ML20059M447

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Proposed Tech Specs Re pressure-temp Limits,Per Generic Ltr 88-11
ML20059M447
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 09/20/1990
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20059M446 List:
References
GL-88-11, NUDOCS 9010040184
Download: ML20059M447 (8)
v • d • e


Text

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I PROPOSED TECilNICAL SPECIFICATION CilANGES 9010040184 900920 ADOCK 0500{<g3 Page 6 DR

3.1.2 pressurization. Heatuo. and Cooldown Limitations Snecification 3.1.2.1 Hydro Tests For_ thermal steady state system hydro tests, the system may be pressurized to the limits set forth in Specification 2.2 when there art fuel assemblies in the core, under the provisions of 3.1.2.3, and to ASME Code limits when no fuel. assemblies are present provided the reactor coolant system limits are to the right of and below the limit line in Figure 3.1.2-1.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.1.2.2 Leak Testa Leak tests required by Specification 4.3 shall be conducted under the provision of 3.1.2.3.

The provisions of Specification 3.0.3 are not applicable.

3.1.2.3 The reactor coolant pressure a..d the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with FJgure 3.1.2-2 and FJgure 3.1.2-3, and are as follows:

Heatup:

Allowabic combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1.2-2.

The heatup rates shall not exceed those shown in Figure 3.1.2-2.

Cooldown:

Allowable combinations of pressure and temperature for a specific q

cooldown shall be to the right of and holow the limit line in Figure 4

3.1.2-3.

Cooldown rates shall not exceed those shown in Figure 3.1.2-3.

3.1.2.4 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 100F.

3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 100F/hr.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 430F.

3.1.2.6 With the limits of Specifications 3.1.2.3 or 3.1.2.4 or 3.1.2.5 exceeded, restore the temperature and/or pressure to within the 11mit within 30 minutes; perform an engineering evaluation to determine the offects of the out-of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the-Reactor Coolant System remains acceptable for continued operations or he in at least il0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS Tavg to less than 200F within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Amendment No. 22, 25, 57 18

- I

3.1.2.7 Prior to reaching fifteen effective full power years of operation, Figures 3.1.2-1, 3.1 2-2 and 3.1.2-3 shall be updated for the next service period in accordance with'10CFR50, Appendix G, Section V.B.

The service period shall be qf sufficient duration to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with the latest revision of Topical Report BAW-1543(5).

The highest predicted adjusted reference temperature of all the beltline region materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.

The provisions of i

Specifications 3.0.3 and 3.0.4 are not applicable.

3.1.2.8 The updated proposed technical specifications referred to in 3.1.2.7 shall be submitted for NRC review at 1 cast 90 days prior to the end of the service period.

Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR Part 50, Appendix G.Section V.C.

3.,

2.9 With the exception of ASME Section XI testing and when the core flood tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the i

motor operators opened before depressurizing the reactor coolant system below 600 psig.

3.1.2.10 With the exception of ASME Section XI testing, fill and vent of the reactor coolant system, and to allow maintenance of the valves, when the reactor coolant temperature is less than 300'F the four liigh Pressure Injection motor operated valves shall be closed with their opening control circuits for the motor operators disabled.

3.1.2.11 The plant shall not be operated in a water solid condition when the i

RCS pressure boundary is intact except as allowed by Emergency Operating Procedures and during System flydrotest.

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Amendment No. 57, M, 75 18a ARKANSAS - UNIT 1 r

RASES All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperaturo and pressure changes.(') These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR.

The maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation.(8) The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100F satisfies stress levels for temperatures below the DTT.(*)

The major components of the reactor coolant prosaure boundary have been analyzed in accordance with Appendix G to 10CFR50.

Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-2106(6).

The limiting wold material is being irradiated as part of the B6W Owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the capsules containing the limiting wold material is discussed in the latest revision to B&W report, BAW-1543. (s) The chemical composition of the limiting weld material in reported in the B&W Report, BAW-1511p.(') The offect of neutron irradiation on the RT f the limiting weld material is reported in the B&W Report., BAW 2075(iT Figures 3.1.2-1, 3.1.2-2, and 3.1.2., present the pressuro-temperature limit curves for hydrostatic test, normal heatup, and normal cooldown respectively.

