ML20059M448

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Unit 1 Pressure-Temp Limits for 15 Efpy
ML20059M448
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/31/1990
From:
BABCOCK & WILCOX CO.
To:
Shared Package
ML20059M446 List:
References
BAW-2106, NUDOCS 9010040186
Download: ML20059M448 (18)
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BAW-2106 MARCH 9990 I

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B ARKANSAS NUCLEAR ONE - UNIT 1

I PRESSURE-TEMPERATURE LIMITS I

I FOR 15 EPPY I

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a HAW-2106 March 1990

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I Arkansas Nuclear One - Unit 1 Pressure Temperature Limits for 15 EFPY I

l Prepared for Arkansas Power and Light Company l

I by B&W Nuclear Service Company Engineering and Plant Services Division P. O. Box 10935 i Lynchburg, Virginia 24506 0935 l ,

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I Prepared by: .\ 3 9 96.

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l Reviewed by: .

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Approved by: [

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CONTENTS Page

1. INTRODUCTION ............................ 1
2. DEVELOPMENT OF TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS . 2
3. RESVLTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
4. REFERENCES ............................. 6 APPENDIX .

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A. Revised P/T Technical Specification and Bases . . . . . . . . . . . . A-1

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1. INTRODUCTION In July 1988 the NRC released Generic letter 88-lll to all Licensees of Operating L

Reactors, the subject being, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its impact of Plant Operations." The purpose of Generic Letter 8811 was to inform utilities operating nuclear power plants of the release of Revision 2 to Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," which became effective in May 1988.2 In addition, the I NRC stated that this Regulatory Guide will be used by the NRC in reviewing

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submittals regarding pressure temperature (P/T) limits and analyses other than pressurized thermal shock (PTS) that require as estimate of the embrittlement of reactor vessel beltline materials.

This report presents the Arkansas Nuclear One - Unit I reactor coolant Technical Specification pressure temperature operating limits for 15 EFPY, in response to -

l Generic letter 8811. Limits for normal operation, both heatup and cooldown, inservice leak and hydrostatic tests, and reactor core operation are provided.

The data used to develop these limitations are based on the analysis of the _

fourth capsule (AN1-C) of the Arkansas Power and Light Company Arkansas Nuclear One, Unit I reactor vessel surveillance program as reported in BAW 2075.3

!" Pddit ica, the revised Technical Specification and Bases are contained in Appendix A.

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2. DEVELOPMENT OF TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS

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. The pressure temperature limits of the reactor coolant pressure boundary (RCPB) of Arkansas Nuclear One, Unit I are established in accordance with the f requirements of 10CFR50, Appendix G.4 The' methods and criteria employed to establish operating pressure and temperature limits are described in topical I report BAW-}0046A.5 The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated op.: cation I occurrences and system hydrostatic tests. The loading conditions of interest include the following:

1. Normal operations, including heatup and cooldown.
2. Inservice leak and hydrostatic tests.

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3. Reactor core operation, The major components of the RCPB have been analyzed in accordance with 10CFR50, f

Appendix G. The closure head region, the reactor vessel outlet n)zzle, and the beltline region have been identified as the only regions of the ieactor vessel (and consequently of RCPB) that require the pressure-temperature limits. Since the closure head region is significantly ' stressed at relatively low temperatures (due' to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods.

4 The reactor vessel outlet nozzle also affects the pressure temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle. After several years of neutron radiation exposure, the RTNDT of the vessel beltline region materials will l control the pressure-temperature limits of the RCPB.

The limit curves for Arkansas Nuclear One, Unit 1 are based on the predicted value of the adjusted reference temperature of the limiting beltline region material at the end of the fifteenth EFPY. The adjusted reference temperature is calculated by adding the predicted radiation induced RTNDT to the initial RTNDT. The predicted RTNDT is calculated using the vessel's neutron fluence and chemistry. Table 1 summarizes the predicted reactor vessel inside surface peak-

fluence value at 15 EFPY for Arkansas Nuclear One, Unit 1. Regulatory Guide 1.99, Rev. 2,2 was used to predict the radiation induced RTNDT value as a function of the material's copper and nickel content and neutron fluence. Using r the 15 EFPY fluence value and the vessel's chemistry, the adjusted RTNDT values I

of the beltline region at the end of the fifteenth full power year are obtained and provided in Table 2. The adjusted RTNDT values are arovided for the 1/4T and 3/4T vessel wall locations (T = wall thickness). The assumed RTNDT of the closure head region and the outlet nozzle steel forgings is 60F, in accordance with BAW 10046A.5 ,

Using the methodology documented in BAW 10046A,5 pressure-temperature limits for the closure head region, the outlet nozzle, and the beltline region were determined for the heatup and cooldown rates summarized in Tables 3 and 4.

