ML20210T557

Proposed Tech Specs,Revising Heatup & Cooldown Curves to Extend Applicability from Existing 7 EFPY to New Limit of 16 EFPY

ML20210T557
Person / Time
Site:	Farley
Issue date:	09/29/1986
From:	ALABAMA POWER CO.
To:
Shared Package
ML20210T530	List: ML20210T525 ML20210T557 ML20210T571
References
TAC-60075, NUDOCS 8610090071
Download: ML20210T557 (9)

Text

I MATERIAL PROPERTY BASIS CONTROLLING MATERI AL LOWER SHELL (PLATE NO.86919 2)

COPPE R CONTENT 0.14 WT%

NICKEL CONTENT . 056 WT%

INITIAL RT NDT e 5' F

1/4T,146.4 0 F RT NDT AFTER 16 EFPY 3/4T,1215' F CURVES APPLICABLE FOR HEATUP RATES UP TO 60* F/HR FOR THE SERVICE PERIOD UP TO 16 E FPY 2500 i iii iii,,iii , j ,

I iii iiiiIi i i I 1

,8 2250 LEAK TEST LIMIT s j r r i r i i

,i E 4

I i f A 2000 s ,J i UNACCEPTABLE ' '

ACCEPTABLE --t--

OPERATION i r OPERATION C-Z

$ 1750 [ s

(

s r J J

- I A W 1500 / [

$ HEATUP RATES UP i

[

$ TO 600 F/HR a  ;

y 1250 Y# /

g P 0 f l- 1000

< f i

9 f Q  !

E 750 -

CRITICALITY LIMIT

~ BASED ON INSERVICE Y HYDROSTATIC TEST 500 TEMPERATURE (2740 F)

FOR THE SERVICE PERIOD UP TO 16 EFPY 250 l I I l O

0 50 100 150 200 250 300 360 400 460 500 INDICATED TEMPERATURE (O F) 8610090071 860929 PDR ADOCK 05000348 P PDR FIGURE 3.4-2 FARLEY UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/4 4-29 AMENDMENT NO.

MATERIAL PROPERTY BASIS CONTROLLING MATERIAL LOWER SHELL (PLATE NO. 86919-2)

COPPER CONTENT . 0.14 WT%

NICKEL CONTENT  ? 056 WT%

INITIAL RTNDT  ? 58 F RTNDT AFTER 16 EFPY 1/4T,146.48 F 3/4T,1215' F CURVES APPLICABLE FOR COOLDOWN UP TO 100* F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY 2500 i J

B F

2250 j D

I F

2000 j n I

!_. r UNACCEPTABLE / ACCEPTABLE OPERATION [ OPERATION E 1500 3 3 1

= l E 1250 O r m I k 1000 , [

9 COOLDOWN RATifS f O

z 22, F/HR l 750 l

2- 0: !

a 20,x _

5%

2:24 3gg g 77 p ur ,r

::!4060L vC

100 250 0

0 50 100 150 200 250 300 350 400 450 500 l INDICATED TEMPERATURE (O F) l

{

FIGURE 3.4-3 FARLEY UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 16 EFPY 3/4 4-30 AMENDMENT NO.

FARLEY - UNIT 1

REACTOR COOLANT SYSTEM BASES Reducing T av to less than 500*F prevents the release of activity should a steam genera $or tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G as required per 10 CFR Part 50 Appendix G.

1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of nonductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2) These limit lines shall be calculated periodically using methods provided bel ow.

3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.

FARLEY-UNIT 1 B 3/4 4-6 AMENDMENT NO.

E REACTOR COOLANT SYSTEM BASES

4) The pressurizer heatup and cooldown rates shall not exceed 100 F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 F.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of AS!E Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-82 and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 16 effective full power years (EFPY) of service life. The 16 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTndt; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTndt. Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in question, can be predicted using Figure B 3/4.4-1, Figure B 3/4.4-2 and the recommendations of Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTndt at the end of 16 EFPY.

FARLEY-UNIT 1 B 3/4 4-7 AMENDMENT NO.

