ML20248C670

Proposed Tech Specs Re Heatup & Cooldown Limit Curves

ML20248C670
Person / Time
Site:	Point Beach
Issue date:	08/03/1989
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20248C652	List: ML20248C648 ML20248C670
References
NUDOCS 8908100116
Download: ML20248C670 (16)

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NOTES TO TABLE 1 I

(a) The chemistry values for the shell plates and forgings were derived from vessel material test reports and surveillance capsule chenistry measure-ments.

The chemistry. values for welds were derived from searches in the WOG Materials Data Base, Rev. O and represent rounded, average values.

RT (b) Initial NDT was determined according to the rules of the ASME Boiler and Pressure Vessel Code,Section III, Paragraph NB-2331.

Additionally, where noted, data were treated statistically to obtain the mean value of initial RTNDT and the corresponding standard deviation (oI).

RT (c) The initial NDT values for weld are generic mean values defined by the PTS Rule at 10CFR50.61 (b)(2)ii.

(d) The chemistry data for SA-775 was utilized since this will result in a conservative calculation for this weld.

RT (e) The plate initial NDT value was measured from material-specific drop weight and Charpy data.

Hence, al equals zero (0) as the test methods precisely determined initial RT The margin added to obtain conserva-tive,upperboundvaluesofadjbe.d reference is therefore 20A or 34 F.

(f) The statistical evaluation in BAW-1803[11] Table 3-5 concludes that the mean initial reference temperature for Linde 80 welds is -6 F and the standard deviation (al) about this mean is 19 F.

FromReg. Guide 1.99 Rev. 2[16], the standard deviation for ARTNDT ( A) is 28 F for welds.

Thus, the margin applied equals M = 2V(19)2 + (28)2 67.67 s 68 F

=

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(g) Tha statistical evaluation in BAW-1895[12] concludes that the average initial reference temperature for SA 508 CL.2 forgings is +3 F with a standard deviation of 30 F.

From Reg. Guide 1.99 Rev. 2 [16], the standard deviation for ARTNDT ( A) is 17 F for base metal.

The margin applied is therefore:

Margin = 2J(30)2 + (17)2

• 69 F 68.8

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l I

DOCUMENTATION FOR TABLE 1 1.

Lukens Steel Company Test Certificate No. RM12965-NS, January 3,1966 for Babcock and Wilcox Company.

2.

WCAP-10736, " Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program", December 1984.

3.

Lukens Steel Company Test Certificate No. RM61766-BB, January 20, 1966 for Babcock & Wilcox Company.

4.

Westinghouse Owner's Group (WOG), " Reactor Vessel Materials Data Bas Revision 0, March 1985.

5.

Bethlehem Steel Corporation Test Report No. 911, July 15, 1968 for Babcock

& Wilcox Company.

6.

WCAP-7712, " Wisconsin Michigan Power Co. and the Wisconsin Ele,tric Power Co. Point beach Unit No. 2 Reactor Yessel Radiation Surveil Program", June 1971.

7.

Bethleham Steel Corporation Test Report No. 917, July 18, 1968 for Babcock

& Wilcox Company.

8.

Lukens Steel Company, A-9811 Materials Test Certificate - File No. 602, January 3, 1966.

9.

WCAP-7513, " Wisconsin Michigan Power Co. Point Beach Unit No. 1 Reactor Vessel Radiation Surveillance Program", June 1970.

10.

Lukens Steel Company, C-1423 Materials Test Certificate - File No. 602, June 20, 1966.

11.

BAW-1803, " Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds", January 1984.

12.

BAW-1895, " Pressurized Thermal Shock Evaluation In Accordance With 10CFR50.61 for Babcock and Wilcox Owners Group Reactor Pressurc Ve. als" January 1986.

13.

U.S. NUCLEAR REGULATORY COMMISSION, 10CFR Part 50.61, " Fracture To Requirements for Protection Against Pressurized Thermal Shock Events".

July 23, 1985.

14.

Letter from C. W. Fay to H. R. Denton (NRC), " Response to 10CFR50.61 Protection Against Pressurized Thermal Shock (PTS) Pcint Beach Nuclear Plant, Units 1 and 2", January 20, 1986.

15.

