ML20005G035

Amends 125 & 129 to Licenses DPR-24 & DPR-27,respectively, Revising Provisions in Tech Specs Re Permissible Heatup & Cooldown Curves

ML20005G035
Person / Time
Site:	Point Beach
Issue date:	01/10/1990
From:	Hannon J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20005G036	List: ML20005G035 ML20005G037
References
NUDOCS 9001180024
Download: ML20005G035 (15)

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%j UNITED STATES

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NUCLEAR REGULATORY COMMISSION a

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1 WISCONSIN ELECTRIC POWER COMPANY-DOCKET NO. 50-266:

' POINT BEACH' NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment. No. 125 -

License No. DPR-24 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

' The application for amendment by Wisconsin Electric Power Company (the licensee) dated: August 3, 1989, as amended October 3, 1989 complies with the standards and requirements of the Atomic Energy-Act of 1954, as amended (the Act), and the Corebsion's rules and s

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.

s 9001180024 900110 PDR ADOCK 05000266 P

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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 Facility Operating License No.

DPR-24. is' hereby amended to' read as follows:

B.. Technical' Specifications I

-The Technical Specifications contained'in Appendices A'and B,

.as revised through Amendment No.125., are hereby incorporated in the license.

The licensee shall operate the facility'in accordance with the Technical Specifications.

3.

This license amendment is effective immediately upon issuance.

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-Technical Specifications'are to-be implemented within 20 days from the s

date'of issuance'.

i FOR THE NUCLEAR REGULATORY COMMISSION i

W John.N. Hannon,. Director i

Project Directorate III-3 l

Division of Reactor Projects - III,-

l IV, V and Special Projects-Office of Nuclear Reactor Regulation Attachment

• I Changes,to the Technical Specifications-Date of Issuance: January 10, 1990 1

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DOCKET NO. 50-301' POINT BEACH NUCLEAR PLANT, UNIT NO. 2 g;

AMENDMENT TO FACILITY OPERATING' LICENSE Amendment No. 129 License No. DPR-27

,7 1.

The Nuclear Regulatory Commission (the Commission) has found that:

d A.

=The application for amendment by Wisconsin Electric Power Company

- i (the. licensee)' dated August 3, 1989, as amended October 3, 1989 4

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; i

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C.

There is reasonable' assurance (1):that the activities authorized by this' amendment-can be conducted without endangering the health

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1 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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Accordingly,-the' license is amended by changes-to theiTechnical-Specifications as' indicated.in the attachment to.this license amendment,-and paragraph 3.B of, Facility Operating License No.

DPR-27 is hereby' amended to' read as follows:

= B.

. Technical Specifications

-The Technical Specifications contained.in Appendices A~and B, as revised through Amendment No.129, are hereby incorporated' A

~in the license.

The-licensee shall operate the facility in accordance with the Technical Specifications.

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'3.

This license amendment' is effective immediately upon: issuance.

The Technical Specifications are to be implemented within 20 days from the date:of issuance.,

FOR THE NUCLEAR REGULATORY COMMISSION John N. Hannon, Director.

Project' Directorate III Division of Reactor Projects - III,

.IV, V and Special Projects Office of Nuclear Reactor Regulation

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Attachment:

Changes to the Technical-Specifications

-Date of Issuance: January 10, 1990 i

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s; Wuf ATTACHMENT TO. LICENSE AMENDMENT NOS.125 AND 129

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TO. FACILITY OPERATING LICENSE NOS.-DPR-24 LAND DPR-27 DOCKET NOS. 50-266.AND 50-301 L.

Revise Appendix A Technical Specifications by removing _ the pages identified below and-inserting the enclosed pages. The revised pages are identified by amendment number and contain marginal-lines indicating the area of change.-

REMOVE-INSERT Figure 15.3.1-1/PBNP Unit No. 1 Figure 15.3.1-1/PBNP Unit Nos. 1&2

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Figure 15.3.1-4/PBNP Unit No. 2 15.3.1-4 15.3.1-4 15.3.1-6 15.3.1-6 15.3.1-7 15.3.1-7 15.3.1-8 ~L 5.3.1-8 Unit 1 15.3.1-8aj 15.3.1-8 Unit 2 15.3.1-17 15.3.1-17 15.3.1 15.3.1-18 15.3.15-1 15.3.15,

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Pressure / Temperature Limits-

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

aC 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.

Thesecondary' side $fthesteamgeneratorwillnotbepressurizedabove 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 maximum'heatup.of 100*F in any one hour and a' maximum cooldown of 200*F.in any one hour, and b.

A maximum spray. water temperature differential between the pres-

.-surizer and spray fluid of not greater than 320 F.

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The reactor vessel material irradiation surveillance specimens shall be s

, removed and examined in accordance with the schedules presented in Tables'15.3.1-1_(Unit 1)and15.3.1-2(Unit 2)todeterminelchangesin l

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. 26,3J,125 Unit 2 Amendment No. 2B,35,129 15.3.1-4

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. produce tensile stresses at.the outer wall of the vessel. These stresses are a

' additive to the pressure induced tensile stresses which are-already present.

The themal induced. stresses at the outer wall of the vessel are tensile c,

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 innerr 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 co;ldown 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.