Differential pressure corrections were then applied to the unadjusted P/T limits to account for the pressure differential between the analyzed regions of the reactor vessel and the system pressure sensor in the reactor coolant system.

The maximum allowable pressure as a function of fluid temperature is obtained through a point by-point comparison of the three limiting regions, adjusted for sensor location. The maximum allowable pressure is taken to be the lowest of the three calculated pressures. The resulting adjusted data points determine the bounding P/T Technical Specification curves. Instrumentation errors for pressure and temperature were not applied to the Technical Specification limits provided in Appendix A.

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Table 1. Arkansas Nuclear One, Unit 1 15 EFPY RV um inside Surface Fluence (Max location) 0.488 E19 n/cm2 (3)

L Table 2. Arkansas Nuclear One, Unit 1 15 EFPY RTNDT(S)

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1/4T (OF ) 1 3/4T (%) -

( 183 139 l

Table 3. Ooerational Constraints for Plant Heatuo

(  !. RC Temperature Constraints: Maximum Heatup Rate s 50F/hr

11. RC Pump Constraints: None m

Table 4. Ooerational Constraints for Plant Cooldown

1. RC Temperature Constraints: Maximum Cooldown Rate _

0 T 2 280 F s 50% in any 1/2 hour period 0

150 F s T < 280% 125 0F in any 1/2 hour period T < 150 0F s 25% in any I hour period ,

in addition, a maximum step temperature change of 350F is allowable when removing all RC pumps from operation with the Decay Heat Removal System operating. -

II. RC Pump Constraints: None i

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}. RESULTS

- The P/T Technical Specification limits for Arkansas Nuclear One, Unit 1 through 15 EFPY for normal heatup, normal cooldown, inservice leak and hydrotests, and I reactor core operation, as required by 10CFR50, Appendix G are provided in Appendix A. The limits have invoked additional operational constraints, l identified in Tables 3 and 4, that must be maintained for future operation.

l Protection against nonductile faDure is ensured by maintaining the reactor coolant system pressure below the upper limits of the pressure-temperature limit j c'trve s . The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not g permitted to be critical until the pressure temperature combinations are to the B right of the criticality limit curve.

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4. REFERENCES s l. NRC Gent ic letter 8811, *NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," dated July 1,1988.
2. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel

( Material, Reaulatory Guide 1.99. Revision 2, May 1988.

3. BAW-2075. Revision 1, " Analysis of Capsule AN1-C Arkansas Power and Ligt.t Company Arkansas Nuclear One, Unit 1,* October 1989.
4. Code of Federal Regulation, Title 10, Part 50 Domestic Licensing of-Production and Utilization Facilities, Appendix G. Fracture Toughness Requirements, f
5. H. W. Behnke, et. al., Methods of Compliance with Fracture Toughness and Operational Requirements of Appendix G to 10CFR50, BAW-10046A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, June 1986.

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APPENDIX A Revised P/T Technical Specification and Bases A-1

3.1.2 Pressurif ation. Heatuo. and Cooldown limitations Soecification 3.1.2.1 Hydro Tests

  • For thermal steady state system hydro tests, the system may be g pressurized to the limits set forth in Specification 2.2 when there L are 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 Tests 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 and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with figure 3.1.2-2 and Figure 3.1.2-3, and are as follows:

Heatup:

Allowable 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 cooldown shall be to the right of and below the limit line in Figure 3.1.2-3. Cooldown rates shall not exceed those shown in Figure 3.1.2-3.

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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 limit within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor i Coolant System remains acceptable for continued operations or be in at least HOT 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 RCS Tavg to 1ess 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 />.

l 3.1.2.7 Prior to reaching fifteen effective full power years of operation. -

l Figures 3.1.21, 3.1.2-2 and 3.1.2-3 shall be updated for the next i

service period in accordance with 10CFR50, Appendix G.Section V.B. _

The service period shall be of sufficient duration to permit the I

scheduled evaluation of a portion of the surveillance data scheduled i in accordance with BAW-1543, latest revision. The highest predicted -

i 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 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 least 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 10CFR50, Appendix G, Section V.C.

