ML19312C014

Proposed Tech Specs 3.1 & 4.2 Incorporating Changes to Pressurization,Heatup & Cooldown Limitations & Reactor Vessel Matl Surveillance Program

ML19312C014
Person / Time
Site:	Oconee
Issue date:	10/01/1975
From:	DUKE POWER CO.
To:
Shared Package
ML19312C012	List: ML19312C011 ML19312C014
References
NUDOCS 7911270789
Download: ML19312C014 (13)

Text

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3.1.2 Pressurization, Heatup, and Cooldown Limitations Specification 3.1.2.1 The reactor coolant pressure and the system heatup and cooldown I rates (with the exception of the pressurizer) shall be limited as follows:

1 1

Heatup:

Heatup rates and allowable combinations of pressure and tempera-tures shall be limited in accordance with Figure 3.1.2-1A Unit 1 3.1.2-1B Units 2&3.

Cooldown:

Cooldown rates and allowable combinations of pressure and tempera-ture shall be limited in accordance with Figure 3.1.2-2A Unit 1 3.1.2-2B Units 2&3.

3.1.2.2 Leak Tests Leak tests required by Specification 4.3 shall be conducted under the provisions of 3.1.2.1.

3.1.2.3 Hydro Tests For thermal steady state system hydro test the system may be pressurized to the limits set forth in Specification 2.2 when there are fuel assemblies in the core under the provisions of 3.1.2.1 and to ASME Code Section III limits when no fuel assem-blies are present provided the reactor coolant system is to the right of and below the limit line in Figure 3.1.2-3.

3.1.2.4 The secondary side of the steam generator shall not be pressurized above 237 psig if the temperature of the vessel shell is below 1100F.

3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 1000F/hr.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 4100F, 3.1.2.6 Pressurization, heatup and cooldown limitations shall be updated based on the results of the reactor vessel materials surveillance program described in Specification 4.2.8.

3.1-3 Entire Page Revised

7911270 y

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Bases - Unit 1 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, startup and shutdown operations, and inservice leak and hydrostatic tests.

The various categories of load cycles used for design purposes are provided in Table 4.8 of the FSAR.

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 BAR-1421(7).

Figures 3.1.2-1A, 3.1.2-2A, and 3.1.2-3 present the' pressure-temperature limit curves for normal heatup, normal cooldown and hydrostatic test respectively. The limit curves are applicable up to the fif th effective full power year of operation. These curves are adjusted by 25 psi and 100F Lar possible errors in the pressure and temperature censing instruments.

The pressure limit is also adjusted for the pressure differential 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 3.1.2-1A for reactor criticality and on Figure 3.1.2-3 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 accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of the reactor vessel in the core region.

The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of the steam generator. At metal temperatures lower than the RTNDT of +600F, the protection agains,t nonductile failure if achieved by limiting the secondary coolant pressure to 20 percent of the preoperational system hydrostatic test pressure. The limitations of Il00F and 237 psig are based on the highest estimated RTNDT of +400F and the preoperational system hydrostatic test pressure of 1312 psig.

The average metal temperature is assumed to be equal to or greater than the coolant temperature. The limitations include margins of 25 psi and 100F for possible instrument error.

The spray temperature difference is imposed to maintain the thermal stresses

, at the pressurizer spray line nozzle below the design limit.

3.1-3a New Page

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Bases Units 2 and 3 All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. (1) These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown opc:ations. 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 rate of 1000F per hour satisfies stress limits for cyclic operation. (2) The 237 psig pressure limit for the secondary side of the steam generator at a temperature less than 1100F satisfies stress levels for temperatures below the DTT. (3) The reactor vessel plate material and welds have been tested to verify conformity to specified requirements and a maximum NDTT value of 200F has been determined based on Charpy V-Notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 400F.

Figures 3.1.2-1B and 3.1.2-2B contain the limiting reactor coolant system l pressure-temperature relationship for operation at DTT(4) and below to assure that stress levels are low enough to preclude brittle fracture.

These stress levels and their bases are defined in Section 4.3.3 of the FSAR.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation. The predicted maximum NDTT increase for the 40-year exposure is shown on Figure 4.10.(4)

The actual shift in NDTT will be determined periodically during plant opera-tion by testing of irradiated vessel material samples located in this. reactor vessel.(5) The results of the irradiated sample testing will be evaluated l and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperature.

