ML20050D922

Proposed Tech Spec Changes Updating Figures 3.4-2 & 3.4-3 & Bases for RCS Heatup & Cooldown Curves

ML20050D922
Person / Time
Site:	Beaver Valley
Issue date:	04/05/1982
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20050D907	List: ML20050D903 ML20050D922
References
NUDOCS 8204120430
Download: ML20050D922 (12)

Text

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3ME CURVE APPLICABLE FOR HEATUP R ATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 6 EFPY AND CONTAINS MARGINS OF 10*F AND 00 PSIG FOR POSSISLE INSTRUMENT ERRORS 2500 -

LEAK TEST LIMITS MATERIAL PROPERTY BA$l$

2000 -

CONTROLLING MATERI AL: PLATE METAL E COPPER COMTENT: 0.20 WT%

$ PHOSPHORUS CONTENT: 0.010 WT%

m RTNOTINITIAL: 27'F g RTNDTAFTER 8 EFPY: 1/4T, 274*F g 3/4T,144*F w

g 1500 -

O N

5 s

E 1000 -

HE ATUP R ATES UP

TO GO*F/HR l

500 -

CRITICALITY LIMIT - BASED ON INSE RVICE l HYDROSTATIC TEST TEMPER ATURE (414*F)

FOR THE SERVICE PERIOO UP TO 6 EFPY 0  !

0 100 200 300 400 500 INDICATED TEMPERATURE ( F)

Figure 3.4-2 Beaver Valley Unit No.1 Reactor Coolant System Heatup Limitations Applicable for the First 6 EFPY 3/4 *-z4 PROPOSED WOIDING 8204120430 820405 DR ADOCK 05000334 PDR r

M MATERIAL PROPERTY BA383 CONTROLLING MATERIAL: PLATE METAL g _ COPPER CONTENT: 020 WT%

PHOSPHORUS CONTENT: 0.010 WT%

RTNDTINITIAL: 270F RTNDTAFTER 6 EFPY: 1/4T,274*F 3/47.144*F 8

g 2000 -

E sas CURVE APPLICABLE FoR COOLDOWN RATES E UP TO 100*F/HR FOR THE SERVICE PERIOD D UP TO 6 EFPY AND CONTAINS MARGINS OF 3

• • 8 io*F ANO so Psia rOR POssinLE INSTRUMENT g ERRORS a.

O 1500 -

Y a

E 1000 COOLDOWN g _ R ATE 5 *F/HR 0

20 40 60 100 0

0 100 200 300 400 500 INDICATED TEMPER ATURE (CF) i l

Fig >re 3.4-3 Beaver Valley Unit No.1 Reactor Coolant System Cooldown Limitations Applicable for the First 6 EFPY 3/4 4-25 PROPOSED b'JRDING

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months by verifying that on a Containment Pressure--High-High signal, the recirculation' spray pumps start automatically as follows:

RS-P-1A and RS-P-2B 210 1 5 second delay RS-P-2A and RS-P-1B 225 t 5 second delay

c. At least once per 18 months, during shutdown, by verifying, that on recirculation flow, each outside recirculation spray pump develops a discharge pressure of 2115 psig at a flow of 2 2000 gpm.

d. At least once per 18 months during shutdown, by:

1 Cycling each power operated (excluding automatic) valve in the flow path not testable during plant operation, through at least one complete cycle of full travel, s

2. Verifying that each automatic valve in the flow path actuates to its correct position on a test signal,

e. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

i BEAVER V ALLEY- U NIT 1 3/4 6-14 PROPOSED WORDING

REACTOR COOLANT SYSTEM BASES 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 at the outer wall of the vessel. These stresses are additive to the pressure induced 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 heatup limit curve, Figure 3.4-2, is a composite curve which was pre-pared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 6 EFPY, The reactor vessel materials have been tested to determine their initial RT NDT; e resu ts of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E > l Mev) irradiation will cause an ir.c rea se in the RT Therefore, an adjusted refe rence temperature, based NDT.

upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2. The heatup and cool-down limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT as well as adjustments for possible error 4 in the pressure and temperature sensing instruments.

l BEAVER V ALLEY- U NIT 1 B 3/4 4-6 PROPOSED WORDINC

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O 5 10 15 20 25 30 35 SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

Fast Neutron Fluence (E > 1 Mev) as a Function of Full Power Service Life Figure B 3/4.4-1 BEAVER VALLEY - l;IT 1 B 3/4 4-6a PROPOSED WOEDI';G

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• e2 l *E 2, I I l* l l l l l l l l l 3 2=1017 4 6 8 1018 2 4 6 8 1018 2 4 6 k FLUENCE (N/CM2, E > 1 MEV)

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TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS DATA (UNIRRA01ATED)

=

Up er Shelf Ene n (Ft-lb)

Component Heat No. Code No. Material Type q T

{)

N r- Closure Head .