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< 3.1.2.9 With the exception of ASME Section XI testing and when the core flood l

t tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the 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 300F the four High 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 RCS pressure boundary is intact txcept as allowed by Emergency Operating Procedures and during Sys.em Hydrotest.

ILA315 All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.I 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.2 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.3 The major components of the reactor coolant pressure 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-2106.4 The limiting weld material is being irradiated as part of the B&W Owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the capsules containing the limiting weld material is discussed in the latest revision to B&W report, BAW-A-4

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1543.5 The chemical composition of the limiting weld material is reported in B&W Report, BAW 1511P.6 The effect of neutron irradiation on the RTNDT of the limiting weld material is reported in the B&W Report, BAW 2075.7

[ Figures 3.1.2-1, 3.': 2 2, and 3.1.2-3 present the pressure temperature limit curves for hydrostatic test, normal heatup, and normal cooldown respectively.

The limit curves are applicable through th+ 4 Siteenth effective full power year

{ of operation. The pressure limit is adjusted for the pressure dif ferential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The pressure temperature limit lines shown on Figure 3.1.2 2 for reactor criticality 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 shift in RTNDT of the beltline region material will be established

[ periodically during operation by removing and evaluating, in' occordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in

1e core region.

Thespraytemperaturedifferencerestrictionbasedonaspssanalysisofthe spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

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

Specification 3.1.2.10 is to ensure that high pressure injection is not the source of pressurizing the reactor coolaat 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

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REFERENCES

1. FSAR, Section 4.1.2.4
2. ASME Boijer'and Pressure Code,Section III, N 415
3. FSAR, Section 4.3.11.5
4. BAW 2106

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5. BAW 1543, Latest Revision'

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7. BAW 2075, Revision 1

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ym ammer e r1 > w ARKANSAS NUCLEAR ONE UNIT 1

APPLICABLE FOR THE FIRST 15.0 EFFECTIVE FULL POWER YEARS L l NOTES: l

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1. THE ACCEPTABLE PRESSURE / TEMPERATURE COMBINATIONS ARE BELOW AND TO THE RIGHT OF THE LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE PRESSURE K 2200 - D"FERENTIAL BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE i ESSURE ON THE REACTOR VESSEL REGION CONTRCitiNG THE LIMIT CURVE. THEY ARE NOT ADJUSTED FOR POSSIBLE INSiRUMENT ERROR.

2000 -

_ee 2. APPLICABLE FOR HEATUP RATES OF,5 50 DEGF/HR.

E g 1800 -

J B 3. APPLICABLE FOR COOLDOWN RATES OF:

$ T)280 DEGF 100 DEGF/HR

$ 1600 -

280 DEGF > T >,150 DEGF

(< 50 DEGFIN ANY % HOUR PERIOD) 50 DEGF/HR E 'E 34 (C 25 DEGF IN ANY % HOUR PERIOD)

T < 150 DEGF 25 DEGF/HR

> hI 1400 -

I L py (C 25 DEGFIN ANY I HOUR PERIOD) 2c

[$

B$

1200 -

4. REACTOR COOLANT PUMP RESTRICTIONS: NONE H

US POINT TEMP PRESS E 1000 (deEF) (psig)

% G g A 70 388 800 B 100 422 g

_c C 125 455 D 150 461 600 -

E E 170 527 B

C Q -

F F

G 205 2M 541 822 400 -

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255 280 1115 1382 J 305 1763 200 -

K 330 2286 L 339 2500 I I I I I I 0 8 50 100 150 200 250 300 350 Indicated Reactor Coolant inlet Temperature. 'F l

.\ 1 L-_) \ l l 1 L___) U M M M M M' ^'M ~M M M M M' 'M ' M F9gure 3.1.2-2 ARKANSAS NUCLEAR ONE UNIT 1 l REACTOR COOLANT SYSTEM, NORMAL OPERATION - HEATUP LIMITATIONS I 2600 -

APPLICABLE FOR THE FIRST 15.0 EFFECTIVE FULL POWER YEARS l 2400 -

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NOTES:

i 2000 -

1. THE ACCEPTABLE PRESSURE / TEMPERATURE COMBINATIONS ARE BELOW AND TO THE j . RIGHT OF THE LIMIT CURVE (S). THE LilAIT CURVES INCLUDE THE PRESSURE '

j l DIFFERENTIAL BETWEEN THE POINT OF SYS1F.M PRESSURE MEASUREMENT AND THE g 1800 -

PRESSURE ON THE REACTOR ' 3SEL REGION CONTROLLING THE LIMIT CURVE.THEY  !

g ARE NOT ADJUSTED FOR POSSIBLE INSTRUMENT ERROR. p e t i 1600 -

g 2. APPLICABLE FOR HEATUP RATES OF,6 50 DEGF/HR.

