Text

Amends 116 & 85 to Licenses NPF-14 & NPF-22,respectively, Revising TS Re Reactor Pressure Vessel pressure-temp Curves
	ML17157B015
Person / Time
Site:	Susquehanna
Issue date:	01/10/1992
From:	Chris Miller; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17157B016	List: ML17157B015; ML17157B017;
References
	NUDOCS 9201220289
	Download: ML17157B015 (32)
	v • d • e

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLYANIA PO>lER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIYE INC.

DOCKET NO. 50-387 SUSQUEHANNA STEAN ELECTRIC STATION UNIT I AMENDllENT TO FACILITY OPERATING LICENSE Amendment No. 116 L'.cense No. NPF-14 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power Light Company, dated April 18, 1991 and its supplements dated September 27, 1991 and January 3, 1992, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 regulations of the Commission; C.

There

'.s reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health, and safety of the publ'.c, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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 w'.th 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 Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14

'.s hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 116 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection'lan.

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This license amendment is effective as of its date of issuance and shall be implemented within 30 days of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Date of Issuance:

January 10, 1992

l

ATTACHMENT TO LICENSE AMENDMENT NO. 116 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness.*

REMOVE 3/4 4-15 3/4 4-16 3/4 4-17 3/4 4-18 3/4 4-19 3/4 4-20 B 3/4 4-3*

B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 B 3/4 4-7 INSERT 3/4 4-15*

3/4 4-16 3/4 4-17 3/4 4-18 3/4 4-19 3/4 4-20 B 3/4 4-3*

8 3/4 4-4 B 3/4 4-5 B 3/4 4-6 B 3/4 4-7

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TABLE 4.4.5-l PRIHARY COOLANT SPECIFIC ACTIVITY SAHPLE AND ANALYSIS PROGRAH TYPE bF HEASUREHENT AND ANALYSIS l.

Gross Beta and Gamma Activity Determination 2.

Isotopic Analysis for DOSE E(UIVALENT I-131 Concentration 3.

Radiochemical for K Determination SAHPLE AND ANALYSIS FRE UENCY At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

At least once per 3) days At least-once per. 6 months*

OPERATIONAL CONDITIONS IN NICH SAHPLE AND ANALYSIS RE UIRED l, 2, 3 4.

Isotopic Analysis for Iodine a) b)

At ~least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds a limit, as required by ACTION b.

At least one sample, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following the change in THERHAL POWER or off-gas level, as required by ACTION c.

l¹, 2¹, 3¹, 4¹ l, 2 5.

Isotopic Analysis, of an Off-gas Sample Including guantitative Measurements for at 1east Xe-133, Xe-135 and Kr-88

--At least once per,31 days Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

¹Until the specific activity of the primary coolant system is restored to within its limits.

REA T R LANT Y 44 LIMITINGCONDITIONFOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limitedin accordance with the limit lines shown on Figure 3.4.6.1-1 for hydrostatic or leak testing, heatup by non-nuclear means, cooldown foHowing a nuclear shutdown and low power PHYSICS TESTS, and operations with a critical core other than low power PHYSICS TESTS, with:

a. A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 100'F in any 1-hour period,

c. A maximum temperature change of less than or equal to 20'F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limitcurves, and

d. The reactor vessel flange and head flange temperature greater than or equal to 70'F when reactor vessel head bolting studs are under tension.

A L

A al

'ith any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limitcondition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUIDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHVH3OWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCERE UIREMENTS 4.4.6.1.1 During system

heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limitlines of Figure 3.4.6. 1-1 at least once per 30 minutes.

SUSQUEHANNA - UNIT 1 3/4 4-16 Amendment No. U.6

VEILLA CERE UIREM%%TS Continued 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined, to be to the right of the criticality limitline of Figure 3.4.6.1-1 within 15 minutes prior to the withdrawal ofcontrol rods to bring the reactor to criticality and at least once per 30 minutes during system heatup.

4.4.6.1.3 The reactor vessel material specimens shall be removed and examined to determine reactor pressure vessel fluence and embrittlement as a function of time and THERMALPOWER as required by 10 CFR 50, Appendix H. The results of these fluence and embrittlement determUiations shall be used to update the curves of Figure 3.4.6.1-1.

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70 F:

a. In OPERATIONAL CONDITION4 when reactor coolant system temperature ls:

1.