TABLE B 3/4.4-1 FARLEY UNIT 1 REACTOR VESSEL TOUGHNESS PROPERTIES Material Cu P Ni Tndt RTndt Upper Shell Energy Component Code No. Type (%) (%) (%) (*F) (*F) MWD [c] NMWD[d]

Closure head dome B6901 A533,B,C1.1 0.16 0.009 0.50 -30 -20[a] 140 _

Closure head segment B6902-1 A533,B,C1.1 0.17 0.007 0.52 -20 -20[a] 138 -

Closure head flange B6915-1 A508, C1.2 0.10 0.012 0.64 60[a] 60[a] 75[a] _

Vessel flange B6913-1 A508, C1.2 0.17 0.011 0.69 60[a] 60[a] 106[a] _

Inlet nozzle B6917-1 A508, C1.2 -

0.010 0.83 60[a] 60[a] -

110 Inlet nozzle B6917-2 A508, C1.2 -

0.008 0.80 60[a] 60[a] -

80-Inlet nozzle 86917-3 A508, C1.2 -

0.008 0.87 60[a] 60[a] -

98 Outlet nozzle B6916-1 A508, C1.2 -

0.007 0.77 60[a] 60[a] -

96.5 Outlet nozzle B6916-2 A508, C1.2 -

0.011 0.78 60[a] 60[a] -

97.5 Outlet nozzle B6916-3 A508, C1.2 -

0.009 0.78 60[a] 60[a] -

100 Nozzle shell B6914-1 A508, C1.2 -

0.010 0.68 30 30[a] 148 -

Inter. shell B6903-2 A533,B,C1.1 0.13 0.011 0.60 0 0 151.5 97 Inter shell B6903-3 A533,B,C1.1 0.12 0.014 0.56 10 10 134.5 100 Lower shell B6919-1 A533,B,C1.1 0.14 0.015 0.55 -20 15 133 90.5 Lower shell B6919-2 A533,B,C1.1 0.14 0.015 0.56 -10 5 134 97 Bottom head ring B6912-1 A508, C1.2 -

0.010 0.72 10 10[a] 163.5 -

Bottom head segment B6906-1 A533,B,C1.1 0.15 0.011 0.52 -30 -30[a] 147 _

Bottom head dome B6907-1 A533,B,C1.1 0.17 0.014 0.60 -30 -30[a] 143.5 -

Inter. shell long. M1.33 Sub Arc Weld 0.25 0.017 0.21 0[a] o[a] _ _

weld seam Inter. to lower G1.18 Sub Arc Weld 0.22 0.011 <0.20[b] 0[a] 0[a] _ _

shell weld seams Lower shell long. G1.08 Sub Arc Weld 0.17 0.022 <0.20[b] 0[a] 0[a] ._ _

weld seams

[a] Estimate per NUREG-0800 "USNRC Standard Review Plan" Branch Technical Position MTEB 5-2.

[b] Estimated (low nickel weld wire used in fabricating vessel weld seams).

[c] Major working direction.

[d] Normal to major working direction.

FARLEY-UNIT 1 B 3/4 4-9 AMENDMENT NO.

1020 9

8 7

6 5

s v SURFACE 4 #

- f 3 s f 1/4T-

/ p r

2 . /  ?

p

, / d f

f f 10 19 9 /

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0 f /

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1018 I

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9 8 I l

7 {

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5 I

( 4 f 3

i 2

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10 17 10 15 20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>l MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE (EFPY)

FARLEY - UNIT 1 B 3/4 4-10 AMENDMENT NO.

l

1020 9

8 7

6 5

4 3

2

, SURFACE 10 19 s /

9 ' g 1/4T.

g f 5 7 / /~

c e / j f 5 #

z #

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17-10 0 5 10 15 20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

FIGURE B 3/4,4-2 FAST NEUTRON FLUENCE (E>l Mey) AT 45' AS A FUNCTION OF FULL POWER SERVICE (EFPY)

FARLEY - UNIT 1 B 3/4 4-10A AMENDMENT N0.

REACTOR COOLANT SYSTEM BASES

=

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the 10 CFR Part 50, Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange must be considered.

This Rule states that the minimum metal temperature of the closure flange regions be at least 120*F higher than the limiting RTndt for these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Farley Unit 1). In addition, the new 10 CFR Part 50 Rule states that a plant specific fracture evaluation may be performed to justify less limiting requirements. As a result, such a fracture analysis was performed for Farley Unit 2. These Farley Unit 2 fracture analysis results are applicable to Farley Unit 1 since the pertinent parameters are identical for both plants.

Based upon this fracture analysis, the 16 EFPY heatup and cooldown curves are impacted by the new 10 CFR Part 50 Rule as shown on Figures 3.4-2 and 3.4-3.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two RHR relief valves or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310 F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures or (2) the start of 3 charging pumps and their injection into a water solid RCS.

3/4.4.11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10CFR Part 50.55a(g)(6)(f).

3/4.4.12 REACTOR VESSEL HEAD VENTS The OPERABILITY of the Reactor Head Vent System ensures that adequate core cooling can be maintained in the event of the accumulation of non-condensable gases in the reactor vessel. This system is in accordance with 10CFR50.44(c)(3)(iii).

FARLEY-UNIT 1 B 3/4 4-14 AMENDMENT NO.