The design value for fast neutron (E > 1 MeV) exposure of the reactor vessel is 3.0 x 1010 n/cm2 +- s at 2,568 MWt rated power and an integrated exposure of 3.0 x 1019 n/cm2 for 40 years nperation. (6) The calculated maximum values are 2.2 x 1010 n/cm2 -- s and 2.2 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load. (4) Figure 3.1.2-1B is based on the design value which is considerably higher than the calculated value.

The DTT value for Figure 3.1.2-1B is based on the projected NDTT at the end of the first two years of operation. During these two years, the energy output has been conservatively estimated to be 1.7 x 106 thermal megawatt days, which is equivalent to 655 days at 2,568 MWt core power. The projected fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 Jc n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value for fast neutron exposure.

The actual shift in NDTT will be established periodical'ly during~ plant operation by testing vessel material samples which'are irradiated cumulatively

, by securing them near the inside wall of the vessel in the core area. To compensate for the increases in the NDTT caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.

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The NDTT shift and the magnitudes of the thermal and pressure stresses are sensitive to integrated reactor power and not to instantaneous power level.

Figure 3.1.2-13 and 3.1.2-2B are applicable to reactor core thermal ratings up to 2,568 HWt.

The pressure limit line on Figure 3.1.3-1B has been selected such that the l reactor vessel stress resulting from internal pressure will not exceed 15 percent yield strength considering the following:

1. A 25 psi error is measured pressure.

2. System pressure is measured in either loop.

3. Mgximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump combinations.

For adequate conservatism in fracture toughness including size (thickness) affect, a maximum pressure of 550 psig below 2750F with a maximum heatup and cooldown rate of 500F/hr has been imposed for the initial two year period as shown on Figure 3.1.2-1B. During this two year period, a fracture toughness criterion applicable to Oconee Units 2 and 3 beyond this period will be developed by the AEC. It will be based on the evaluation of the fracture toughness properties of heavy section (thickness) steels, both irradiated and unirradiated, for the AEC-HSST program and the PVRC program, and with considerations surveillance programs. of test results of the Oconee Units 2 and 3 reactor. l The spray temperature difference restriction 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 NDT' for the shell.

REFERENCES (1) FSAR Section 4.1.2.4.

(2) ASME Boiler and Pressure Code,Section III, N-415.

• j (3) FSAR Section 4.3.10.5.

(4) FSAR Section 4.3.3.

(5) FSAR Section 4.4.6.

(6) FSAR Sections 4.1.2.8 and 4.3.3.

(7) Analysis of Capsule OCl-F from Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1421 Rev. 1, i September 1975. l 3.1-5

a 2400 ,

E F g j 2200 -

"o E 2000 -

THE ACCEPTABLE PRESSURE AND TEMPERATURE COMBINAll0NS ARE BELOW AND 10 1HE RIGHT OF THE LIMIT CURVE. THE REACTOR MUST NOT BE MADE y ~"

CRiflCAL UNill THE PRESSURE-TEMPERATURE COM81 Nail 0NS ARE 10 THE g RIGHT OF THE CRITICALITY LIMil CURVE. MARGINS OF 25 PSIG AND 10f ARE INCLUDED FOR P055tBLE INSTRUMENT EPROR.

m 1600 -

3 c *

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,, 1400 -

E

" 1200 -

w e

' APPLICABLE FOR HEATUP RATES OF < 100 F/HR CRITICALITY T

$ 1000 _ ([ 50 F IN ANY 1/2 HOUR PERl00)- Lluli e

3 F y 800 -

0 8

nEssuRE TEwPERAluRt

$o 600 -

. PolNT A 415 70

% #8 C e 525 176

& 400 -

A C 525 215 o D 775 215

- 2250 E 308

@ 200 I 965 276 *

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E G 2250 348 0 I I I I I I I I I  ! I l l I I I I 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 indicated Heactor CO0lant System Temperature, Tc,*F Unit 1 Only REACTOR COOLANT SYSTEM HEATUP LlHITATIONS, APPLICABLE FOR FIRST 5 EFPY sest rems OCONEE NUCLEAR STATION Figure 3.1.2-1 A

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POINT TEMP. PRESS.

