ML20207S561

Amends 101 & 91 to Licenses DPR-39 & DPR-48,respectively, Extending Applicability of Heatup & Cooldown Curves to 14 EFPY Yrs

ML20207S561
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	03/13/1987
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207S557	List: ML20207S556 ML20207S561 ML20207S563
References
NUDOCS 8703200035
Download: ML20207S561 (19)

Text

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[b UNITED STATES P

1. 'n NUCLEAN REGULATORY COMMISSION y,

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COMMONWEALTH EDISON COMPANY DOCXET NO. 50-295 ZIONNUCLEARPOWERSTATTON,UNITJ

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l AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.101 License No. DPR-39 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated December 5,1986, supplemented February 26, 1987 and February 27, 1987 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; 8.

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 (1) 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-39 is hereby amended to read as follows:

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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised throuch Amendment No.101, 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 NUCLF R REGULATORY COMMISSION Steven A. Var \\g, Directork Lws

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PWR Project Directorate #3 Division of PWR Licensing-A, NRR

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13, 1987

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-304 ZION NUCLEAR POWER STATTON, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 91 License No. DPR-48 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated December 5,1986, supplemented February 26, 1987 and February 27, 1987, 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; R.

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 (111 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 Comission's regulations and all applicable requirerrents have been satisfied.

?.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby amended to read as follows:

-?-

(2) Technical Specifications The Technical Specifications contained in Appendices A and R, as revised through Amendment No.91

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendnent is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION fA.LAN >

\\ 'd, Steven A. Varga, Director y

PWR Project Directorate #3 Division of PWR Licensing-A, NRR

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13, 1987 O

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ATTACPMENT TO LICENSE AMENDMENTS AMENDMENT NO. 101 FACILITY OPERATING LICENSE NO. OPR-39 AMENDPENT NO. 91 FACILITY OPERATING LICENSE NO. DPR-48 DOCKET NOS. 50-295 AND 50-304 Revise Appendix A as follows:

Remove Pages Insert Paaes vli vii viii viii 79 79 84 84 85 85 86 86 87 87 88 88 89 89 90 90 91 91

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92 92 1

93 93 l

93a 1

. - ~

l*

I1 ure 3

Page 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Fast Neutron Fluence (E > 1 MeV) as a Function of 86 Full Power Service Life (EFPY) for Zion Unit 1 3.3.2-4 Fast Neutron Fluence (E > 1 MeV) as a Function of 87 Full Power Service Life (EFPV) for Zion Unit 2 3.3.6-1 Dose Equivalent 1-131 RC Limit versus Percent of Rated Thermal Power 124c

3. 4 -1 High Steam Line Flow Setpoint.

131a 3.11-1 Restricted Area 80undry 226 6.1-1 Minimum Shift Crew Composition 327 I

i 1

l 1

l i

i i

LIST OF FIGURES (Continued) vil Amendment Nos. 101, 91 10580/10s90 i

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Pzge 1.1 Operational Modes 6b I

j 1.2 Survei11ance Frequency Notatlen 6c 3.1 -1 Reactor Protection System-Limiting Operations Conditions and Setpoints 30 l

l 3.1-2 Reactor Protection System Instrument Numbers 33 i

j 3.3.2-1 Zion Unit 1 Reactor Vessel Toughness Data 88 3.3.2-2 Zion Unit 2 Reactor Vessel Toughness Data 89 i

3.3.4-1 In Service Inspection Program 106 i

5 3.3.5-1 Reactor Coolant Systems and Chemistry Specifications 122 i

3.4-1 Engineered Safeguards Actuation System-Limiting Conditions on 129 j

Operation and Setpoints i

3.4-2 Engineered Safeguards System Instrument Numbers 132 i

i 3.7-1 Neutron Flux High Trip Points with Steam Generator Safety Valves 160a l1 Inoperable - Four Loop Operation 3.7-2 Neutron Flux High Trip Points with Steam Generator Safety Valves 160b j

j Inoperable - Three Loop Operation 4

1 3.11-1 Maximum Permissible Concentration of Dissolved or Entralned Noble 226a j

Gases Releases from the Site to Unrestricted Areas in Liqu.id Effluents 1

3.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 228 3.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation 236

[

l 3.14-1 Radiation Monitoring Instrumentation 251 1

3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Bus 268 d

3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 269 LIST OF TABLES 1150t/1151t vill Amendment Nos. 101, 91

LICITING CONBI110N FOR OPERATION j.

SURVEILLANCE REQUIREMECT 3.3.2 PRESSURIZAllDN AND SYSTED li;IEGRI1Y 4.3.2 PRES $UtilAll0N AND SYSitM INILGRIIY 4

A. Heatup and Cooldown A. The reactor coolant temperature and pressure -

shall be determined to be within the limits at j

1he Reactor Coolant System temperature and least once per 30 minutes during system l

pressure (with the exception of the heatup, cooldown, and inservice. leak and i

pressurizer) shall be Ilmited in accordance hydrostatic testing operations.

with the Ilmit lines shown in Figures 3.3.2-1 l

and 3.3.2-2 during heatup, cooldown and inservice leak and hydrostatic testing with:

i

1. a. A maximum heatup rate of 20*F/hr i

appilcable up to and including 180*F RCS Indicated temperature. A maximum heatup rate of 60*F/hr appIlcable for RCS Indicated temperatures greater than 4

180*F.

j

b. A maximum cooldown of 100.*F in any I t

hour period.

