ML20094Q295

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Amends 158 & 148 to Licenses DPR-77 & DPR-79,respectively, Incorporating New RCS pressure-temp Limit Curves Applicable Up to 16 EFPYs
ML20094Q295
Person / Time
Site: Sequoyah  
Issue date: 03/31/1992
From: Hebdon F
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20094Q302 List:
References
NUDOCS 9204090261
Download: ML20094Q295 (2)
v • d • e


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JLNX SSEE VALLEY AUTHORITY DOCKET NO. 50-327 SE0VOYAH NKLEAR PL ANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 158 License No. OPR-77 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 1, 1991, and superseded September 6, 1991, complies with the standards and requi p ments of the Atomic Energy Act of 1954, as amended (the Act), and'the Commission's rules and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; 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.

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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 2.C.(2) of facility Operating License Itu. OpR-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 158, 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 its date of issuance, to be implemented within 30 days.

FOR THE flVCLEAR REGULATORY COMMISSION

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Attachment:

Changes to the Technical Specifications Date of Issuance: March 31, 1992 3

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4h ATTACHMENT TO LICENSE AMENDMENT NO.158 FACILITY OPERATING LICENSE N0. OPR-77 QQCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contoln marginal lines

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indicating the area of change. Overleaf pages* are provided to maintain document completoness.

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CURVES AFPLICABLE FCR COOLDOUN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY, MARGINS OF CO PSIG AND 10'F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.

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,R_EACTOR COOLANT SYSTEM BASES 3/4.4.9 PPESSURE/ TEMPERATURE LIMITS The temperature and pressure changes during heatup and coeldown are limited to be consistent with the requirements given in the A C Boiler and Pressure Vessel Code, Section Ill, Appendix G.

1)

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the first full power service period.

a)

Allowable combinations of pressure and temperature for specific temp st.ure change rates are below and to the right of the limit lines shown.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

b)

Figures 3.4-2 and 3.4-3 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 heatup and cooldown rates that can be achie.ed over certain pressure-temperature ranges.

2)

These limit lines shall be calculated periodically using methods provided below.

3)

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

4)

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200 F/hr respectively.

The spray shall not be used if the temperature difference between the pressurizer ar'i the spray fluid is greater than 560*F.

5)

System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

10 CFR 50, Appendix G, addressed metal t.mperature of the closure head flange and vessel regions.

Appendix G states that the minimum metal tem-perature of the closure flange region should be at least 120 degrees Fahrenheit (F) higher than the limiting RT for this region when the pressure exceeds 20 percent of the preserv b hydrostatic test pressure (561 pounds per square inch gauge (psig) for Westinghouse Electric Corpora-tion plants).

For SQN, Unit 1, the minimum temperature of the closure flange and vet,sel flange regions is 90 degrees F since the limiting initial RT for the closure head f tange is -40 degrees F (see Table B 3/4.4-1).

Th b numbers (561 psig and 90 degrees F) include a margin foi instrumenta-tion error of 10 degrees F and 60 psig.

The SQN Unit I heat up and cool-down curves shown in Figures 3.4-2 and 3.4-3 ;re not impacted by this regulation.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, and ASTM E185-82, and in accordance with additional reactor vessel require-ments. These properties are then evaluated in accordance with Appendix G to 10 CFR 50 and Appendix G of the 1986 ASME Boiler and Pressure Vessel Code, Section 111, Division 1 and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

SEQUOYAH - UNIT 1 B 3/4 4-6 Amg#entNo.12,133

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4 REACTOR COOLANT SYSTEM BASES Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of The$T,FPYservice l

16 ef fective full power years of service life.

E life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT ofNolimitingunirradiated g

material.

TheselectionofsuchalimitikTRT assures that all components in the Reactor Coolant System will NToperated conservatively in accordance with applicable Code requirements.

The reactor vessel materialb have been tested to determine their initial RT

the results of these tests are shown in Table 9 3/4.4-1.

Reactor opeNtionandresultantfastneutron(Egreaterthan1MEV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature,baseduponthefluhNe. of the material in question, has been predicted using Regulatory Guide 1.99, Revision 2 and a peak surface fluence of 1.94 x 1018 n/cn3 for 16 effective full power years (Reference WCAP 12970, "Heatup and Cooldown Limit Curves for Normal Operation,"

June 1991).

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for thi, shift in RT at the end of 16 EFPY, as well as adjustments for possib1 errorsint$Tpressure and temperature sensing instruments.

Values of delta RT determined in this manner may be used until the results from the mhNrial surveillance program, evaluated according to ASTM E185, are available.