A 40 550 8 275 550 C 275 1400 2400 _

0 380 2275 2200 -

2000 -

a

. 1800 - 4 5

UPPER PRESSURIZATION e LIMIT 1400 - -C E

% 1200 _

2 .

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[ 1000 -

3L3 "a> 800 -

2 3 600 -

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A l,i l

400 - l MAXIMUM HEATUP RATE,*F/HR l

200 -

50 _ 100

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' I ' I ' l ' I ' I 0 100 200 300 400 500 275 Indicated Reactor Coolant System Temperature, F 1 OCONEE NUCLEAR STATION Units 2 & 3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS

( A PPLICABLE UP TO AN INTEGRATED EXPOSURE OF 1.7 x 10 18 n/ c m2 OR OTT - 14 4 ' F )

F i gu r e 3.1.2 -18 3.1-6 a

2600 .

$2400 - [

- THE ACCEPTABLE PRES $uRE AND 1EMPERATURE COMBINAll0NS ARE BELOW AND 10 MARGINS OF 25 PSIG AND 10f ARE INCLUDED

{2200 - THE RIGHT OF Tile LIMll CURVE.

FOR POS$18LE INSTRUMENT ERROR.

C 2000 _ l. WHEN THE DECAY MEAT REMOVAL (DMR) SYSTEM 15 OPERAllNG WiiH NO RC PUMPS OPERAllNG, THE INDICAIED DHR SYSTEM RETURN TEMPERATURE 10 IHE REAClok

". VESSEL SMALL BE USED.

E 2. A MAXIMUM STEP TEMPERA 1uRE CHANGE OF 75*F 15 ALLOWABLE WHEN WEMOVING THE E 1600 -

ALL RC PUMPS FROM OPERATION Wl1H IHE DHR SYSTEM OPERAllNG.

$1EP IEMPERATURE CHANGE 15 DEFINED AS THE RC TEMP (PRIOR TO '

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STOPPING ALL RC PUMPS) MINUS THE DHR REluRN TEMP (AfTER STOPPING ALL RC RUMPS) till 100*f/HR RAMP DECREASE iS

{ 1400 -

O ALLOWABLE B0lH SEFORE AND AFTER THE STEP TEMP CHANGE.

APPLICABLE FOR COOLDOWN RATES 3

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• 1200 - 0F I 100 F/HR (s 50 F IN ANY si Note: Applicable cooldown rates -

• O 532 to 4320F, s1000F/HR; l/2 HOUR PEH100) (2) 432of to Cold shutdown, 500F h 1000 -

in any 1/2 hour period 0 P0 INT PRES $uRE iEMPERAluRE 3 800 -

A 375 70 8 525 125 C 525 alb 3 600 -

8 D 900 2th

% C E 2250 377 y

400 -

{ 200 -

1 I l l I I I I I l l I I 0 I l I I 220 240' 260 280 300 320 340 360 380 60 80 100 120 140 160 180 200 40 indicated React 0r C00lant System Temperature. TC. F (I)

Unit i Only REACTOR COOLANT SYSTEM C00LDOWN

. LlHITATIONS, APPLICABLE FOR FIRST 5 EFPY at OCONEE NUCLEAR STATION Figure 3.1.2-2A

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PolNT TEMP PRESS A 380 2275 8 275 1400 C 275 550 0 250 550 E 250 450  !

2400 - F 175 450 A

/ G 175 200 H 120 200 2200 -

RC PUMP COMBINATIONS ALLOIABLE:

ABOVE 185F ALL BELOW 185F 1 -4,1 B;0-4.2 -8;1 - A.0-8;0- A,1 -B g 2000

$ (1) WHEN DECAY H(AT REMOVAL SYSTEM (OH) IS g 100 OPERATING WITHOUT ANY RC PUMPS OPERATING. l INDICATED OH RETURN TEMP. TO THE REACTOR 3

- 1600 -

VESSEL SHALL BE USED.