The limit curves are applicable through the fifteenth effectivo full power year of operation. The pressure limit is adjusted for the pressure differential between the point. of system pressure measurement and the If miting component for all operating reactor coolant pump combinations.

The pressure-temperature limit lines shown on Figure 3.1.2-2 for reactor critienlity and on Figure 3.1.2-1 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.

The actual shif t. In RT of the bolt 11no region material will be established periodically during op Ntion by removing and evaluating, in accordance with Appendix 11 to 10CFR50, reactor vossol material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region.

The spray temperature difference restriction based on a stress analysis of the spray lino nozzlo is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit. Temperature requirements for the stons generator correspond with the measured NDTT for the shell.

Amendment No. 22, /f, H M 19

The heatup and cooldown rates stated in this specification are intended as the maximum changes in temperature in one direction in a one hour period. The actual temperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total temperature dif ference from exceeding the limit for the one hour period.

Specification 3.1.2.9 is to ensure that the core flood tanks are not the source for pressurizing the reactor coolant nystem when in cold shutdown.

Specification 3.1.2.10 is to ensure that high pressure injection is not the source of pressurizing the reactor co31 ant system when in cold shutdown.

Specification 3.1.2.11 is to ensure that the reactor coolant system is not operated in a manner which would allow overpressurization due to a temperature transient.

REFERENCES (1) FSAR, Section 4.1.2.4 (2) ASME Boller and pressure Code, Section Ill, N-415 (3) FSAR, Section 4.3.11.5 I

i (4) BAW-2106 (5) BAW-1543, latest revision 4

l (6) BAW-1511p (7) BAW-2075, Revision 1 i

i Amendment No. 2, 12, 15, 53, 75 20 ARKANSAS - UNIT 1

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- NOTES:

7 2400

1.THE ACCEPTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE I

a RIGHT OF THE LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE LIMITIPG PRESSURE Z

DIFFERENTIAL BETWEEN THE POINT OF SYSTEM PRESOURE MEASUREMENT AND THE P

PRESSURE ON THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE. THEY m

2200

?

ARE NQI ADJUSTED FOR POSSIBLE INSTRUMtiNT ERROR.

8 2000

2. APPLICABLE FOR HEATUP RATES OF s 50 DEG F/HR.

4 g

- 3. APPLICABLE FOR COOLDOWN RATES OF:

5 1800 T2280 DEG F 100 DEG F/HR (550 DEG F IN ANY 1/2 HOUR PERIOD) 3 280 DEG F > T2150 DEG F 50 DEG F/HR (s25 DEG F IN ANV 1/2 HOUR PERIOD) j g

T<150 DEG F 25 DEG F/HR (s25 DEG F11; ANY 1 HOUR PERIOD)

E - 1600 2

E&

- 4. REACTOR COOLANT PUMP RESTRICTIONS:

2*

E. 7 7

NONE.

I us s 1400 Eo SI m

H

$ 15 o~

j PRESS TEMP 3

o W

NR 3

800 A

388 70 5

B 422 100 E

C 455 125 600 D

461 150 C

D d

E 527 170 f

F 3

F 541 205 400 A

UC' G

822 220 H

1115 255 I

1382 280 200 J

1763 305 K

2286 330 L

2500 339 A

i i i l i i I i l i i i l I I i i l i i I i I i I i i i i E

O 50 100 150 200 250 300 350 Indicated Reactor Coolant Intet Temperature, 'F REACTOR COOLANT SYSTEM INSERVICE HYDROSTATIC TEST HEATUP AND COOLDOWN a

LIMITATIONS APPLICABLE FOR FIRST 15.0 EFFECTIVE FULL POWER YEARS

y, 2600 3

C a

M 28 NOTES:

1.THE ACCEPTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE u

o l

E RIGHT OF THE LIMIT CURVE (S). THE LIMIT CURVES ;NCLUDE THE LIMITING PRESSURE f

7

-g DIFFERENTIAL BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE

/

8 PRESSURE ON THE REACTOR VESSEL REClON CONTROLLING THE LIMIT CURVE. THEY f

ARE NQT AD.*USTED FOR POSSIBLE INSTRUMENT ERROR.

j

)

s 2

2. APPLICABLE FOR HEATUP RATES OF 5 50 DEG F/liR.

N

_- 3.WHEN THE DECAY HEAT REMOVAL SYSTEM IS OPERATING W!TH NO RC L

p I

PUMPS OPERATING. THE INDICATED DHR SYSTEM RETURN TEMPERATURE En- - 1600 TO THE REACTOR VESSEL SHALL BE USED.