The first capsule was removed at the end of the first core cycle.

Successive capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H.

The heatup and cooldown curves and the low temperature overpressure protection set-points must be recalculated when the delta RT determined from the surveillancecapsuleexceedsthecalculateddhNaRT f r the equivalent NDT capsule radiation exposure.

SEQUOYAH - UNIT 1 8 3/4 4-7 Amendment No.157, 158

REACTOR COOLANT SYSTEM This page 1,tentionally deleted.

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REACT 0p COOLANT SYSTEP PASES

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Allowable presue -temperature relationships for various heatup and cooldown rates are calculated usina rethods derived from Appendix G in Section !!! of the ASMF Poller and Pressure Vessel Code as reouired by Appendix G to 10 CFR Part 50 and these rethods are discussed ira detail in WCAP-792a-A.

The ceneral rethod for calculating beatup and cooldown limit curves is based upon the principles of the linear elastic fracture rechanics (LETP) technology.

In the calculation procedures a semi-elliptical surf ace defect with a depth of one cuarter of the wall thickness. T, and a lenath of 3/2T is assumed to exist at the inside of the vessel well as well as at the outside of the vessel wall.

The dirensions of this postulated crack, referred to in Appendix C of AS"E 111 as the reference flaw, amely exceed the current capabilities of inservice inspection technioves.

--Therefore, the reactor operation lirit curves developed for this reference' crack are conservative and provide sufficient safety rargins for protection aoainst non-ductile failure. To assure thai the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limitino value of the nil ductility reference temperature, PTNDT' IS used and this includes the radiation induced shif t, delta PTNDT, corresponding to the end of the period for which beatup and cooldown curves are aenerated.

The ASFT approach for calculatino the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined themal and pressure stresses at any tire during heatue j

or cooldown cannot be creater than the reference stress intensity f actor, ryp, for the retal temperature at that tire.

K is obtained from the reference jp fracture touchness curve, defined in appendix 0 to the AS"f Code. The Vjp curve is oiven by the ecuation:

Kjp = 26.7P + 1.223 exp [0.014E(T Di@T + l#0II III Sf000YAH - UNIT 1 P 3/4 4-10

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REACiORCOOLANTSYSTEJj BASES thermal stresses and different K

's for donotofIseteachotherandthe;hressuresteady-stateandfiniteheatuprates temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite haatup rates when the 1/4T flaw is considered.

Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature 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 be defined.

Rather, each heatup 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 situations, the final limit curves are pro-duced as follows.

A composite curve is constructed based on a point-by point comparison of the steady-state and finite heatup vate data.

At any given tem-perature, the allowable pressure _is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

The ieak test limit curve shown in Figure 3.4-2 represents the minimum temperature requirements at the leak test pressure specified by applicable codes.

The leak test limit curve was determined by methods of Branch Technical Position MTFB 5-2 and 10 CFR 50, Appendix G.

The criticality limit curve shown in Figure 3.4-2 specifies pressure-temperature limits for core operation to provide additional margin during actual power production.

The pressure-temperature limits for core operation (except for low power physics tests) require the reactor vessel to be at a temperature equal to or higher than the minimum temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum pressure-temperature curve for heatup and cooldown.

The maximum temperature for the in-service hydrostatic test for the SQN Unit 1 reactor vessel is 327 degrees F.

A vertical line at 327 degrees F on the pressure-temperature curve, intersecting a curve 40 degrees F higher than the pressure-temperature limit curve, constitutss the limit for core operation for the reactor vessel.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature

-sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile f ailure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

SEQUOYAH UNIT - 1 B 3/4 4-13 Amendment No.158 l

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1[@l[H(( VALLEY AUTHORITY DOCKET NO. 50-328 SE0VOYAH NUCLEAR PLANT. UNIT 2 AMENDMDil.10 FAClllTY OPEPATING llCENSE Amendment No.148 License No. OPR-79 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 1, 1991, and superseded September 6, 1991 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 Chapth 1; 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.

4.

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-79 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.148, 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 its date of issuance, to be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

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Frederick J. Hebdon, Director Project Directorate !! 4 i

Division of Repctor Projects - 1/11 i

Office of Nuc1Far Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 31 1992 l

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ATTACHMENT T0 t ICEN$[ A!!(NDMENT N0.148 FAClllTY OPERATING LIC.MSE NO. DPR-72 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Overleaf pages* are provided to maintain document completeness.