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/ (2) IN THE TEMPERATURE RANGE 260F TO 175F, A 5 1400 -

MAIIMUM STEP TEMPERATUFE CHANGE OF 75F y

- IS ALLOWABLE FOLLOWED Ef A ONE HOUR 3 1200 -

MINIMUM HOLO ON TEMPERATURE. IF THE STEP CHANGE IS TAKEN BELOW 250F RC TEMPERATURE, 3

- 1000 - THi MAXIMUM ALLOWAdLE STEP SHALL BE THAT

$ WHICH YlELOS A FINAL TEN ERATURE OF 175F.

ch 800 -

THE STEP TEMPERATURE CHANGE IS DEFINED AS g RC TEMPERATURE (BEFORE STEPPING ALL RC PUMPS)

MINUS THE OH RETURN TEMPERATURE TO THE REACTOR b 600 - /

VESSEL. .

j C, 400 -

/

E g g 200 -

UPPER PRESSURI Z ATION __'

0 -

LIMIT MAllMUM C00LOOWN RATE. "F/HR

- (2)-.

100 i 50 t I t 1  ! f f  ! t t

$30 ' 4 I I 260 i 175 120 i 275 400 300 200 100 600 500 i Indicates Reactor Coolant System Temperature.*F II) l f

REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS l

(APPLICA8tE UP TO OTT = 185'F)

Units 2 r, 3 pu mn; OCONEE NUCLEAR STATION Y Figure 3.1,2 2B 3.1-7 a

y2100 c2 -

A D 2500 _

O THE ACCEPTABLE PRESSURE AND TEMPERATURE CONDINATIONS ARE E BELOW AND TO THE RIGHT OF THE LIMIT.CUEVE. MARGINS OF n 2300 -

2S PSIG AND 10F ARE INCLUDED FOR POSSIBLE INSTRUMENT-5 ERROR. FOR COOLDOWN, NOTES 1 AND 2 ON FISURE 3.1.2-2A ARE APPLICABLE.

5 2l00 -

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$1900 -

2

{1700 _

$ APPLICABLE FOR HEATUP OR C00LDOWN E 1500 -

RATES OF s 500F/HR (s 250F IN ANY g 1/2 HOUR PERIOD).

A 1300 -

Y ~ C T

y {1100 -

w

• r0!NT PRES $URE TEMPERAluRE E 900 -

A 525 70

% 8 525 190 sE c 1980 190 o D 2500 274 e

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l l l l l l l l l l l 1 1 I l l 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 Indicated Reactor Coolant System Temperature. Tc,*F REACTOR COOLANT SYSTEM HEATUP AND C00LDOWN LIMITATIONS FOR INSERVICE HYDROSTATIC TESTS (NO FUEL ASSEMBLIES IN THE CORE), APPLICABLE FOR FIRST 5 EFPY OCONEE NUCLEAR STATION Figure 3.1.2-3

q r 3.1.3 Minimum conditions for Criticality -

Specification 3.1.3.1 The reactor coolant temperature shall be above 525cF except for portions of low power physics testing when the requirements of Specification 3.1.9 shall apply.

3.1.3.2 Reactor coolant temperature shal.1 be above the criticality limit of 3.1.2-1A (Unit 1) or above DTT + 100F (Units 2 snd 3).

3.1.3.3 When the reactor coolant temperature is below the minimum tempera-ture specified in 3.1.3.1 above, except for portions of low power

. physics testing when the requirements of Specification 3.1.9 shall apply, the reactor shall be suberitical by an amount equal to or greater than the calculated reactivity insertion due to depressuri-zation.

3.1.3.4 The reactor shall be maintained suberitical by at least 1%Ak/k until a steam bubble is formed and a water level between 80 and 396 inches is established in the pressurizer.

3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality. The regulating rods shall then be positioned within their position limits defined by Specification 3.5.2.5 prior to deboration.

Bases At the beginning of the initial fuel cycle, the moderator temperature coeffi-cient is expected to be slightly positive at operating temperatures with the operating configuration of control rods.(1) Cal.ulations show that above 5250F, the consequences are acceptable.

Since the moderator temperature coefficient at lower temperatures will be 1 less negative or more positive than at operating temperature,(2) startup and operation of the reactor when reactor coolant temperature is less than 5250F is prohibited except where necessary for low power physics tests.

