ML20245K670

Proposed Tech Specs,Updating Pressure/Temp Curves & Associated Info

ML20245K670
Person / Time
Site:	LaSalle
Issue date:	06/21/1989
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20245K665	List: ML20245K662 ML20245K670
References
0167T, 167T, NUDOCS 8907050216
Download: ML20245K670 (18)

Text

-

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE'/ TEMPERATURE LIMITS REUTOR COOLANT SYSTEM i

LIMITING CONDITION FOR OPERATION I

go 3.4 6 e l-In 3.4.6.1 The reactor coolant system temperature and pressure shall be limited

)!

in accordance with the limit lines shown on Figure 3.4.6.1-1*(1) curvefAmed

1. for hydrostatic or leak testing; (2) curves B :.J 7 for heat.up by non-nuclear i

means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C an* f for operations with a critical core other than low power i

PHYSICS TESTS, with:

}3 a.

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

A maximum cooldown of 100 F in any one hour period, j

i c.

A maximum temperature change of less than or equal to 20 F in any i

one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange temperature greater than or equal to 80 F when reactor vessel head bolting studs are under tension.

,_.s APPLICABILITY:

At all times. -$

l ACTION:

With 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-limit condition 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 SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l, SURVEILLANCE REQUIREMENTS i

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 j;

the right of the limit lines of Figure 3.4.6.1-lgurvey A r.edadf' or B ar.4-O',

y!

as applicable, at least once per 30 minutes.

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i, LA SALLE - UNIT 1 3/4 4-16 7907050216 890621 NDR ADDCK 0500 3

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIRE.ENTS (Continued) 4.4.6.1.2 -The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figures 3.4.6.1-1 Ago i

curves C andae* within 15 minutes prior to the withdrawal of control rods to~

3 s4 4*I~I l,

1 bring.the reactor to criticality.

I l 1l

4.. 6.1. 3 The reactor vessel material specimens shall be removed and examined 4

to' determine reactor pressure vessel fluence as a function of time and THERMAL POWER as required by 10 CFR Part 50, Appendix H in accordance with the schedule l

l:

in Table 4.4.6.1.3-1.

The results of these fluence determinations shall be used

.to update the curves of. Figures 3.4.6.1-li A4D J.+.6.1-la l

l l

l' 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be l

verified to be greater than or equal to 80"F 1

l a.

In 0PERATIONAL CONDITION 4 when the reactor coolant temperature is:

j 1

1.

5 100"F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

5 85 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.

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- - - - ' -O

REACTOR COOLANT SYSTEM-BASES

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V 1

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components 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 specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress l' lits-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 j

the tensile stresses induced by the internal pressure.~ Therefore, a pressure-1 temperature. curve based on steady state conditions, i.e., no thermas stresses, represents a lower bound of all similar curves for finite heatup rates when j

the inner wall of the vessel is treated as the governing location.

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

The thermal gradients established during heatup produce

.l tensile stresses which are already present.

The thermal-induced stresses at v

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 cc. rolling location, each heatup rate of interest must be analyzed on an individual basis.

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

Reactor NDT.

operation and resultant fast neutron, E greater than 1 Mev,. irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, NDT.menat based upon the fluence, phr: A s content and copper content of the material in question, can be predicted using D m ri tions of Regulatory Guide 1.99, Revision,2* gsre G ~:/4.^.0 ; a d the.recommenda-

" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The pressure /

i temperature limit curve, Figure 3.4.6.1-1, incluues predicted adjustments for at the end of . '%m.e. sigwsu crescrsve twL Powse vrAns (E this shift in RT NDT l

WWM FjeHRE 3,+, g, g. f a Ncu4pg5 pgEpir.rs p ADJ45rMErJTF W PsIMOT. AT WE E#p oF um The actual shift in RT f the vessel material will be established pl.gegg NDT periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50, Appendix H, irradiated reactor vessel material 4

specimens installed near the inside wall of the reactor vessel in the core.

l area.