REMOVE JHSERT 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 B3/4 4-7 B3/4 4-7 B3/4 4-8 B3/4 4-B B3/4 4-9 B3/4 4-9 B3/4 4-10 B3/4 4-10 B3/4 4-13*

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APPLICABLE UP TC 16 Er;Y Arendnent f;o. 148 SEQUOYAH - UNIT 2 3/4 4-30

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O 25 50 100 200 300 400 500 IN;:CATED TEMPERATURE ('F) l FIGURE 3.4-3 SEQUOYAH UNIT 2 REA['[O [OCLANT $Y$i[M COOLDOnN LIMITATIONS AcPIc!:AeLE up ic it Erc<

Arendment No.148 i

SEOUOYAH - UNIT 2 3/4 4-3I

RpCTORCOOLANTSYSTEM BASE 3 PRESS _URE/ TEMPERATURE LIMITS (Continued) l 5)

System preservice hydrotests and in service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler j

and Pressure Vessel Code,Section XI.

10 CFR 50, Appendix G, addresses metal temperature of the closure head flange and vessel regions.

Appendix G states that the minimum metal i

temperature of the closure flange region should be at least 120 degrees l

Fahrenheit (F) higher than the limiting RT for this region when the pressure exceeds 20 percrt of the preservk[ hydrostetic test pressure (561 poundt per square inct p ge (psig) for Westinghc.4e Electric Corpora-tion plants).

For SQN, Urnt ?, the minimum temperature of the closure flange and vessel flange regions is 117 degrees F since the ihniting initial RT for the closure head flange is -13 degrees r see Table B 3/4.4-1).NNesenumbers(561psigand117degreesF)intad i margin s

for instrumentation error of 10 degrees F and 60 psig.

Tne SQN Unit 2 heat up and cooldown curves shown in Figures 3.4-2 and 3.4-3 are not impacted by this regulation, y

The fracture toughass properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-82, and _in accordance with additional reactor vessel require-ments.

These properties are then evaluated in accordance with Appendix G to 10 CFR 50 and Appendix G of the 1986 ASME Boiler and Pressure Vessel Code,Section III, Division 1 and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

Heatup and cooldown limit curses are calculated using the most limiting value of the nil-ductility reference temperature, RT at the end of The$T,FPYservice l

16 effective full power years of service life.

E life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT fNelimitingunirradiated N

material.

TheselectionofsuchalimitikTRT assures that all components in tht. ReactorCoolantSystemwillUSToperated conservatively in accorMnce with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT

the results of these tests are shown in Table B 3/4.4-1.

Reactor ophItionandresultantfastneutron(Egreaterthan1MEV)irradiationcan cause an increase in the RT Therefore, an adjusted reference tempera-ture, based upon the fluenchDdf the material in question, has been predicted using Regulatory Guide 1.99, Revision 2 and a peak surface fluence of 0.864 x 1018 n/cm2 for 16 effective full power years (Reference WCAP 12971, "Heatup and Cooldown Limit Curves for Normal Operation," June 1991.

The i

heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include pre-dicted adjustments for this shift in RT at the end of 16 EFPY, as well asadjustmentsforpossibleerrorsintNhTpressure and temperature sensing instruments.

SEQUOYAH - UNIT 2 B 3/4 4-7 Amendment No.148

REACTOR COOLANT SYSTEM BASES-PRESSURE / TEMPERATURE LIMITS'(Continued)-

Values of ARTNDT determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM ElB5, are available.

The first capsule was removed at the end of the first core cycle.

Successive capsules will be removed in accordance with the require-ments of ASTM E185-82 and 10 CFR 50, Append'< H.

The heatup and cooldown curves and the low-temperature overpressure protection setpoints must be

- i recalculated when the ART determined from the surveillance capsule NDT exceeds the calculated ART for the equivalent capsule radiation exposure.

ND1 Allowable pressure-temperature relationships for various heatup and cool-down rates are calculated c;ing methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Apptr-dix G to 10 CFR Part 50-and these methods are discussed in detail in WCAF-7924-A.

The general method for calculating heatup %nd cooldown limit curves is based upon the principles of the linear elastic-fracture mechanics (LEFM) technology.

In the calculation procedures a semi-elliptical surface defect with a depth of one quarter of the wall thickness, T,.and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. -The dimenstions of this postulated crack, referred to in Appendix G of ASME III as the reference flaw, amply

. exceed the current: capabilities of inservice inspection techniques.

Therefore, the reactor cperation limit curves developed for this reference crack lare conservative and provide-sufficient safety margins for protection against non-ductile failure.

To assure that the radiation embrittlement-effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT,_is-used and this includes the radiation induced shift, ARTNDT, c rresponding to the end of the period for which heatup and cooldown curves are generated.