Letter from C. W. Fay tc H. R. Denton (NRC), " Correction to Pressurized Thermal Shock (PTS) Submittal Dated January 20, 1986 Point Beach Nuclear Plant, Units 1 and 2", March 14, 1986.

16.

U.S. NRC Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", May 1988.

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1 TABLE 3 i

SIGNIFICANT PARAMETERS FOR HEATUP AND C00LDOWN LIMIT CURVE REVISION AP.PLICABLE TO JAN0ARY 1, 1995 POINT BEACH NUCLEAR PLANT Method and Assumptions of Heatup/Cooldown Limit Curves 1.

Controlling Welds:

Unit 1 SA-847 Lower Shell Axial Weld Unit 2 SA-1484 Inter.-to-Lower Shell Girth Weld 2.

One set of heatup and cooldown curves will be calculated to be applicable to both units at Point Beach for clarity and simplification.

From Table 1 the chemistry of Unit 2 weld SA-1484 is more limiting than that of Unit 1 weld SA-847.

Additionally, the Unit 1 weld SA-847 is located 15 off the peak fluence (cardinal) axes, and thus, P-T curves calculated for Unit 2 are bounding for Unit 1.

3.

Weld Chemistry:

SA-1484 Cu = 0.26 wt.%

Ni = 0.60 wt.%

4.

Trend and Fluence Attenuation Formulas:

Regulatory Guide 1.99, Rev. 2 5.

WCAP-10638 was used to extrapolate fluence.

L3P cores are assumed to continue through 1995, and no credit for flux reduction measures is taken.

6.

Vessel Chemistry Flygnce Fluence ART Location Factor (x10 n/cm2)

Factor NDT ARL Wall 180 2.05 1.196 215.3 277.3 1/4 T 180 1.39 1.091 196.4 258.4 3/4 T 180 0.64 0.875 157.5 219.5 7.

Heatup/Cooldown Limit Curve Calculational Procedure:

WCAP-8738 8.

Margins for instrumentation uncertainties are not applied in the calculation of the heatup and cooldown curves.. These additional margins are not required by 10CFR50 Appendix G and ASME Section III Appendix G.

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. TABLE 4-COMPARIS0N FOR PTS' RULE PURPOSES 1 L

REFERENCE TEMPERATURES FOR REAC10R VE5SEL BELTIIllE WELD MATERIALS

• l POINT BEACH NUCLEAR PLANT

'RG 1.99 REV. 2 vs. 10CFR50.61 p-

, Unit 1 SA-775/812**

SA'-847**

SA-1101***

Axial Weld Axial Weld Circumferential' Weld:

~ PERIOD Inter.'Shell Lower Shell Inter. to Lower Shell-RG 1.99 RTPTS RG 1.99 RTPTS RG 1.99 RTPTS Present(05/31/89) 225.0 180.8 231.9 215.5 234.0 198.1

License Expiration (10/05/2010)'

256.8 207.1-265.1 249.1 261.9 228.1 w/L3P' Cores Continued License Expiration (10/05/2010) 250.0 200.8 253.2-236.0 251.0 215.0-w/ Expected Flux Reductions +

From Planned L4P and Hf Insert Core. Designs Unit-2 SA-1484***

Circumferential Weld 4

Inter. to Lower Shell l

RG 1.99 RTPTS Present(05/31/,89) 265.4 248.4 l

.L-icense Expiration (03/08/2013) 300.8 292.9 w/L3P Cores Continued License Expiration (03/08/2013) 287.2-273.8 w/ Expected Flux Reductions From Planned L4P and Hf Insert Core g

Designs RT

• Predicted PTS values assume a cumulative lifetime capacity factor of 80%.

I All values are in F.

• • Applicable PTS screening criterion - 270 F.

• • • Applicable PTS screening criterion - 300 F.

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B.

Pressure / Temperature-Limits 1

l Specification:

1.

The Reactor Coolant System temperature and pressure shall be limited in accordance with the limit lines shown in Figure 15.3.1-1 and 15.3.1-2 r

during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100 F in any one hour, b.

A maximum cooldown of 100 F in any one hour, and c.

An average temperature change of $10 F per hour during inservice leak and hydrostatic testing operations.

2.

The secondary side of the steam generator will not be pressurized above 200 psig if the temperature of the steam generator vessel shell is below 70 F.

