ML18043A799

Request for Change to Tech Specs of License DPR-20,re Heatup & Cooldown Curves

ML18043A799
Person / Time
Site:	Palisades
Issue date:	07/02/1979
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18043A798	List: ML18043A797 ML18043A799 ML18043A800
References
NUDOCS 7907120337
Download: ML18043A799 (19)

Text

• CONSUMERS POWER COMPANY Docket 50-255 Request for Change to the Technical Specifications License DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 for the Palisades Plant be changed as described in Section I below:

I. Change( s)

A. Change the first sentence of Section 3.l.2e(l) to read:

(1) US Nuclear Regulatory Commission Regulatory Guide 1.99 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data.

B. Change Section 3.l.2.e(2) to read:

(2) Before. the end of the integrated power period for which Figures 3-1, 3-2 and 3-3 apply, the limit lines on the'figures shall be updated for a new integrated power period. The total integrated reactor thermal power from start-up to the end of the new power period shall be converted to an equivalent inte-grated fast neutron exposure (E ~ 1 MeV). Such a conversion shall be made consistent with the dosimetry evaluation of the initial surveillance program capsule which was removed at the beginning of the Cycle 3. For

. - purposes of determining . .....fluence 18 ...

.

at the reactor vessel beltline until a fluence of 8.0 x 10 nvt -

is realized at the inner vessel wall at the beltline region, the

--f~ll~wing basis. ~-~s~-~~l~shed: . -3~36~1019 n~ ~-a;~~a~e~* -~;*' ......

the reactor vessel beltline for 2540 MWt for 40 years at an 80%

load factor. This conversion has resulted in a correlation of 12 1.132 x 10 nvt per 1 MWdt.

C. Change the second paragraph of the Bases for Section 3.1.2.to read:

'I'he reactor vessel plate and material opposite the core has been pur-chased to a specified Charpy V-Notch test result of 30 ft-lb or greater

• at an NDTT of +10°F or less. The testing of baseline specimens 7 901120'337 JO

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2

• associated with the reactor surveillance program indicates that the vessel plate has the hightest RTNDT of plate, weld and HA.Z specimens and that the plate will continue to be limiting up to and including the fluence to which the Figures 3-1, 3~2 and 3-3 apply.( ) The 4

RTNDT has been determined to be 0°F. ( 3 ) An RTNDT of 0°F is used as an unirradiated value to which irradiation effects are added. In addition, this plate has been ~00% volumetrically inspected by ultra-sonic test using 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 component function and has a maximum NDTT of +40°F. ( 5 )

D. Change the fourth paragraph of the Bases for Section 3.1.2 to read:

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 sample exposure by application of the calculated azimuthal neutron flux variation. The maximum integrated fast neutron (E > 1 MeV) exposure of the reactor vessel is computed to be 3.36 x lo19 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 appro-6 priate US NRC Regulatory Guide. ( ) The actual shift in RTNDT will.

be established periodi'cally during plant operation by testing of re-actor vessel material samples which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4-11 of the FSAR. 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.

E. Change the ninth paragraph of the Bases for Section 3.1.2 to read:

The revised pressure-temperature limits are applicable to reactor 18

• vessel inner wall fluences of up to 8.0 x lo nvt or approximately

3

• 7.1 x 10 6

MWd of thermal reactor power. The application of appro-priate fluence attenuation factors at the 1/4 and 3/4 thickness 18 locations results influences of 4.48 x lo nvt and-1.12 x lo 18 nvt, respectively. From Reference 6, these fluences are extrapolated to RTNDT shifts of 69°F and 34°F, respectively, for the limiting base metal transverse oriented material. The criticality condition which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 201°F.

The most limiting wall location is at 1/4 thickness. The minimum criticality temperature 201°F is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2100 psig operation pressure.

F. Change the References for Section 3.1.2 to the following:

References (1) FSAR, Section 4.2.2 (2) ASME Boiler and Pressure Vessel Code,Section III, N-415

• ( 3)

( 4)

Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program:

Properties," August 25, 1977 Unirradiated Mechanical

\

Battelle Columbus Laboratories Report, "Palisades Nuclear Plant*

Reactor Vessel Surveillance Program: Capsule A-240,"

March 13, 1979 (5) FSAR, Section 4.2.4 (6) US Nuclear Regulatory Commission, Regulatory Guide 1.99, "Ef:(ects of-Residual Elements on Predicted-Radiation. Damage.

to Reactor Vessel Materia.J.s," 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

. 4

• G. Change Section 3.1.3.b to read:

b) In no case shall the reactor be made critical if the primary coolant temperature is below 201°F.