E Dome C6213-18 B6610 A533B CL. 1 .15 .010 -40 0* 121 -

i Closure Head Seq. A5518-2 86611 A533B CL. 1 .14 .015 -20 -20*

h 131 -

H Closure Head

- Flange ZV3758 -

A508 CL. 2 .08 .007 60* 60* > 100 -

Vessel Flange ZV3661 -

A508 C1. 2 .12 .010 60* 60* 166 -

Inlet Nozzle 9-5443 -

A508 C1. 2 .10 .008 60* 60* 82.5 -

Inlet Nozzle 9-5460 -

A508 C1. 2 .10 010 60* 60* 94 -

, Inlet Nozzle 9-5712 -

A500 C1. 2 .08 007 60* 60* 97 -

$ Outlet Nozzle 9-5415 -

A508 C1. 2 .008 60* 60* 97 -

h tn Outlet Nozzle 9-5415 -

A508 C1. 2 - 007 60* 60* 112.5 -

@ t' Outlet Nozzle 9-5444 -

A508 C1. 2 .09 007 60* 60* 103 -

g Upper Shell 123V339 -

A508 C1. 2 ,

010 40 40* 155 -

@ 4 Inter. Shell C4381-2 B6607-2 A5330 C1.1 .14 .015 -10 73 123 82.5 h Inter. Shell C4381-1 B6607-1 A533B C1. 1 .14 015 -10 43 128.5 90 Lower Shell C6317-1 B6903-1 A5338 C1. 1 .20 .010 -50 27 134 80 Lower Shell C6293-2 B7203-2 A5338 Cl. 1 .14 015 -20 20 129.5 83.5 Trans. Ring 123V223 -

A508 C1. 2 - - 30 30* 143 -

Bottom Hd.

Seg. C4423-3 B6618 A533B C1. 1 .13 .008 -30 -29* 124 -

Bottom Ho.

Dome C4482-1 B6619 A5338 C1. 1 .13 .015 -50 -30* 125.5 -

Core Region Welds .30 37 .013 -

0* -

> 100 Weld HAZ - - -40 -40 -

136.5

• Estimated Per NRC Standard Review Plan Branch Technical Position MTEB 5-2 MWD - Major Working Direction NMWO - Normal to Major Working Ofrection

REACTOR COOLANT SYSTEM BASES Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (referen e nilductility temperature). The most limiting RT e ae a n e e eg n e ret m vessel NDT is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART s designatd as the NDT* NDT higher of either the drop weight nil-ductility transition temperature r the temperature at which the material exhibits at least 50 ft (TNDT) lb of ' impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

RT in reases as the material is exposed to fast-neutron radiation.

NDT Thus, to find the most limiting RT at any time period in the reactor's NDT life, A RT NDT due to the radiation exposure associated with that time period must be added to the original unirradiated RT e ex en esWt NDT.

I in RT is enhanced by certain chemical elements (such as copper and NDT phosphorus) present in reactor vessel steels. The Regulatory Guide 1.99 trend curves which show the ef fect of fluence and copper and phosphorus contents on A RT eac essel steels are show in Ngure B 3/4.4-2.

NDT Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARTNDT can e es ma gu . . ast-neutron fluence (E > 1 Mev) at the 1/4 T (wall thickness) and 3/4 T (wall thickness) vessel locations are given as a function of full-power service life in Figure B 3/4.4-1. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RT NDT'

> BEAVER V ALLEY- U NIT 1 B 3/4 4-7a PROPOSED WORDING

REACTOR COOLANT SYSTEM BASES

'Ihe preirradiation fracture-toughness properties of the Beaver Valley Unit I reactor vessel materials are presented in Table B 3/4.4-1. The fracture toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review plan. I The postirradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Beaver Valley Unit 1 Vessel Material Surveillance Program.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K , for the combined thermal and pressure stresses at any time during heatup and cooldown cannot be greater than the reference stress intensity factor, K , for the metal temperature at that time. K g is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. II The K cune s ghen by the equation:

IR s.

K IR

= 26.78 + 1.223 exp [0.0145 (T-RT

• NDT where K IR is the reference stress intensity factor as a function of the metal temperature T and the metal reference nilductility temperature RT NDT" Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix C to the ASME Code I as follows:

CK g +K <K -}

_ IR

1. " Fracture Toughness Requirements," Brench Technical Position MTEB No. 5-2, Sect ion 5. 3. 2-14 in Standa rd Review Pl.$n_, NUREG-75/087, 1975.

2. ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices,

" Rules for Construction of Nuclear Vessels," Appendix C. " Protection Against Nonductile Failure," pp. 461-469, 1980 Edition, American Society of Mechanical Engineers, New York,1980.

, BEAVER V ALLEY- U NIT 1 B 3/4 4-7b PROPOSED WORDING

REACTCR COOLANT SYSTEM BASES where K

73 is the stress intensity factor caused by membrane (pressure) stress K

It s the stress intensity factor caused by the thermal gradients K is a fun tion of temperature to the RT "" *

• IR NDT C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical At any time during the heatup or cooldown transient, K s determined by IR the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K7 , for the

, reference flaw are computed. From equation (4-2), the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations', composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state s i tua t ion. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T location IR f or finite cooldown rates than for steady-state operation. Furthermore, if cenditions exist such that the increase in K 7g exceeds K It, the calculated allowable pressure during cooldown will be greater than the steady-state value.

BEAVER V ALLEY- U NIT 1 B 3/4 4-8 pK0 POSED WORDING

REACTOR COOLANT SYSTEM 5

BASES The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and insures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K IR f r the 1/4T crack during

(.. heatup is lower than the K IR f r the 1/4T crack during steady-state conditions at the same coolant temperature.

During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and lower K IR's do not of fset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. The refo re , both cases have to be analyzed in order to insure that at any coolant temperature the lower vellue of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

BEAVER V ALLEY- U NIT 1 B 3/4 4-8a PROPOSED WORDING

. . l REACTOR COOLANT SYSTEM

^%

BASES The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a *1/4T deep outside sur-face flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given tem-perature, the allowable pressure is taken to be the lesser of the three values s

taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, 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. Then, com-posite curves for the heatup rate data and the cooldown rate data are adjusted I for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

The actual shift in NDTT of the vessel material will be established l periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples and l

BEAVER V ALLEY- U NIT 1 B 3/4 4-9 PROPOSED WORDING

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