The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.lak/k.

During physics tests, special operating precautions will be taken. In addi-tion, the strong negative Doppler coefficient (1) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduc-tion of moderator density.

The requirement that the reactor is not to be made critical below the limits of Specification 3.1.2-1 provides increased assurance that the proper relation-

ship between primary coolant pressure and temperature will be maintained rela-l tive to the NDTT of the primary coolant system. Heatup to this temperature j will be accomplished by operating the reactor coolant pumps.

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,] O If the shutdown margin required by Specification 3.5.2 is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant pressure.

The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% suberitical will assure that the reactor coolant system cannot become solid in the event of a rod withdrawal accident or a startup accident.(3)

The requirement that the safety rod groups be fully withdrawn before criti-cality ensures shutdown capability during startup. This does not prohibit rod latch confirmation, i.e., withdrawal by group to a maximum of 3 inches sithdrawn of all seven groups prior to safety rod withdrawal.

The requirement for regulating rods being within their rod position limits ensures that the shutdown margin and ejected rod criteria at hot zero power are not violated.

REFERENCES (1) FSAR, Section 3 (2) FSAR, Section 3.2.2.1.4 (3) FSAR, Supplement 3, Answer 14.4.1 3.1-9

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4.2.3 The structural integrity of the Reactor Coolant Syst._3 boundary shall be maintained at the level required by the original accep-

[ tance standards throughout the life of the station. Any evidence,

! as a result of the tests outlined in Table IS-261 of Section XI of the code, that defects have developed or grown, shall be inves-tigated, including evaluation of comparable areas of the Reactor

! Coolant System.

4.2.4 To assure the structural integrity of the reactor internals throughout the life of the unit, the two sets of main internals i bolts (connecting the core barrel to the core support shield and to the lower grid cylinder) shall remain in place and under tension.-

This will be verified by visual inspection te determine that the welded bolt locking caps remain in place. All 'ocking caps will be inspected af ter hot functional testing and whanever the inter-nals are removed from the vessel during a refueling or maintenance shutdown. The core barrel to core support shield caps will be l inspected each refueling shutdown.

4.2.5 Sufficient records of each inspection shall be kept to allow com-parison and evaluation of future inspections.

j 4.2.6 The inservice inspection program shall be reviewed at the end of

, five years to consider incorpor scion of new inspection techniques l and equipment which have been proved practical and the conclusions i

of this review and evaluation shall be discussed with the AECfDOL.

! 4.2.7 At approximately three-year intervals, the bore and keyway of each i reactor coolant pump flywheel shall be subjected to an in-place, volumetric ernmination. Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric ernmination shall be performed, if the interval measured from the previous such inspections is i

greater than 6 2/3 years.

4.2.8 Reactor vessel material surveillance capsules shall be withdrawn after 1,8,16 and 24 effective full power years of operation. The withdrawal schedules may be modified to coincide with those re-fueling outages most cicsely approaching the withdrawal schedule.

Specimens thus withdrawn shall be tested in accordance with Appendix G of 10 CFR 50. A report of the test results shall be i

forwarded to the Commission in accordance with Specification 6.6.1.6.

4.2.9 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspected along their i longitudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal. The elbows to be inspected are identified in B&W Report 1364 dated December 1970.

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4.2-2 i

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Bases i

The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition. The program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.

The reactor vessel material surveillance program is based on the requirement of Appendix H to 10 CFR 50 and is described in B&W report BAW-10006A, Rev 3.

The capsule withdrawal schedule id based on the following:

First capsule - At the time when the predicted shift of the adjusted reference temperature is approximately 500F Second and third capsules - At approximately one-third and two-thirds of the time interval between first and fourth capsule withdrawal.

Fourth capsule - Three-fourths of service life.

Fifth capsule - Standby The specific capsule withdrawn at each of these intervals will be based on current need in best defining the material properties of the reactor vessel.

Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steel base metal when explosively clad with sensitized stainless steel. If no degrada-tion is observed during the two annual inspections, surveillance require-ments will revert to Section XI of the ASNE Boiler and Pressure Vessel Code.

1 4.2-3

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