4 2

4. REACTGR COOLANT PUMP RESTRICTIONS:

/

.F NONE.

/

{

E y

5. CURVE N-O-P-O IS THE CRTTICALITY CURVE.

o (/

g PRESS TEMP o

I 8

g O 1200 i

N (psig)

(deg F) a

$sc lii A

331 70 f

1000 j

o" B

331 125 C

369 150 D

415 170

}

I E

498 195 3

800 F

541 204 g

H G

541 230 H

695 234 600 8

809 255 E

d J

1011 280 D

F G

K 1234 300 400 C

L 1708 330 A

5 F

M 2250 355 N

O 340 0

1234 340 P

1708 370 N

O 2250 395 0

3 I I I I I I I I I I I I I I I I I I I I I I I i i i I I I I i 1 y

0 50 100 150 200 250 300 350 400 Indicated Reactor Coolant intet Temperature, 'F e

h iu REACTOR COOUU4T SYSTEM NORMAL OPERATION - HEATUP LIMITATIONS

'u APPLICABLE FOR FIRST 15.0 EFFECTIVE FULL POWER YEARS

1 I

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9 3

NOTES:

1.THE ACCEPTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE 2400 R!GHT OFTHE LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE LIMITING PRESSURE g

DIFFERENTIAL BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE g

9 2200 PRESSURE ON THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE. THEY 7

7 y

ARE NQI ADJUSTED FOR POSSIBLE INSTRUMENT ERROR.

8 2000

2. APPLICABLE FOR COOLDOWN RATES OF:

T2280 DEG F 100 DEG F/HR (s50 DEG F IN ANY 1/2 HOUR PERICO) /

.?

E.

280 DEG F > T2150 DEG F 50 DEG F/HR (s25 DEG F IN ANY 1/2 HOUR PERIOO) /

T<150 DEG F 25 DEG F/HR (s25 DEG FIN ANY 1 HOUR PERIOD)/

5 1800 I

3. WHEN THE DECAY HEAT REMOVAL SYSTEM IS OPERATING WITH NO RC g - 1600 PUMPS OPERATING. THE INDICATED DHD SYSTEM RETURN TEMPERATURE f

g&

TO THE EEACTOR VESSEL SHALL BE USED.

)

s >-

4. A MAXIMUM STEPTEMPERATURE CHANGE OF 35 DEG F IS ALLOWABLE

/

1400 WHEN REMOVING ALL RC PUMPS FROM OPERATION WITH THE DHR /

5E SYSTEM OPERATING. THE STEP TEMPERATURE CHANGE IS

/

3 m

DEFINED AS THE RC TEMPERATURE (PR!OR TO STOPPING ALL /

8 8 1200 O

RC PUMPS) MINUS THE DHR RETURN TEMPERATURE TO cc

$ ~li; THE REACTOR VESSEL (AFTER STOPPING ALL RC PUMPS).

I THE 50 DEG F/HR COOLDOWN RATE IS ALLOWABLE M

1000 BOTH BEFORE AND AFTER THE STEP TEMPERATURE /

'/

E PRESS TEMP CHANGE.

[

PO!NT (psg)

(deg F) g

/

_=

800 A

260 70 y

5. REACTOR COOLANT PUMP RESTRICTIONS: /

B 287 100 5

NONE.

f C

312 125 600 D

316 150 j"

F /

E 366 170 E

f F

430 195 400 G

589 220 f

C B

H 907 255 1

1192 280 200 J

2078 330 K

2250 339 I

0 mi 0

50 100 150 290 250 300 350 400 O

indicated Reactor Cootent inlet Temperature, 'F w

.o ip REACTOR COOLANT SYSTEM NORMAL OPERATION - COOLDOWN LIINTATIONS APPUCABLE FOR FIRST 15.0 EFFECTWE FULL POWER YEARS

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