ML20006B113

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Amends 71 & 55 to Licenses NPF-11 & NPF-18,respectively, Revising Tech Spec 3/4 4.6 Re pressure-temp Limits of RCS in Order to Comply W/Generic Ltr 88-11 & Reg Guide 1.99,Rev 2
ML20006B113
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 01/16/1990
From: Jocelyn Craig
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20006B114 List:
References
RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, NUDOCS 9001310440
Download: ML20006B113 (20)
v • d • e


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NUCLEAR RECULATORY COMMISSION q#o UNITED STATES

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COMMONb'EALTW. EDISON COMPAWY 00CKE7.WO. 50 373 LASALLE COUNTY-STAT 10k.. UNIT.1 AMEWDMENT.TO. FACILITY.0PERAT!WG LICENSE Amendment No. 71-License No. NPF-11 1.

The Nuclear Regulatory Connission (the Connission or the NRC) has found that:

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee) dated June 21, 1989 complies 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 Connission; C.

There is reasonable assurance: (1)thattheactivitiesauthorizedby this amendinent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this anendment 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 amendnent 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 enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:

(2) Technical. Specifications.and Environsental. Protection Plan The Technical Specifications contained in Appendix A, as revised through Anandment No. 71, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

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

9001310440 900116

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This amendment is effective upon date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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1 M

khnW.Craig, Director L

Project Directorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects

Enclosure:

Changes to the Technical Specifications Datt of Issuance: January 16, 1990 l

t ENCLOSURE 10. LICENSE ANENDMENT.W0. 71.

FACillTY.0PERATIWG.LICEWSE.NO.. WPF 11 00CKET.WO.450 373 Replace the following pages of the Appendix "A" Technical Specifications with l

the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT XIX XIX 3/4 4-16 3/4 4-16 3/4 4 17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 4-1Ba B3/4 4-4 B3/4 4-4 B3/4 4-5 B3/4 4-5 B3/4 4-7 t

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i INDEX LIST OF FIGURES FIGURE PAGE l

3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS.........................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na:B o038 10 H O) i 2

VOLUME / CONCENTRATION REQUIREMENTS..................

3/4 1-22 3.4.1.5-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED).............................

3/4 4'4c

3. 4. 6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE........................

3/4 4-18

3. 4. 6.1-la MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE....................................

3/4 4-18a 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..........

3/4 7-32 B 3/4 3-1 REACTOR VESSEL WATER LEVEL.........................

B 3/4 3-7 d

B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS...................

B 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS...............................

5-2 5.1.2-1 LOW POPULATION ZONE................................

5-3 6.1-1 DELETED............................................

6-11 6.1-2 DELETED............................................

6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION.....................

6-13 LA SALLE - UNIT 1 XIX Amendment No. 71 P

, REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4.6.1-1 and 3.4.6.1-la (1) curves A for hydrostatic or leak testing; (2) curves B for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C for operations with a critical core other than low power PHYSICS TESTS, with:

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

b.

A maximum cooldown of 100'F in any one hour period, A maximum temperature change of less than or equal to 20'F in any c.

one hour period during inservice hydrostatic end 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.

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 i

remains acceptable for continued operations or be in at.least HOT SHUTDOWN i

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

i SURVEILLANCE REQUIREMENTS 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 limit lines of Figures 3.4.6.1-1 and 3.4.6.1-la curves A or B, as applicable, at least once per 30 minutes.

I

  • During shutdown conditions for hydrostatic or leak testing or heatup by nonnuclear means the average coolant temperature limit of Table 2.1 for cold shutdown and hot shutdown may be increased to 212'F.

'LA SALLE - UNIT 1 3/4 4-16 Amendment No.71

, 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 Figures 3.4.6.1-1 and 3.4.6.1-la curves C 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 3.4.6.1-1 and 3.4.6.1-la.

l 4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80'F:

In OPERATIONAL CONDITION 4 when the reactor coolant temperature is:

a.

1.