Since the neutron spectra at the material specimens and vessel inside j

radius are essentially identical, the irradiated specimens can be used with y

- i t

LA SALLE - UNIT 1 B 3/4 4-4

REACTOR COOLANT SYSTEM BASES i

PRESSURE / TEMPERATURE LIMITS (Continued) pp 3 A,, 9, j. I a f

confidence in predicting reactor vessel material transition temperature shif t.

l The operating limit curves of l igure 3.4.6.1-1 shall be adjusted, as required, i

'I on the basis of the specimen data and the recommendations of Regulatory Guide 1.99, Rev. g. 2, 5.t.b.l-la N

pp The pressure-temperature limit lines shown in Figure 3.4.6.1-1 for reactor l

3 criticality and for inservice leak and hydrostatic testing have been established using the requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and I,jdrostatic testing, General Electric " Transient

[,

Pressure Rise Affecting Fracture Toughness Requirement for Boiling Water il

!)

Reactors," NED0-21778-A, December 1978, and " Protection Against Non-Ductile Failure" of the ASME Boiler and Pressure Vessel Code,1971 Edition, including I

Summer 1972 Addenda.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES i

Double isolation valves are provided on each of the main steam lines to

q minimize the potential leakage paths from the containment in case of a line j

break.

Only one valve in each line is required to maintain the integrity of j;

the containment.

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

The maximum closure time has been selected to l

contain fission products and to ensure the core is not uncovered following i

line breaks.

1 3/4.4.8 STRUCTURAL INTEGRITY Il ll 1;

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

4 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 1974 Edition and Addenda through Summer 1975.

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 l

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).

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 temperature r 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.

u LA SALLE - UNIT 1 B 3/4 4-5

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0 10 20 30 40 Service Life (Years )

Calculated Fast Neutron Fluence (E>l Mev) at hT As a Function of ll Service Life at 90% of RATED THERMAL POWER and 90% Availability l]

Bases Figure B 3/4.4.6-1 i

LA SALLE - UNIT 1 B 3/4 4-7 DEL ETE l

1 m__.__._____._.________

4 REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS-REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ano 3.4. r,.1 I A Thereactorcoolantsystemtemperatureandpressurefshallbelimited 3.4.6.1 in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve 4 A aumk

1. for hydrostatic or leak testing; (2) curver B and4' for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C aoM for 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-bour. 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 limit curves, and d.

The reactor vessel flange and head flange temperature greater than or equal to 86*F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times. 4 ACTION:

With 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-limit condition 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 SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS i

I

)

4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall I

be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curves A and A' or B and B',

as applicable, at least once per 30 minutes.

(fooTworr)

DyRWc, sMuTDOWN con DLTiorJ5 fcR HYORSTATIC OA LEA K TEST'ING OR HEATup SY N ON Ntt c.L srA rt MEArJS, "THE AVERAG E cooLA4T TEMPERATtA R.E LIMIT of TAB.LE Ed fDR. COLD SHurpoWrJ AND hot SHHTDOWN MAY BE iWCRE:ASGD To 21Z DEGRE E S -F.

LA SALLE - UNIT 2 3/4 4-17 ea w. e.* e e

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curves C and=4f within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality.

4.4.6.1.3 The reactor vessel material specimens shall be removed and examined to determine reactor pressure vessel fluence as a function of time and THERMAL POWER as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1.

The results of these fluence determinations shall be used to update the curves of Figures 3.4.6.1-law cl 3 4.G.l-hx.

l 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be i

verified to be greater than or equal to 86*F:

j a.

In OPERATIONAL CONDITION 4 when the reactor coolant temperature is:

1.

5106 F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 2.

5 91*F, at least once per 30 minutes.

i b.

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

LA SALLE - UNIT 2 3/4 4-18 w.v m

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3. 4. 6.1 - I q LaSattc - (44rT. 2 3/4 4 -j$,

9 REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components 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 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 of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the l

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 I

an individual basis.