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c. A maximum temperature change of 110*F in

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any I hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit j

curves.

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'APPLICA8illTY: At all times.

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With any of the above Ilmits exceeded, j

restore the temperature and/or pressure to within the limit within 30 minutes; i

perform an engineering evaluation to j

determine the effects of the out of limit condition on the structural i

integrity of the Reactor Coolant System; l

determine that the Reactor Coolant System remains acceptable f or continued

j operation or be in at least N000 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce RCS igyg and pressure to less than 200*F and 500 psig, respectively, within the t

i following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

10580/10590 79 8

065FA Amendment Nos. 101, 91 i

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FOR THE SERVICE PERIOD UP TO 34 ET 2M 00 to 800 ISO 200 250 300 350 aos 450 soo INDICATE TDOERATURE (DE. M The 20'F/hr heatup rate is applicable for all RCS indicated temperatures.

The 60*F/hr heatup rate is applicable for RCS indicated temperatures greater than 180*F.

Zion 1 and 2 Reactor Coolant System Heatup Limitations applicable for up to 14 EFPY and Heatup Rates up to 20'F/hr and 60*F/hr. Curves contain margins of 10*F and 60 psig for possible instrument errors.

REACTOR COOLANT SYSTEM HEATUP LIMITATIONS 4

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l Bases 3.3.2 & 4.3.2 FRACTURE TDUGHNESS PROPER 11ES The temperature and pressure changes during heatup and cooldown are Ilmited to be consistent with the requirements given in the ASME Boller and Pressure Vessel Code, Section 111, Appendix G, and 10Cf R50 Appendix G.

I The fracture toughness properties of the ferritic nuterials in the reactor vessel are determined in accordance j

with the NRC Standard Review Plan, ASIM E185-73, and in accordance with additional reactor vessel requirements.

1hese properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section ill of the 4

ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-1924-A, " Basis for Heatup and Cooldown Limit Curves", April 1975.

a I

Heatup and cooldown limit curves are calculated using the most limiting value of the nll-ductility reference temperature, RTNDT, at the end of 15 effective full power years (EFPY) of service Ilfe. The 15 EFPY service 1

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life period is chosen such that the limiting RTMDT at the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material. The selection of such a limiting RTND1 assures that all components in the Reactor Coolant System will be operated conservatively in accordance with app 11 cable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDi; the results of these tests are shown in Tables 3.3.2-1 and 3.3.2-2.

Irradiation can cause an increase in the RIReactor operation and resultant fast neutron (E greater than 1 MeV)

NDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel content of the noterial in question, can be predicted using Regulatory Guide 3

l 1.99, Revision 2, "Ef fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

hestup and cooldown 11mit curves of Figures 3.3.2-1 and 3.3.2-2 include predicted adjustments for this shift in The RTNDT at the end of 15 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section til of the ASME Boller and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detall in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear 4

elastic f racture mechanics (LEFM) technology.

In the calculation procedures, a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/21 is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section lit as the reference flaw, amply exceed the current capabilities of inservice l

inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are i

conservative and provide sufficient safety margins for protection against nonductile failure.

i To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value i

of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shif t, j

ARTHDI, corresponding to the end of the period for which heatup and cooldown curves are generated.

10500/10590 90 0651A Amendment Nos. 101, 91

l i*

The ASME approach for calculating the allowable 11mit curves for various heatup and cooldown' rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the references stress intensity f actor, Kgg for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code.

The j

Kgg curve is given by the equation:

4 j

KIR = 26.78 + 1.223 exp {0.014.5(T-RTNOT + 160)]

(1)

I j

Where:

K g is the reference stress intensity factor as a function of the metal temperature T and the metal i

nil-ductility reference temperature RINDT.

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

j CKIM + Kit 5 Kgg (2) i j

Where:

Kgg =

the stress intensity factor caused by membrane (pressure) stress, j

Kit =

the stress intensity factor caused by the thermal gradients, q

l Kgg =

constant provided by the Code as a function of temperature relative to the RTHOT of the j

material, l

C 2.0 for level A and B service limits, and

=

i C

1.5 for inservice hydrostatic and leak test operations.

=

At any time during the heatup and cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RINDT, and the reference fracture toughness curve.

The thermal 1

stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, Kit, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

j Figures 3.3.2-1 and 3.3.2-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the J

heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

1 The leak test limit curve shown on the heatup curves (Fig. 3.3.2-1) represent minimum temperature requirements at j

the leak test pressure specified by ASME Section 111 and the NRC Standard Review Plan NUREG-0800.

Allowable combinations of pressure and temperature for specified temperature change rates are below and to the right of the limit ilnes shown. Limit Ilnes for cooldown rates between those presented may be obtained by 1

interpolation.

i l

10580/10590 91 Amendment Nos. 101, 91

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HEAIUP 1hree separate calculations are required to dhtermine the limit curvas.for finite heatup rates.