5100*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.

1 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.

l 1

LA SALLE - UNIT 1 3/4 4-17 Amendment No.71'

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i LA SALLE - UNIT 1 3/4 4-18a Amendment No. 71 I

I

_ 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, 1

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 I

so that the maximum specified heatup and cooldown rates are consistent with L

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 t

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 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 initial RTNDT.

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 will cause an increase in the RTNDT.

Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using the recommendations of Regulatory Guide 1.99, Revision 2. " 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 HDT *t the end of sixteen effective full power years (EFPY) while Figure 3.4.6.1-la includes predicted adjustments in RTNDT at the end of life fluence.

e The actual shift in RT of the vessel material will be established NDT 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 Since the neutron spectra at the material specimens and vessel inside area.

radius are essentially identical, the irradiated specimens can be used with LA SALLE-UNIT 1 B 3/4 4-4 Amendment No. 71

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REACTOR COOLANT SYSTEM BASES I

PRESSURE / TEMPERATURE LIMITS (Continued) i confidence in predicting reactor vessel material transition temperature shift.

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

i The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and

+

3.4.6.1-la for reactor 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 Require-ment 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 l

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.

i 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 1

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 i

and that an alternate method of coolant mixing be in operation.

LA SALLE-UNIT 1 B 3/4 4-5 Amendment No. 71

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'g umTEo sTATEF NUCLEAR RECULATORY COMMISSION WASHING TON, D, C. 20006 t

COMMONWEALTW.E0lS0W.COMPAkY J

DOCKET.NO. 50 374 LASALLE. COUNTY. STAT 10NeUNIT.2 l

ANENDMENT.TO. FACILITY.0PERATING. LICENSE 1

Amendment No. 55 License No. NPF-18 1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found

(

that:

l A.

The application for amendment filed by the Commonwealth Edison Comparty i

(the licensee), dated June 21, 1989 complies with the standards and t

requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the I

provisions of the Act, and the regulations of the Commission;

[

C.

There is reasonable assurance: (i) that the activities authorized by l

this amendment can be conducted without endangering the health and i

safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's rsgulations set forth in 10 CFR Chapter 1; l

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 anendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

l 2.

Accordingly, the license is anended by changes to the Technical Specifica-tions as indicated in the enclosure tc this license amendment and paragraph i

2.C.(2) of the Facility Operating License No. NPF-18 is hereby amenced to y

read as follows.

l (2) Technical Specifications.and Environmental. Protection. Plan The Technical Specifications contained in Appendix A, as revised through Anendment No. 55, and the Environmental Protection Plan contained in i

Appendix B, are hereby incorporated in the license.

The licensee shall l

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

i l

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3.

This amendment is effective upon date of issuance.

i FOR THE NUCLEAR REGULATORY COMMISSION L

/ John W. Craig, Director-Project Directorate III-2 l

Division of R66ctor Projects - III, l

IV, V and Special Projects i

l

Enclosure:

1 Changes to the Technical l.

Specifications i

i Date of Issuance:

January 16, 1990 l

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ENCLOSURE.TO LICENSE AMENDMENT NO. 55

[AtiLITYOPERATING. LICENSE.NO.NPF-18 p0CKET.NO.50374 Replace the following pages of the Appendix *A" Technical Specifications with the enclosed pages. The revised pages are identified by amendrent nunter and contain a vertical line indicating the area of change.

REMOVE INSERT XIX XIX 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-18 3/4 4-19 3/4 4-19 3/4 4 19a B3/4 4-4 B3/4 4-4 B3/4 4-5 B3/4 4-5 B3/4 4-7 i

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. LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

i CONCENTRATION REQUIREMENTS........................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na 0 0

2 10 16 10 H O) 2 VOLUME / CONCENTRATION REQUIREMENTS.................

3/4 1-22 3.4.1.5-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE

[

FLOW (% OF RATED)..................................

3/4 4 Sc 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE t

VS. REACTOR VESSEL PRESSURE.......................