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

Reactor i

NDT.

operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, NDT.

based upon the fluence, Prs $$Ucontent and copper content of the material in question, can be predicted using Bases Fi tionsofRegulatoryGuide1.99, Revision //gureB3/4.4.6-1andtherecommenda-

" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The pressure /

temperature limit curve, Figure 3.4.6.1-1, includes predicted adjustments for this shift in RT at the end of

f r"- -- t, S, = d 0- S I!;r:. 3. 'l. C.1Ne eff;;ti.e N-- i f Mt f=1 cycle Orb rd crees.

1. , O', e d C' 4" rigurc 3.4.0.1-1 :re effective fer-10 cffect4= full powef_.

-y::r (EFPYt. SirTEc4 cffccTwr fuu. Powee YMe5 (, EFpY) WHIL.E FsG uRE 3 A,(,,j M m u uper pRepic n o ADJusruturs IM RTus.T Ar rHE Ego urr n.g e g es, The actual shift in RT f the vessel material will be established NDT periodically during operation by removing and. evaluating, in accordance with ASTM E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the material specimens and vessel inside radius are essentially identical, the irradiated specimens can be used with LA SALLE - UNIT 2 B 3/4 4-4

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) ep 3.4. (o. l-k confidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and the recommendations of Regulatory Guide 1.99,Rev./.2.

. Mo 3.4. G. l-la The pressure-temperature limit lines shown in Figure 3.4.6.1-1 for _ reactor 4

criticality and for inservice leak and hydrostatic' testing have been established using the requirements of Appendix G to 10 CFR Part 50 for reactor criticality.

and for inservice leak and hydrostatic testing, General Electric " Transient Pressure Rise Affecting Fracture Toughness Requirement for Boiling Water Reactors," NED0-21778-A, December 1978, and " Protection Against Non-Ductile Failure" of the ASME Boiler and Pressure Vessel Code,1971 Edition, including Summer 1972 Addenda.

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 1974 Edition and Addenda through' Summer 1975.

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).

3/4.4.9 RESIDUAL HEAT REMOVAL l

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.

LA SALLE - UNIT 2 B 3/4 4-5

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CALCULATED FAST NEUTRON FLUENCE (E>1 MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE AT 90% OF RATED THERMAL POWER AND 90% AVAILABILITY BASES FIGURE B 3/4.4.6-1 i

l l

DELETE LA SALLE - UNIT 2 B 3/4 4-7

u' o-

.c-l 6

l' ATIAC3(ENT C SIGNIFICANT HAZARDS CONSIDERATION

-Commonwealth Edison has. evaluated the proposed: Technical Specification

' amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2, in accordance with the proposed amendment, will ant 1)

Involve a significant increase:in the' probability or consequences of an, accident previously evaluated because:

The proposed change is administrative becauso it does not' change: the physical facility. The proposed change is.the result of a re-evaluation.

that will put more stringent limits on the pressure - temperature relationship at the station for operation of both Units. The revised temperature limit of 200*F still provides forfsubcooling of the reactor vessel coolant below the boiling point.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:

The proposed change will'not introduce any new concerns for safety at the station. Reactor coolant system failures are previously addressed in the UFSAR, as well as pressure and temperature effects on the reactor coolant a

system. The change to'the 200*F temperature limit does not imply'any new.

l or different accident but. recognizes the actual boiling' point of,the j

reactor coolant.

- 1 1

i 3)

Involve a significant reduction in the margin of safety becauses.

j The margin of safety,.if anything, will be increased because the new l

pressure - temperature curve limits will help ensure the continued l

Integrity of the Reactor Vessel over the life.of the plant.

. ]

Guidance has been provided in 51.44 FR 7744 (Reference I.C.2.e.li) for the application of standards to license change requests for determination of the existence of significant hazards considerations. This-document provides

'j examples of amendments which are and are not'11kely considered to involve significant hazards considerations.

___________mm__

ATLJ -

1 i

This proposed amendment does not involve a significant relaxation of l

the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions of operations. Therefore, based on the l

guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration.

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