As is dont in the

~

cooldown En21ysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall.

1he thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure.

The metal' temperature at the crack tip lags the coolant temperature; therefore, the K p for the 1/4T track during heatup is lower than the K p for the 1/41 track during i

i steady-state conditions at the same coolant temperature.

During heatup, especially at the end of the transient, j

conditions may exist such that the ef fects of compressive thermal stresses and dif ferent Kgg's f or steady-state and finite heatup rates do not of f set each other and the pressure-temperature curve based on steady state l

conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/41 flaw is.

considered.

Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the j

lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

j The second position of the heatup analysis concerns the calculation of pressure-temperature limitations for the l

case in which a 1/4T deep outside surface flaw is assumed.

Unitke the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature i

and thus tend to reinforce any pressure stresses present.

These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot l

be defined.

Rather, each he.mtup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate i

situations, the final limit curves are produced as follows.

A composite curve is constructed based on a j

point-by-point comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

j 1he use of the composite curve is necessary to set conservative heatup limitations because it is possible for i

conditions to exist such that over the course of the heatup ramp the controlling condition switches from the l

Inside to the outside and.the pressure limit must at all times be based on analysis of the most critical criterion.

i i

lhe heatup rates of the Reactor Coolant System are limited to 20*F/hr for RCS indicated temperatures equal or less than 180*F and 60*F/hr for RCS indicated temperatures greater than 180*F to comply with the requirements of 10CFR50 Appendix G.

j Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors i

in the pressure and temperature sensing instruments by the values indicated on the respective curves.

}

C001DOWN for the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is 1

always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which i

l 10580/10590

{

92 06siA Amendment Nos. 101, 91

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l l

Increase with increasing cooldown rates. All ble press %re-temperature relations are generated ter both stecdy-state cod finite cooldown rate situations. From these relations, composite limit curves are constructed i

for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is i

based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the i

material temperature at the tip of the assumed flaw.

i temperature than the fluid adjacent to the vessel ID. During cooldown, the 1/41 venel location is at a higher This condition, of course, is not true for the steady-state -

i situation.

It follows that at any given reactor coolant temperature, the AT developed during cooldown results 4

j in a higher value of K g at the 1/41 location for finite cooldown rates than for steady-state operation.

i i

Turthermore, if conditions exist such that the increase in Kg during cooldown will be greater than the steady-state value. g exceeds Kit, the calculated allowable pressure

}

l 1he above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

q PRESSURIZER LIMITS Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

HYOROSTATIC TESTING LIMIT CURVE l

t Allowable pressure-temperature relationships for leak and hydrostatic testing are also calculated using methods i

derived f rom Non-Mandatory Appendix G2000 in Section III of the ASME Boller and Pressure Vessel Code.

The approach specified is the same as described for heatop and cooldown limits except that the safety factor on Kgn i

is reduced to 1.5 and there are no significant thermal transients or gradients. Thus the governing equation for the leak and hydrostatic testing analysis is:

l s

1.5 Kgg < Kgg INADVERTANT SAFETY INJECil0N i

in the event of an inadvertant safety injection actuation, the affected reactor will trip immediately, place the j

reactor in the hot shutdown condition.

After f>0 seconds safety injection may be reset and injection terminated as j

required. An inspection of the primary system while at hot shutdown will prevent possible degradations in the primary system from undergoing further immediate thermal shock imposed during a cooldown.

If degradations in the primary system are discovered, an orderly controlled cooldown will be planned to minimize the elfects of thermal shock on these degradations on the affected unit.

i 10sa0/10590 93 0651A Amendment Nos. 101, 91 i

i 1

1

,y

_e REFERENCES I,

1.

ASME Boller and Pressure Vessel Code, Section Ill,1976 Summer Addenda.

2.

WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves" April 1975.

3.

ASME Boller and Pressure Vessel Code,Section III, N-331.

4.

ASME Boller and Pressure Vessel Code, Section 111, N-415.

5.

FSAR, Chapter 4.3.

6.

WCAP-8724, "ASME 111, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 1 Reactor Vessel".

7.

WCAP-8727 "ASME Ill, Appendix G Analysis of the Commonwealth Edison Company Zion Unit-2 Reactor Vessel".

8.

WCAP-10677, " Adjoint Flux Program for Zion Units 1 and 2".

9.

Regulatory Guide 1.99 Revision 2.

10.

Code of Federal Regulations, 10CFR50 Appendix G, " Fracture Toughness Requirements."

11.

WCAP-11247, "Heatup and Cooldown Limit Curves for the Commonwealth-Edison Company Zion Units 1 and 2 Reactor Vessel".

12.

WCAP-10962, " Zion Units 1 and 2 Reactor Vessel Fluence and Ripts Evaluations".

13.

CWE-86-563, " Low Temperature Overpressure Protection System Setpoing Analysis", August 26, 1986.

14.

CWE-865-588, low Temperature Overpressure Protection System Setpolng Analysis Extensions" October 77, 19Bri.

10580/10590 93a 0657A Amendment Nos '101, 91 s