3.

The pressurizer temperature shall be limited to:

a.

A maximur; heatup of 100 F in any one hour and a maximum cooldown of 200 F in any one nour, and b.

A maximum spray water temperature differential between the pres-surizer and spray fluid of not greater than 320 F.

4.

The reactor vessel material irradiation surveillance specimens shall be removed and examined in accordance with the schedules presented in Tables 15.3.1-1 (Unit 1) and 15.3.1-2 (Unit 2) to determine changes in j

material properties.

The results of these examinations shall be con-sidered in the evaluation of the prediction method to be used to update Figures 15.3.1-1 and 15.3.1-2.

Revised figures shall be pro-l vided to the Commission at least sixty (60) days before the calculated exposure of the applicable reactor vessel exceeds the exposure for which the figures apply.

Unit 1 Amendment No.

Unit 2 Amendment No.

15.3.1-4 l

l o__

L stresses at the outer wall of the vessel.

These stresses are additive to the pressure induced tensile stresses w" ch are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

During cooldown the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stress at the outside wall.

The heatup and cooldown curves are composite curves which are prepared by deter-mining the most conservative case with either the inside or outside wall controlling for any heatup or cooldown rate up to 100 F in any one hour.

In developing these curves, an initial unirradiated RT f -6 F was utilized NDT as reported in BAW-1803 dated January 1984.

(Reference 5) This value is based upon a statistical evaluation of Linde 80 weld material test data consisting of measured reference temperatures, drop weight data, and related pre-irradiated Charpy data.

A standard deviation (o ) of 19 F was also calculeteu for this y

data set.

Both the initial RT and standard deviation values in BAW-1803 may NDT be revised as additional data are obtained.

As a result of fast neutron irradiation, there will be an increase in the RT NDT with nuclear operation.

The maximum integrated fast neutron exposure of the 19 2

vessel is computed to be 3.5 x 10 neutrons /cm for 40 years of operation at Unit 1 Amendment Unit 2 Amendment 15.3.1-6

)..

1518 MWt and 80 percent load' factor.(2) This maximum fluence is the exposure expected at the inner reactor vessel wall, which will be reduced when flux reduction measures are implemented.

However, the neutron fluence used to predict the ART shift is the one quarter shell thickness neutron expcsure.

NDT The relationship between fluence at the vessel ID wall and the fluence at the one quarter and three quarter shell thickness locations is as presented in Regulatory Guide 1.99 Revision 2, " Radiation Damage to Reactor Vessel Materials."

(Reference 6)

Once the fluence is determined, the adjusted reference temperature used in revis-ing the heatup and cooldown curves is obtained by utilizing the method in Section 1.1 of. Regulatory Guide 1.99 Revision 2 (Reference 6) for the limiting weld material of both Unit 1 and Unit 2.

The heatup and cooldown curves presented in Figure 15.3.1-1 and 15.3.1-2 were l

calculated based on the above information and the methodt of ASME Code Section III (1974 Edition), Appendix G, " Protection Against Nonductile Failure", and are applicable up to the operational exposure indicated on the figures.

The regulations governing the pressure-temperature limits (10 CFR 50 - Appendix G and ASME Code Section III - Appendix G) do not require additional margins for instrumentation uncertainties be added to the heatup and cooldown curves.

This is because the inclusion of instrumentation uncertainties, in addition to other conservatism in the methods for calculating the pressure temperature limits, is not necessary to protect the vessel from damage.

I Unit 1 Amendment Unit ~2 Amendment 15.3.1-7

t I

The actual temperature shift of the vessel material will be established periodically during operation by removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and vessel inside radius are identified by a specified lead factor, the measured temperature shift for a sample is an excellent indicator of the effects of power operation on the adjacent section of the reactor vessel.

If the experimental temperature shift (at the 30 ft-lb level) does not substantiate the predicted shift, new prediction curves and heatup and cooldown curves must be developed.

The pressure-temperature limit lines shown on Figure 15.3.1-1 for reactor l

criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.

The spray should not be used if the t emperature difference between the pre.csurizer and spray fluid is greater than 320F.

This limit is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

The temperature requirements for the steam generator correspond with the measured NDT for the shell.