H. Change the third paragraph of the Bases for Section 3.1. 3 to read:

During physics tests, special operating precautions will be taken.

In addition, .the strong-negative Doppler coefficient( 3 ) and the small integrated ~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 201°F provides increased assurance that the proper relationship between primary coolant pressure and temperature will be maintained relative to the NDTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the primary coolant pumps .

. I. Replace present Figures 3-1, 3-2 and 3-3 with the new figures 3-1, 3-2 and 3-3, respectively .

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• II. Discussion The enclosed Battelle report entitled, "Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule A-240," March 13, 1979, proyides justification for this Technical Specifications change. This change is being submitted as a result of the latest information obtained from our Reactor Vessel Surveillance Program.

III. Conclusion(s)

Based on the foregoing, both the Palisades Plant Review Committee and the Safety and Audit Review Board have reviewed these changes and find them acceptable.

By

• Executive Sworn and subscribed to before me this 2nd day of July 1979.

e/Y~/~~

Dorothy H Bartkus (Signed) (SEAL) -

Dorothy H Bartkus, Notary Public Jackson County, Michigan My commission expires March 26, 1983 .

• 3.1 3.1.2 PRIMARY COOLANT SYSTEM (Cont'd)

Heatup and CooldoW'Il 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 pressure and temperature for any heatup rate shall be below and to the right of the limit li~es as sho*..m

• on Figure 3-1. 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 limit lines as shown on Figure 3-2. The average cooldown rate shall not exceed l00°F/h in any one-hour time period.

c) Allowable ccmbinations of pressure and temperature for inservice testing from heatup are as shown in Figure 3-3. Those curves in-clude allowances for the temperature change rates 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 pressurizer shall not exceed 200°F/h in any one-hour time period.

e) Before the radiation exposure of the reactor vessel exceeds the ex-posure for which the figures apply, Figures 3-1, 3-2 and 3-3 shall be updated in accordance with the following criteria and procedui:e:

(1) US Nuclear Regulatory Commission Regulatory Guide 1.99 has been used to predict the increase in transition te:r.perature based: on integrated fast neutron flux and surveillance test ..

data.

If measurements on the irradiated specimens show increase above this curve, a new curYe shall be constructed such that it is above and to the left of all applicable data points.

(2) Before the end of the integrated power period for which Figures 3-1, 3-2 and 3-3 apply, the limit lines on the figures shall be updated for a new* integrated power period. The total in-

• tegrated reactor tlle~-::i:=.l power from star-c-u9 to the end of the n~w power period shall be converted to an* equivalent integrated fast neutrcn exposure (:8~1 MeV). Such a conversion shall be made consistent with the dosirr:etry evaluation of the initial 3-4

• 3.1.2 Heatun and Cooldown Rates (Cont'd)

(2) (Cont'd) surveillance program capsule which vas removed at the begin-ning of the Cycle 3. For purposes of determinin<<: fluence a.t

- 18 the reactor vessel beltline until a fluence of 8.0 x lQ _nv:t is realized at the inner vessel vall at the _beltl~ne reg!_c::ip, the following basis is established: 19 3,36 x lo nvt cal-

• culated at the reactor vessel beltline for 2540 MH fqr t

40 years at a 80% load factor. This conversion has resulted 12 in a correlation of 1.132 x 10 nvt per 1 MWdt.

(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 associated with the RT~m~

' 1 -

increase during the period since the curves T.;ere la.st con-structed. The RTNDT increase will be based npon surveillance program testing of the specimens in the initial surveillance capsule .

• Basis All components in the primary coolant system are designed* to* i-Ti thstand the effects of cyclic loads due to primarf system tempe.rature and pres-sure changes. (l) These*cyclic loads are introduced by normal :.mit load transients, reactor trips and start-up and shutdown operation.