1 6

.SEQUOYAH - UNIT 2 8 3/4 4-8 Amendment No. 147, 148

i TABLE B 3/4.4-3 h

on5 SEQUOYAH-UNIT 2 REACTOR '

6Hhy

'ATA g

E i

J 1

i E

50 t.

AVERAGE UPPER SHELF j

co HEAT MATERIAL.

CU Ni NDTT MIL TEMP *F FTL8 l

NT COMPONENT NO.

GRADE

  • F PMWD' NMWD2 NDT PMWD' NMWDZ

.p CL Hd. Dome 52899-1 A533BCL1

-13 28 48*

-12 75' 141.g*

125.

CL Hd. Ring A508CL2 5

34 54*

5

-13

<-67*

<-67

-13 Hd Flange 4890 A508CL2

-22

-47

-27

-22 155 5*

Vessel Flange 4832 A508CL2 3

Inlet Nozzle 4868 A508CL2

-22 41 61*

1 79a Inlet Nozzle 4872 A508CL2

-22 12 32*

-22 108a Inlet Nozzle 4877 A508CL2

-31 1

21*

-31 113a Inlet NOIzle 4886 A508CL2

-31

-52

-32*

28 138 cz, 8

1 Outlet Nozzle 4867 A508CL2

-31 19 39*

-21 85 w

d Outlet Nozzle 4873 A508CL2

-22 21 41*

-19 76 4

Outlet Nozzle 4878 A508CL2

-40

-6 14*

-40 105' 143*5' Outlet Nozzle 4887 A508CL2

-22

-11 9*

-22 5

25 45*

5 104 Upper Shell 4885 A508CL2 Inter Shell 4853 A508CL2 0.13 0.74 -22 19 70 10 138 93 y

Lower Shell 4994 A508CL2 0.14 0.76 -40 8

38

-22 140 5 100 3

Trans. Ring 4879 A508CL2 5

27 47*

5 98 Bot. Hd. Rim 52835-1B A533BCL1

-4 48 68*

8 81'a Bot. Hd. Rim 52835-1B A533BCL1

-22 25 45*

-15 81' Bot. Hd. Rim 52899-2 A533BCL1

-13 39 59"

-1 62 8

E 80t. Hd.

5297-1 A533BCL1

-31 14 34*

-26 99.5 Weld Weld 0.13 0.11

-4 14

-4 101 17

-13 120 l

HAZ HAZ

-13 2

e 5

i Paralled to Major Working Direction 2 Normal to Major Working Direction L

S

  • Estimate based on USAEC Regulatory Standard Review Plan, Section 5.3.2 MTEB 5-2 a % Shear not reported

l This page intentionally deleted.

SEQUOYAH - UNIT 2 B 3/4 4-10 Amendment No.147' 148

l REACTOR COOLANT SYSTEM BASES PRESSURE /TrMPERATURE LIMITS (Continued)

The 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.

HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates.

As is don? in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditicas as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vesce! wall.

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

The metal temperature at the for the 1/4T crack crack tip lags the coolant temperature; therefore, the KIR for the 1/4T crack during steady-state during heatup is lower than the KIR conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K

's f r steady-state and finite heatup rates IR do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered.

Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature 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 be defined.

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

SEQUOYAH - UNIT 2 B 3/4 4-13 1

TcACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Contir,ued)

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows.

A composite curve is constructed based on a 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.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

The leak test limit curve shown in Figure 3.4-2 represents the minimum temperature requirements at the leak test pressure specified by applicable codes.

The leak test limit curve was determined by methods of Branch Tecnnical Position MTEB 5-2 and 10 CFR 50, Appendix G.

The criticality limit curve shown in Figure 3.4-2 specifies pressure-temperature limits for core operation to provide additional margin during actual power production.

The pressure-temperathe limits for core operation (except for low power physics tests) require the reactor vessel to be at a temperature equal to or higher than the minimum temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum pressure-temperature curve for heatup and cooldown.

The maximum temperature for the in-service hydrostatic test for the SQN Unit 2 reactor vessel is 274 degrees F.

A vertical line at 274 degrees F on the pressure-temperature curve, intersecting a curve 40 degrees F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments ty the values indicated on the respective curves.

Although the pressurizer operates in temperature renges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and-testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1).

Components of ths reactor coolant system were designed prior to the issuance of Section XI of the ASME Boiler and Pressure Vessel Code.

These components will be tested to the extent practical within the limitations of the original plant design, geometry and materials of construction.

SEQUOYAH - UNIT 2 B 3/4 4-14 Amendment No.148 l

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