20 i p $If 1400 -

3. WHEN THE DECAY HEAT REMOVAL SYSTEM IS OPERATING WITH NO RC PUMPS

& Ed OPERATING. THE INDICATED DHR SYSTEM RETURN TEMPERATURE TO THE POINT TEMP PRESS S REACTOR VESSEL SHALL BE USED.

[g$ 1200 -

K O< IN A 70 331 i

$h g- L REACTOR COOLANT PUMP RESTRUCTIONS: NONE. B 125 331 1000 - C 150 369

& J D 170 415 i' i r E 195 498

$ 800 F 204 541 g

l G 230 541

.5 ' H 234 695 H

600 I HEATUP LIMIT I F G K 300 1234 l E L 330 1708 400 -

D M 355 2250

g C N 340 0 CRITICALITY LIMIT j 0 340 1234 M -

P 370 1708  !

Q 395 2250 l 1 I I I I N g j 0 50 100 150 200 250 300 350 400 Indicated Reactor Coolant inletTemperature. 'F

4 M M M M . m ' Iiiil M Wigu E .1. M M M E E E E " E ARKANSAS NUCLEAR ONE UNIT 1 REACTOR COOLANT SYSTEM, NORMAL OPERATICM - COOLDOWN LIMITATIONS l 2600 - APPLICABLE FOR THE FIRST 15.0 EFFECTIVE FULL POWER YEARS NOTES:

2400 -

1.THE ACCEPTABLE PRESSURE / TEMPERATURE COMBINATIONS ARE BELOW AND TO THE RIGHT OF THE. LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE PRESSURE K '

DIFFERENTIAL BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE 2200 -

PRESSURE ON THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE. THEY ARE NOT ADJUSTED FOR POSStBLE INSTRUMENT ERROR.

J 2000 -

2. APPLICABLE FOR COOLDOWN RATES OF:

4 Y T > 28G DEGF 100 DEGF/HR ((SO DEGF IN ANY % HOUR FERIOD)

S 280 DEGF > T > 150 DEGF 50 DEGF/HR ((25 DEGF IN ANY % HOUR PERIOD) -)

g 1800 -

T< 150 DEGF 25 DEGF/HR (< 25 DEGF IN ANY 1 HOUR PERIOD)

= #

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E 3.WHEN THE DECAY HEAT REMOVAL SYSTEM IS OPERATING WITH NO RC IE s#

1600 -

PUMPS OPERATING, THE INDICATED DHR SYSTEM RETURN TEMPERATURE TO THE REACTOR VESSEL SHALL BE USED.

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4. A MAXIMU M STEPTEMPERATURE CHANGE OF 35 DEGF IS ALLOWABLE

, 5-8 WHEN dEMOVING ALL RC PUMPS FROM OPERATION WITH THE DHR h m1200 u

- SYSTEM OPERATING. THE STEP TEMPERATURE CHANGE IS ,

DEFINED AS THE RC TEMPERATURE (PRIOR TO STOPPING ALL

$3 RC PUMPS) MINUS THE DHR RETURN TEMPERATURE TO 2 1000 THE REACTOR VESSEL (AFTER STOPPING ALL RC PUMPS).

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" THE 50 DEGF/HR COOLDOWN RATE IS ALLOWABLE H s BOTH BEFORE AND AFTER THE STEPTEMPERATURE (degF) (psig)

.0 CHANGE. A 70 260 800 -

E B 100 287

5. REACTOR COf ' %NT PUMP RESTRICTIONS: C 125 .312 NONE. n. .G i t a L. i D 150 316 600 - .
o. E 170 366 F F 195' 430 400 -. E" G '220 589 B C D H 35 M7 A

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I 280- 1192 200 - J 330 2078 K 339 2250 ,

o I I I I I I I I O '50 100 150 200 250 300 350 400

' Indicated Reactor Coolant inlet Temperature *F

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