M 100'F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

M 80'F, at least once per 30 minutes.

b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

SUSQUEHANNA - UNIT 1 3/4 4-17 Amendment No. 116

1600 1400 A

8 C

1200 1000 800 600 NON-BELTUNE UMITS WlTH

~ RECIRC. INLET NOZZLE RTwtOF 404F FOR BdcC CRD PENETRATION RTg)g OF 34 F FOR A 400 312 ps' SYSTEM HYOROTEST UMIT WlTH FUEL IN VESSEL B NON-NUCLEAR HEATING LIMIT C NUCLEAR (CORE CRITICAL)

LIMIT 200 80lTV8'4tÃ CURVES A, 8 and C ARK VAUD FOR 32 EFPY OF OPERATION 100 200 300 400 500 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F) 600 REACTOR VESSEL PRESSURE VS. MINIMUMVESSEL TEMPERATURE FOR UNIT 1 Figure 3.4.6. 1-1 CIJRAIJI RAMNL UHIT 1

AMENDMENT NO. 116

This Page Has Been Intentionally Left Blank SUS UEHANNA-UNIT I 3/4 4-19 Amendment No. 116

This Page Has Been Intentionally Left Blank SUSQUEHANNA - UNIT 1 3/4 4-20 Amendment No. >>6

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued)

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.

When the conductivity is within limits, the pH, chlorides and other impurities affecting, conductivity must also be within their acceptable limits.

With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE E(UIVALENT I-131, but less than or equal to 4.0 micro-curies per gram DOSE E(UIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

AS 44 Allcomponents in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The. various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specifled heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions, i.e., no thermal

stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination ofpressure-temperature limitations forthe case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which 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.

'I The reactor vessel materials have been tested to determine their initialRT~T. The results of these tests are shown in Table B 3/4.4.6-1.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation willcause an increase in the RT~T.

Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Regulatory Guide 1.99, Revision 2, Radiation Embrittlement ofReactor Vessel Materials." The pressure/temperature limitcurve, Figure 3.4.6.1-1, includes predicted adjustments for this shift in RT~T for the 32 EFPY condition.

The actual shift inRT~T ofthe vessel material willbe established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limitcurves ofFigure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99. Revision 2.

SUSQUEHANNA - UNIT 1 B 3/4 4-4 Amendment No. 116

BAS (Continued)

The pressure-temperature limitlines shown in Figure 3.4.6.1-1 curves C and A, for reactor criticalityand for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for hiservice leak and hydrostatic testing.

Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line brcak.

Only one valve in each line is required to maintain the integrity of the containment.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

44 The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components willbe maintained at an acceptable level throughout the lifeof the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XIof the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through 1972.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed 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 NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).

4 4 ID L

ATREM AL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

A - UNI'I' B 3/4 4-5 Amendment No. 116

AS TABLEB 446-1 I:;"".'Mfhn'e':"".'-:','g i",;:;Corn'paitent:":;.;

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-50 33 N/A

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~OTE:

~ These values are given only for the benefit of calculating the 32 EFPY RT~T MATEMAL TYPE:OR-"'"-'"-":%EL'9:SEAM!R9::" "~l'-::

'hajj(HEGBI3'F; STARTlNG~j Shell Ring Bottom Head Dome Bottom Head Torus Top Head Dome Top Head Torus Top Head Flange

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Vessel Flange Feedwater Nozzle Recirculation Inlet Nozzle Weld Closure Studs SA-533 GR B DL1 SA-508, CL.2 SA-540 GR B24 C1232-2 C9942-2 C9942-2 C9220-2 C9355-1 N/A N/A Q2Q49W Q2Q49W No CNVS Available 82552

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Fast Neutron Fluence (E>) Mev) at l.D. Surface as a Function of Service Life*

Boses Figure B 3/4.4.6 1

At 90% of RATED THERMAL POWER ond 90K availability

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POWER 5 LIGHT COMPAI'IY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 85 License No.

NPF-22 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by the Pennsylvania Power 5 Light Company, dated April 18, 1991 and its supplements dated September 27, 1991 and January 3, 1992, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 regulations of the Commission; C.

D.

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 set forth in 10 CFR Chapter I; 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 Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows:

(2)

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i 1P The Technical Specificat'.ons contained in Appendix A, as revised through Amendment No. 85 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

PPSL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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This license amendment is effective as of its date of issuance and shall be inplemented within 30 days of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Charles L. Hiller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 10, 1992

ATTACHMENT TO LICENSE AMENDMENT NO. 85 FACILITY OPERATING LICENSE NO.

NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness'.*

REMOVE 3/4 4-17 3/4 4-18 3/4 4-19 3/4 4-20*

B 3/4 4-3*

B 3/4 4-4 B 3/4 4-5*

B 3/4 4-6 B 3/4 4-7 INSERT 3/4 4-17 3/4 4-18 3/4 4-19 3/4 4-20*

B 3/4 4-3*

B 3/4 4-4 B 3/4 4,-5*

B 3/4 4-6 B 3/4 4-7

1

RE ontinued The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limitline of Figure 3.4.6. 1-1 within 15 minutes prior to the withdrawal ofcontrol rods to bring the reactor to criticalityand at least once per 30 minutes during system heatup.