During unit start-up and shutdown, the rates of temperature and pres-sure changes are limited. A maximum plant heatup and cooldown rate of l00°F per hour is consistent with the design munber of cycles and sat-*

is fies* stress'. limits far cyclic operation. ( 2 )

Tne re*actor:'.vesseLplate and material opposite the core has been pur-chased to a specified Charpy V-Notch test result of 30 ft-lb or greater 0

at an NDTT of +10 For less. The testing of base line specimens associ-ated with the reactor surveillance program indicates that the vessel plate has the highest RTNDT of plate, weld and HAZ specimens and that the plate will continue to be li."lliting up to and including the fluence 4

to which the Figures 3-1, 3-2-and 3-3 apply. ( ) The RTNDT has been

. d t o .oe 0°F . ( 3 ) An RTi*TDT. or~ _0°F ~s

. t ermine ae - used as an unirr~aiavea

. ' .... -

value to vli.ich irradiation e:'fects are added.. In addition, this plate has _been 100% volumetrically inspected by ultrasonic test using both 3-5

--

• 3.1.2 Heatup 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. (5)

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) 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 flwc measured at the samples and reactor vessel inside radius should be nearly itj.entical, 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 sample exposure by application of the calculated azimuthal neutron flux variation.

maximum integrated fast neutron ( E ) 1 MeV) exuosure of the reactor The vessel is computed to be 3. 36 x lo 19 n~-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 reactor vessel material samples which are irradiated cumulatively by securing them".near _the instde wall of the reactor ve*ssel as described in Section 4.5.3 a,nd Figure 4-11 of the FSAR. To compensate for any increase in the RT caused by irradiation, limits on th_e 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 (Cont'd) stress intensity factor computed( 7 ) is a function of RTNDT' operating temperature, and vessel wall 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 54 psi pressure allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltlin~. 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.

Di.iring 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 thermal gradient stresses are tensile there. Dliring heatup, either the* 1/4 thickness or J/4* thickness location may be limiting depending upon heatup rate.

Figures 3-1 through 3-3 ~efine 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 heatir:g capacity tends t0<- restrict both

. 0 normal heatup and cooldown rates to less than 60 F per hour.

The revised pressure-temperature limits are applicabie to reactor

• • • • • • 8 .-

vessel inner wall fluences of up to 8.0 x 101 nvt or approximately 6

7.1 x 10 MWd 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 4.48 x 10 nvt and 1.12 x 10 nvt, respectively. From Reference 6, these fluences are extrapolated to RTNDT shifts of 69°F and 34°F, respectively, for the limiting base 3-7

• 3.1.2 Heatup and Cooldown Rates (Cont'd)

Basis (Cont'd) metal transverse oriented material. The criticality condition which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics,. considerations) is 201°F.

The most limiting wall location is at 1/4 thickness. The minimum criticality temperature 201°F is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2100 psig operation pressure.

The restriction of heatup and cooldown rates to l00°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.

inservice test curves incorporate allowances for the thermal gradients associated with the heatup curve used to attain inservice test* pres-The sure. These curves differ* from heatup curves only with respect to margin for primary membrane stress.( 7 ) For heatup rates less than 60°F/h, the hypothetical o°F/h (isothermal heatup) at the l/4T location is controlling and heatup curves converge. Cooldovm curves cross for various cooldown rates, thus a composite curve is drawn. Due to the shifts in RTNDT' .NDTT requirements associated with nonreactor vessel. materials a:re, for all practical~. purposes , no longer limiting.

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

References (1) FSAR, Section 4.2.2 (2) ASME 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 SU!"reillance Program: Capsule A-240", March 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 ~.inimum Conditions for Criticality 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

• 0 temperature is below 201 F.

c) When the primary coolant temperature is below the minimum temper-ature specified in "a" above, the reactor shall be subcritical 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 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 pr,essurizer.

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 . Minimum Conditions for Criticality Basis (Cont'd)

(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 primar'J 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 0 .1%A p.

During physics tests, special operating precautions will be taken.

In addition, the strong negative Doppler coefficient ( 3 ) and the small integrated A. 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 201°F provides increased assurance that the proper relationship between primary

-* coolant pressure and temperature will be maintained relative to the l

NDTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the primary coolant pumps.

If the shutdown margin required by Specification 3.10.l is maintained, there is no possibility of a.Ii 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'br the dilution of boron'.so as to preclude*

the possible 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

• -* .... .

l. ,, .

• • *

(Pages 3-14 through 3-16 deleted.)

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