4 The_heatup and cooldown curves are composite curves which are prepared by determining _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 of -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 (cy) of 19 F was also calculated for this 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 with nuclear operation. The maximum integrated fast neutron exposure NDT Unit 1 - Amendment No.24,125

. Unit 2 - Amendment No.25,129 15.3.1-6

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' Lof. the vessel is computed to be 3.5 x 1019 2

neutrons /cm for 40 years of operation s

at-1518 MWt and 80 Lpercent load factor.52 This maximum fluence is the-exposure L

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- expected at the inner' reactor vessel wall,'which.will be reduced when flux y

reduction measures are implemented. However,-the neutron fluence used to predict the' ART NDT shift is the one-quarter shell thickness neutron exposure. The

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relationship betweenifluence 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."

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'(Reference'6)-

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Once the fluence is determined 'the adjusted reference temperature used in 1

revising the:heatup and cooldown curves is obtained by utilizing the method in j

Section 1.1 of Regulatory Guide l' 99 Revision 2 (Reference 6) for the limiting weld material of both' Unit-1 and Unit 2.

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The h'eatup 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 methods of ASME Code Section i

i 111 (1974 Edition),AppendixG,"ProtectionAgainstNonductileFailure,"and i

.are applicable up to the operational exposure indicated on the figures.

The regulations governing the pressure-temperature limits (10 CFR 50 - Appendix G l

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 conservatisms in-the methods for calculating the pressure temperature limits, is p

not necessary to protect the vessel from damage.

1 Unit 1 - Amendment No. 2A,E3,98.125 Unit 2 - Amendment No. E7,E9.J02,129 15.3.1-7

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The actual temperature-shift of the vessel material will be established periodi-cally during operation by-removing and evaluating reactor vessel material irradia-tion surveillance specimens installed near the inside wall of-the reactor vessel fin 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 f t-.lb level) does not substantiate the predicted shif t, new prediction curves and heatup and cooldown curves must be developed.

LThe. pressure-temperature limit lines shown on Figure 15.3.1-1 for reactor critical-ity and'for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum ternperature requirements' of Appendix G to 10 CFR 50 for

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reactor criticality and for inservice leak and hydrostatic testing.

The spray should not be used if the temperature difference between the pressurizer and spray fluid is greater than 320*F.

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 in Table 15.3.1-1 for Unit 1 and Table 15.3.1-2 for Unit 2.

These schedules have been' developed based upon the requiremerits of the Code of Federal Regulations.

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 dates will provide materials 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 Unit 1 - Amendment No. 24,98,125 15.3.1-8 l

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>The actual temperature shift of'the vessel material will be established periodi-

=cally during operation by removing and evaluating reactor vessel material irradia-tion 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 L 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

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curves and heatup and cooldown curves must be developed.

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

ity 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 temperature difference between the pressurizer and spray fluid is greater than 320'F.

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 0

NDT for the shell.

The reactor vessel materials surveillance capsule removal schedules are presented L

in Table 15.3.1-1 for Unit 1 and Table 15.3.1-2 for Unit 2.

These schedules have L

been developed based upon the requirements of the Code of Federal Regulations.

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

l When the capsule lead factors are considered, the dates accommodate the weld data needs.'of all the participants in the Babcock and Wilcox : laster Integrated Reactor i

Vessel Surveillance Program.

Additionally, the schedule will provide plate-forging material data as well as fluence data corresponding to the expiration of the current Unit-2 license and the end of a twenty-year license extension.

References (1)

FSAR, Section 4.1.5 (2) Westinghouse Electric Corporation, UCAP-10633 (3) Westinghouse Electric Corporation, WCAP-8743 (4) Westinghouse Electric Corporation, WCAP-8733 (5) Babcock 3 Wilcox, BAW-1803 I

(6) Regulatory Guide 1.99, Revision 2 i

Unit 2 - Amendment E7,J02,J28,129 15.3.1.8

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_15.3.15._0verpressure Mitigating System Operations f

r LApplicability-i 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

.Tolspecify functional requirements and-limiting conditions for operation on

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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 systera 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:

a.

Both pressurizer power operated relief valves operable at a setpoint of < 425 psig, 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. M,125 Unit 2 - Amendment No. ;D,129 15.3.15-1

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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 prcent 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 e

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of low level physics tests) to the left of the reactor core criticality curve presented in Figure 15.3.1-1.

The reactor shall be maintained suberitical by at least 1% h until 4.

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 b9ron concentrations in the coolant will be lower and the moderator coefficients will be either less positive or will be negative.

- At all times, the modcrator 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.

Unit 1 - Amendment No. ELS6,125 Unit 2 - Amendment No. 57,90, 129 15.3.1-17

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The limitations of the moderator temperature coefficient'are provided to ensure that the assumptiens used in the accident and transient analyses remain valid j

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 pre-cautions will be +aken.

In addition, the strong negative Doppler coefficientI3}

and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

3 l-The requirement that the reactor is not to be made critical below the Reactor

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Core Criticality Curve provides assurance that a proper relationship between j

reactor coolant pressure and temperature will be maintained during syste, neatup and pressurization.

Heatup to this temperature will be accomplished j

by operating the reactor coolant pumps.

However, as provided in 10 CFR Part 50 j

Appendix G.Section IV. A.3, the reactor core may be taken critical below this curve for the purpose of low level physics tests.

r 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.II)

The requirement or bubble formation in the pressurizer when the reactor has passed the threshold of 1% subcriticality will assure that the Reactor Coolant System will not be solid when criticality is achieved.

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References:

- FSAR Table 3.2.1-1 FSAR Figure 3.2.1-9 FSAR Figure 3.2.1-10 Unit 1 - Amendment No. 26,125 Unit 2 - Amendment No. 90,129 15.3.1-18 3

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