The reactor vessel materials surveillance capsule removal schedules are presented inTable15.3.1-1for!/ nit 1and-Table 15.3.1-2forUnit2.

These schedules have been developed based upon the requ,irements of the Code of Federal Regulations, d

Title 10, Chapter 50, Appendix H and with consideration of ASTM standard E-185-82.

When the capsule lead factors are considered, the scheduled removal Unit 1 Amendment Unit 2 Amendment 15.3.1-8

i dates will provid: caterials data representative of about 10%, 20%, 50%, 90%, and 110%_of the actual _ reactor vessel exposure anticipated during the vessel life.

References

-(1) FSAR, Section 4.1.5 (2) Westinghouse Electric Corporation, WCAP-10638 (3) Westinghouse Electric Corporation, WCAP-8743 (4) Westinghouse Electric Corporation, WCAP-8738 (5) Babcock & Wilcox, BAW 1803 (6) Regulatory Guide 1.99, Revision 2 i

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Unit 1 Amendment l

Unit 2 Amendment 15.3.1-8a u.___..._______

F.

MINIMUM CONDITIONS FOR CRITICALITY Specification:

1.

Except during low power physics tests, the reactor shall not be made critical when the moderator temperature coefficient is more positive than 5 pcm/ F.

2.

Reactor power shall not exceed 70 percent of Rated Power if the moderator temperature coefficient is positive.

3.

In no case shall the reactor be made critical (other than for the purpose of low level physics tests) to the left of the reactor core criticality curve presented in Figure 15.3.1-1.

4.

The reactor shall be maintained subcritical by at least 1 percent k until normal water level is established in the pressurizer.

Basis:

During the early part of the fuel cycle, the moderator temperature coefficient is calculated to be slightly positive at coolant temperatures below 70' percent of rated thermal power.(1)(2) The moderator coefficient at low temperatures will be most positive at the beginning of life of. the fuel cycle, when the boron concentration in the coolant is the greatest.

Later in the life of the fuel cycle, the boron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative.

At all times, the moderator coefficient is negative when > 70 percent of rated thermal power.

Suitable physics measurements of moderator coefficient of reactivity will be made as part of the startup program to verify analytic predictions.

I Unit 1 Amendment Unit 2 Amendment 15.3.1-17 I

The limitations of the moderator temperature coefficient are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle.

This requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest.

During physics tests, special operating precautions will be taken.

In addition, the strong negative Doppler coefficient (3) and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction of mode;ator density.

The requirement that the reactor is not to be made critical below the Reactor Core Criticality Curve provides assurance that a proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization.

Heatup to this temperature will be accomplished by operating the reactor coolant pumps.

However, as provided in 10 CFR Part 50, Appendix G, Section IV.A.3, the reactor core may be taken critical below this curve for the l

purpose of low-level physics tests.

l If the specified shutdown margin is maintained (Section 15.3.10), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.(1)

The requirement for bubble formation in the pressurizer when the reactor has passed the threshold of 1 percent subcriticality will assure that the reactor coolant system will not be solid when criticality is achieved.

1

References:

(1) FSAR Table 3.2.1-1 (2) FSAR Table 3.2.1-9 (3) FSAR Figure 3.2.1-10 Unit 1 Ameadment Unit 2 Amendment 15.3.1-18

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15.3.15 OVERPRESSURE MITIGATING SYSTEM OPERATIONS Applicability l

Applies to operability of the overpressure mitigating system when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test.

Objective To specify functional requirements and limiting conditions for operation on the use of the pressurizer power operated relief valves when used as part of the overpressure mitigating system and to specify further limiting conditions for operation when the reactor coolant system is operated without a pressure absorbing volume in the pressurizer.

Specification A.

System Operability 1.

Except as specified in 15.3.15. A.2 below, the overpressurization mitigating system shall be operable whenever the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressurization temperature for the inservice pressure test, as specified in Figure 15.3.1-1.

Operability requirements are:

l a.

Both pressurizer power operated relief valves operable at a setpoint of 1 25 psig.

4 b.

The upstream isolation valves to both power operated relief valves are open.

2.

The requirements of 15.3.15.A.1 may be modified to allow one of the two power operated relief valves to be inoperable for a period of not more than seven days.

Unit 1 Amendment No.

Unit 2 Amendment No.

15.3.15-1