I The reactor vessel material specimens shall be removed and examined to determine reactor pressure vessel fluence and embrittlement as a function of time and TE&RMALPOWER as required by 10 CFR Part 50, Appendix H. The results of these fluence and embrittlement determinations shall be used to update the curves of Figure 3.4.6.1-1.

The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70'F:

a.

In OPERATIONALCONDITION4 when reactor coolant system temperature ls:

1.

M 100'F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

M 80'F, at least once per 30 minutes.

b.

Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

ANNA-UNIT2 3/4 4-17 Amendment No. 85

1600

'I 400 A

8 C

1200 Q

1000 V)

V) 800 O

600 I

g 400 Q

312 ps'ON-BELTUNE UMITS WITH STEAM OUTLET NOZ2LE RT mrOF 30 F FOR 88cC; RECIRC. OUTLET NOZ2LE RT~TOF 24 F FOR A A SYSTEM HYOROTEST UMIT WITH FUEL IN VESSEL.

8 NON-NUCLEAR HEATING LIMIT C NUCLEAR (CORE CRITICAL)

LIMIT 200 BOLTut 70 F CURVES a, 8 ANO C ARE VAUO FOR 32 EFPY OF OPERATION 100 200 300 400 500 MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F) 600 REACTOR VESSEL PRESSURE VS. MINIMUMVESSEL TEMPERATURE FOR UNIT 2 Figure 3.4.6.1-1 4

4 18 mzmzmT HO. a5

This Page Has Been Intentionally Left Blank SUSQUH~lNA - UNIT 2 3/4 4-19 Amendment No. 85

REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1040 psig.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2".

ACTION:

With the reactor steam dome pressure exceeding 1040 psig, reduce the pressure to less than 1040 psig within 15 minutes or be in at least.HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1040 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"Not applicable during anticipated transients.

SUSQUEHANNA - UNIT 2 3/4 4"20

REACTOR COOLANT SYSTEM BASES 3/4. 4. 4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.

When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits.

With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4. 4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 microcuries per gram DOSE EtlUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

= Information obtained on iodine spiking will be used to assess the param-eters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves preven'ts the release of activity to the environs should a steam line rupture occur outside contain-ment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

SUS(UEHANNA - UNIT 2 B 3/4 4-3 Amendment No. 6O

R L

Y BASKS Allcomponents in the reactor coolant system are designed to withstand the effects ofcyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall'produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions, i.e., no thermal

stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination ofpressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which 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.

The reactor vessel materials have been tested to determine their initialRT~T. The results of these tests are shown in Table B 3/4.4.6-1.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation willcause an increase in the RT~T. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials". The pressure/temperature limitcurve, Figure 3 4.6.1-1 includes predicted adjustments for this shift in RT~T for the 32 EFPY condition.

The actual shift inRT~T ofthe vessel material willbe established periodically during operation by removing and evaluating, in accordance with ASTME185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wallofthe reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limitcurves ofFigure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and recommendations ofRegulatory Guide 1.99, Revision 2.

SUSQUEHANNA - UNlT2 B 3/4 4-4 Amendment No. 85

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown in Figure 3.4.6. 1-1, curves C

and A, for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through 1972.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(1)

~

3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

BASES TABLEB 3 4 4.6-1 EA OR VESSEL TOUG S

-:<<Heat/Lat!j<,"~'CU{%)::,,"',:::ANl{%) @RTNDi'(t@~

,,>Qgggfg'j'i'":

Plate SA-533 GR B CL.1 6C1053/1 0.10 0.58

+10 N/A

+50 Weld N/A 624263/

E204A27A 0.06 0.89

-20 50 N/A

+30

~OTEt

~ These values are given only for the benefit of calculating the 32 EFPY RTNDT

~=.', XEGHRST 8FARTNG'i-:;',;

'$"3@8)gg'~"+"'~{~p}>

t><'I':+~:y'hell Ring //5 Bottom Head Dome Bottom Head Torus Top Head Side Plates Top Head Flange Vessel Flange Feedwater Nozzle Steam Outlet Nozzle Weld Closure Studs SA-533 GR B Cl.1 SA-508, Cl.2 Bottom Head Flanges to Shell Top Head Other Non-Beltline SA-540 GR B24 All C0472 C0473-1 125H446 2L2393 Q2Q62W Q2Q64W All All All All

+10

-10

+30

-20

-20 0

Meet requirements of 45 ft-lbs and 25 mils lateral expansion at +10'F

7 IO X

6 t4 E

4 C

G:

Z 0

0 20 40 Service Life (Years*)

Fast Neutron F)uence (E)1 Mev) at LD. Surface as a Function of Service Life*

Bases Figure B 3/4.4.6-1 At 90% of RATED THERMAL POWER and 90% availability

ML17157B015

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