3/4 4 19 i

3.4.6.1-la MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE............................

3/4 4-19a 4.7-1 SAMPLING PLAN FOR $NUBBER FUNCTIONAL TEST..........

3/4 7-33 i

B 3/4 3-1 REACTOR VESSEL WATER LEVEL........................

B 3/4 3-7 B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS..................

8 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS..............................

5-2 5.1.2-1 L OW PO P U L AT I ON Z ON E...............................5-3 6.1-1 DELETED...........................................

6-11 6.1-2 DELETED...........................................

6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION.....................

6-13 i

t i

t LA SALLE - UNIT 2 XIX Amendment No.55

REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS

_ REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the 1.imit lines shown on Figures 3.4.6.1-1 and 3.4.6.1-la;.

4 (1) curves A for h non-nuclear means,ydrostatic or leak testing; (2) curves B for heatup by.

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

a.

A maximum he tup of 100'F in any one hour period, b.

A maximum cooldown of 100'F in any one hour period, A maximum temperature change of less than or equal to 20'F in any c.

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.

APPLICABILITY:

At all t imes.*

j 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 conc tion on the structural integrity of the reactor couiant 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 />.

SURVEILLANCF. REQUIREMENTS l

4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing cperations, the reactor coolant system-temperature and pressure shall be determined to be within the above required heatup and cooldown limits'and to i

i the right of the limit lines of Figures.3.4.6.1-1 an.1 3.4.6.1-la curves A.or B, as applicable, at least once per 30 minutes.

1

  • During shutdown conditions for hydrostatic or leak testing or heatup by nonnuclear means, the average coolant temperature limit of Table 2.1 for cold shutdown and hot shutdown may be increased to 212 F.

LA SALLE - UNIT 2 3/4 4-17 Amendment No.55

' 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 Figures 3.4.6.1-1 and 3.4.6.1-la curves C 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-3.4.6.1-1'and 3.4.6.1-la.,

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4. 4. 6.1. 4 The reactor vessel flange and head flange temperature shall'be verified to be greater than or equal to 80'F:

In OPERATIONAL CONDITION 4 when the react'or coolant. temperature is:

a.

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.

1 85 F, at least once pe.- 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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REACTOR COOLANT SYSTEM J

BASES 1

3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand I

the effects of cyclic loads due to system temperature and pressure changes.<

These cyclic loads are introduced by normal load transients, reactor trips, i

3 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 j

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 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, I

Subsequently, for the cases in which the outer wall of the vessel becomes the t

stress controlling location, each heatup rate _of interest must be analyzed on an individual basis, l

i 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.

based upon the fluence, nickel content and copper content of the material in question, can be predicted using the recommendations of Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The pressure / temperature limit curve, q

Figure 3.4.6.1-1, includes predicted adjustments forlthis shift in RTNDT at the end of sixteen effective full power years (EFPY) while Figure 3.4.6.1-la includes predicted adjustments in RT at the end of life fluence.

1 NDT 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 Since the neutron spectra at the material specimens and vessel inside area.

radius are essentially identical, the irradiated specimens can be used with LA SALLE - UNIT 2 B 3/4 4-4 Amendment No.55

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. REACTOR COOLANT SYSTEM q

BASES

.1 PRESSURE / TEMPERATURE LIMITS (Continued) i confidence in predicting reactor vessel material transition temperature shift.

l The operating limit curves of Figures 3.4.6.1-1 and 3.4.6.1-la shall be adjusted,-

as required, on the basis of the specimen data and the recommendations of Regulatory Guide 1.99, Rev. 2.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and 3.4.6.1-la for reactor 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 Toughnefis Require-ment for Boiling Water Reactors," NE00-21778-A, December 1978, and " Protection Against Non-Ductile Failure of the ASME Boiler and Pressure Vessel Code,197.1-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.

i Components of the reactor coolant system were designed to provide access l

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).

I 3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal l

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 Amendment No.55 r

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