate shall be below and to the right of the l:i..."'lit lines as shown on Figure '3-2. The a_verage c.ooldmm rate shall not exceed 100°F/h in any one-h0ur time period.
c) AJ,.lowable combinations of pressure and temperature for inservi.:e
- testing from heatup are as shown in ?igure 3-3. Those* ccirves in-clude allowances for the temperature change rate*s noted above.
Interpolation between limit lines for other than the noted temper-ature change rates is permitted in 3.l.2a, b or c.
d) The average heatup and cooldown rates for the pressurize!* shall not exceed 200°F/h in any one-hour time period.
e) Before the radiation exposlire of the reactor Yessel exceeds the ex-posure for which the figures apply, ?igures 3-1, 3-2 and 3-3 stall be updated. in accordance with the following cri te:?:'ia: and procedure:
(1) US Nuclear Regulatory Commission Regulatory Guide 1.99 has been. used to predict the increase in .transition te."::perature based on*. integrated fast neutron fltLx and su=veillance test data. I If measurements on the irradiated specimens show increase above this .curve,* a new curve sh::i.11 be constructed such that it ::.s above and to the left of all applicable data points.
(2) Before the end of the integr~ted po¥er period fer which Fi~ures 3-1, 3-2 and 3-3 apply, the limit lines on the figures shall be updated for a n'2w integrated power period. T'he total in-tegrated reactor the~~~l power from s~ar~-Up to the end of the fast neutron exposure (2 >l MeV). Such s conver3ion shall be made consisten~ ~ith the d~si~etry evaluation of the initial 3-4
. . . - -- . 3 .1. 2- Heatun and Cooldown Rates (Cont'd)
(2) (Cont'd) surveillance program capsule which was removed at the begin-ning of the Cycle 3. ** For purposes of determinin.s flt:.e!'lce at the reactor vessel beltline until a fluence of 6.75 x lo18 nvt is realized at the inner vessel wall at the beltline region, 19 the following basis is established: 3.64 x lo nvt cal-
- culated at. the reactor vessel beltiine for 2540 t-11\ for 40 years at a 80% load factor. This conversion has resulted 12 in a correlation of 1.227 x 10 nvt per 1 MWdt.
(3) The li!llit lines in Fi5~res 3-1 through ~-3 shall be ~oved parallel to the temperature axis in the direction of in-creasing temperature a distance associated with the RT,~~
!'I.Lr.!.'
increase during the pe::-iod since the cur-res *..-ere .i.as~ con-structed. The *RTNDT increase will oe based npon surveillance program testing of the specime!'ls in the initial surveillance capscle; Basis All components in the pri.::iar.r coolant syst.em are designed* to withstand the effects of cyclic loads due to primar.r system temperature and pres-sure changes. (l) These cyclic loads are introduced by normal :.init load transients, reactor trips a.."'l.d start-up and shutdo*..-n operation.
During unit start-up and shutdow-n, the rates of temperature and pres-sure changes are li:nited. A maximum plant heatup and cooldown rate of 100°? per hou= is consistent *..rith the design number of cycles and sat-2 isfies stress limits for cyclic operation.( )
The reactor vessel plate and mater.ial o.pposite the core has been ~ur chased to a specified Charpy V-Ifotch test result of 30 ft-lb or greater at an l'TDTT of +10°? or less. The testing of base line specimens assoc~-
ated with the reactor surveillance progra.11 indicates that the vessel weld has the. highest_ RTNDT of plate, weld and HAZ specimens I
at the fluence to which the Figures 3-1, 3-2 and 3-3 apply.
4 I
(
- 6 )The unirradiated RTNDT has been determined to be o°F. I
( 3
- 8 )A
- used as an unirradiat. ed value to which irradiation effects are added. In addition, this plate has been 100% volumetrically inspected by ultrasonic test using both 3-5
3.1.2 Heatun and Cooldown Rates (Cont'd)
Basis (Cont'd)*
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 component
- function and has a maximum NDTT of+ 4o°F. C5 ) 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 used to prediCt the integrated fast neutron ( E) l MeV) fltL'<:es 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 for a sample can be applied to the adjacent section of the reactor vessel for later stages *in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured s!.mple exposure by application of the calculated azimuthal neutron flux variation. The maximum i*ntegrated fast neutron ( E > 1 MeV) ei<posure of the reactor 19 . .
vessel is computed to be 3.64 x 10 nvt for 40 years' operation at 2540 MWt and 80% load factor. The predicted RTNDT shift for the base metal has been predicted based upon surveillance data and the 6
appropriate US NRC Regulatory Guide.( ) The actual shift in RTNDT will be established periodically during plant operation by testing*
of reactorvessel material samples which are irradiated cumulatively by securing them riear the inside wall of the reactor vessel as described in Section 4,5,3 and Figure 4-11 of the ?SAR. To compensate for any increase in the RT caused by irradiation, limits on the pressure-temp-erature relationship are periodically changed to stay within the stress limits during heatup and cooldown.
Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components. This procedure is based on the principles of linear elastic fracture mech~
anics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical values. The 3-6
3.1.2 Heatup and Cooldown Rates (Cont'd)
- - Basis-- ( corit 'd) stress intensity factor computed(T) is a function of RT~TDT' operating temperature, and vessel wall temperature gradients.
Pressure-temperature limit calculational procedures for the reactor coolant pressure bounda.r'J are defined in Reference 8 based upon Reference 7. The limit lines of Figures 3-1 through 3-3 consider a 54_psipressure allowance to account for. the fact that pressure is measured in the pressurizer rather than at the vessel beltline. In 0
addition, for calculational purposes, 5 F and 30 psi were taken as measurement error allowances for temperature and pressure, respect-ively. By Reference 7, reactor vessel wall locations at 1/4 and 3/4.
thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw 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 RTNDT is higher than that at the 3/4 thickness location and thermaI gradient stresses are tensile there. During heatup, either the 1/4 thickness or. 3/4* thickness location may be limiting depe.nding 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 th_e heatup and cooldown. rat~s which can* be achieved.
Pump parameters and pressurizer heating *capacity tends to-.-restrict ooth 0
normal heatup and cooldown rates to less than 60 F per hour.
\
The revised pressure-temperature limits are applicable to reactor 18 vessel inner wall fluences of up to 6.75 x lo nvt or approximately 5.5 x 10 6 IvIWd of thermal reactor power. The application of appro-priate fluence attenuation factors at the 1/4 and 3/4 thickness loca-
. . 18 18 tions results in fluences of 3.8 x 10 nvt and .95 x 10 nvt, respectively. From Reference 6, these fluences are extrapolated to RT shifts of 150°F and 75°F, respectively, for the limiting weld NDT 3-7
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 mechanics' considerations) is 286°F. The most limiting wall location is at 1/4 thickness. The minimum criti-cality temperature 286°F is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2100 psig operation pressure.
. 0 The restriction of heatup and cooldown rates to 100 F/h and the maint-enance of a pressure-temperature 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, 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 pres-sure. These curves differ from heatup curves only with respect to margin for primarJ membrane stress.( 7 ) For heatup rates less than 6o°F/h, the hypothetic~l o°F/h (isothermal heatup) at the l/4T location is controlling and heatup curves converge. CooldoWn. curves cross for various cooldown rates, thus.a composite curve is drawn. Due to the shifts in RTNDT' NDTT requirements associated wi.th norireactor vessel materials are, for all practical purposes, no longer limiting.
3.1.2 Heatun and Cooldown Rates (Cont'd)
References (1) FSAH, Section 4.2.2 (2) A.SME Boiler and Pressure Vessel Code,Section III, N-415 (3) Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties,"
August 25, 1977.
(4) Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surreillance Program: Capsule A-240", r:Iarch 13, 1979.
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3.1.2 Heatup and Cooldown Rates (Cont'd)
References (Cont'd)
(5) FSAR, Section 4.2.4 (6) US Nuclear Regulatory Commission, Regulatory 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,"
August 31, 1973.
3.1.3 ~.ininrum Conditions for Criticality a) Except during low-power physics test,.the reactor sh8.J.l 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 I temperature is below 286~F. * .
c) .When the primary coolant temperature is below the minimum temper-ature specified in "a". above, the* reactor shall be subcritica:l by an amount equal to or greater than the potential reactivity in-sertion due to depressurization'.-
d) No more than one control rod a.t a time shall be exercised or with-drawn until after a steam bubble. and normal water level are esta-blished 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 temper-ature coefficient is expected to be slightly negative at operating temper-atures with all control rods withdrawn. (l) However, the uncertainty of the calculation is such that it is possible that a slightly positive coefficient could exist.
3-12
3.1.3 Minimt.1I11 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.(l, 2 ) It is, there-fore, prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating temperature
( ~525°F). Assuming the most pessimistic rods out moderator coefficient, the maximum potential reactivity insertion that could result from depressurizing the coolant from 2100 psia to saturation pressure at 525°F is o .1%A p.
During physics tests, special operating precautions will be taken.
In a~dition, the strong negative Doppler coefficient ( 3 ) and the small integratedA p would limit the magnitude of a power excursion resulting from a reduction of moderator density. The requirement that the reactor is not to be made critical below 286°F provides increased assurance that the proper relationship between primary I coolant pressure and temperature will be maintained relative to the RTNDT of the primary coolant system pressure boundary material. Heatup to this
- temperature will be accomplished by operating the primary coolant pl,l.ID.ps.
If the. shutdown margin 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.
Normal water level is established in the pressurizer prior to the withdrawal of control rods'or the dilution of boron so as to preclude the possibl~ overpressurization of a solid primary coolant system.
References (1) FSAR, Table 3-2
( 2) FSAR, Table 3-6
( 3) FSAR, Table 3-3 3-13
"I. ..
(Pages 3-14 through 3-16 deleted;)
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