ML17331B277

Proposed Tech Spec 3/4.4.9.1 Re Heatup & Cooldown Curves

ML17331B277
Person / Time
Site:	Cook
Issue date:	02/22/1994
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331B276	List: ML17331B274 ML17331B277
References
AEP:NRC:0894U, AEP:NRC:894U, NUDOCS 9403040283
Download: ML17331B277 (16)

Text

Attachment 2 to AEP:NRC:0894U EXISTING TECHNICAL SPECIFICATIONS FOR DONALD C. COOK NUCLEAR PLANT UNIT 2 MARKED TO REFLECT PROPOSED CHANGES 9403040283 940222 PDR P

ADOCK 050003l6 PDR

r' 0

O O 2600 IIEAGTOR COOLANT SYSTE ATUP LIMITATIONS APPLICABLE FOR FIRST FFEGTIVE FULL POWER 2400 YEARS. ( MARGINS OF 60 PSIG ANO IO~F AflE INCLUDED FOR POSSIBLE INSTAUMFNT ElIROR )

2200

'EAK TEST LIMIT QJ 2000 U)

(0 180 UJ s-I CL ill HOT 58 f 160 KFPT Rl NDT (l/4 l) i)8 F ACCEPTABLE 3/Ii - l500f OPF RATION 1400 UNACCEPTABLE OPERATION 1200

~ PRESSURE "TEMPERATURE 1000 LIMIT FOR NEATUP RATES O

O' RITIGALITY

UP TO 60 F/HR LIMIT 800 C)

C3 600 UJ K 400 C) 200 f$

~ ) 50 1 00 150 200 250 300 350 400 450 O AVERAGE REACTOR COOLANT SYSTEH TEHPERATURE (deQ F)

FIGURE 3.4-2 FACTOR COOLANT SYBTEH>PRESS'~ - ~~@~ )LIHIgg VKRNN 60 degW~ AATF CRITICALITY LIHIT AN) HYDROSTATIC mSY LIHIT

2600 REACTOR COOLANT SYST M OOLOOWN I.IMITATIONS APPLICABLE FOR FIRST I FFECTIVE FULL POWER 2400 YEARS. ( MARGINS OF 60 PSIG ANO 10 F ARE INCLUOEO FOR POSSIBLE INSTRUMENT RRO .)

2200 2000

~TE+R+A PROf ERlg +A/IS BASE HE)AL 18 0 I ~ 0.5 UNACCEPTABLE OPERATION 16 0 EFPT ATMDI ()/il) 11B F SD F 1400 ACCEPTABLE OP E RATION 1200 PRESSURE- TEMPERATURE 2: Pyr4 LIMITS oo@ 1000 P~

<8 800 o

I COOLOOWN RATE F I IIR O

0 400 20 40 80 200 100 50 150 200 250 300 350 40P 45p AVERAGE REACTOR COOLANT SYSTEH TEHPERATURE (deQ =- F)

FIGURE 3.4-3 fKACTOR COOLANT SYSTKH ~~ ~~ggyc LZg~ V59HUS COOLDHN RATES

ll

~

~

REACTOR COOLANT SYSTEM EASES

4. 4. 9 PRESSURE EMPERATURE LIMITS All components in tha Reactor Coolant System are designed to vith-stand tha effects of cyclic loads dua to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor tzips, and startup and shutdovn opezations. The Section various categories of load cycles used for design purposes ara provided temperature 4.1.4 in of the FSAR. During stirtup and shutdovn, the rates of and pressure changes are limited so that the maximum specified haatup and cooldovn rates are consistent vtth tha design assumptions and satisfy the s tr e s s limits for cyc 1 ic opera t ion.

An ID or OD one-quarter thickness surface flav is postulated at the location in the vessel vhich is found to be the limiting casa. There aza sevezal factors vhich influence tha postulated location. Tha thermal induced banding stress during hcitup is compressive on tha inner surface vhi.le tensile on the outcr suzficc of the vessel vali. During cooldovn the bending stzcss profile is reversed. In addition, the matazial tough-nass is dependent upon irradiation and temperature and therefore tha fluence profile thzough the reactor vessel vali, the rata of heitup and also the rate of cooldovn influence the postulated flav location.

The hcatup limit curve, Figurc 3.4-2, i.s a c"=posite cuzve vhic'.; vas .

prepared by date=in ng thc most conservative case, vith either 0 the inside oz ou side vali controlling, for any hcatup zatc up to 60 F per hour.

The cooldovn li=it curves of Figurc 3.4-3 ize cc=.posite curves vhich vere prepared based upon the same type analysis yith the exception that the controlling location i.s alvays thc inside vali vhcrc the cooldovn thermal gzadicnts tend to produce tensile strcsscs vhile producing compzcssive stzcsscs it-thc outside vill. The hcitup ind cooldovn curves vere prepared based on the most limiting aluc of thc predicted adJusted reference tc pczaturc it the end of 12 EFPY.

Thc

/5 reactor vessel materials have been tested to determine their initial RT<DT', The results of these rests aze shovn in Table b 3/4.4-1.

Reactor operation and resultant fist ncutzon (E > 1 McV) izzadiation v$ 1L cause in increase in tha RT . Therefore an adJustcd reference tam-must, be predicted In accordance vt.th Regulatory Cuide 1.99, NDT'czatura Revision 2. This prediction i.s based on the fluence ind a chemistry factor determined from ona of tvo Positions presented in the Regulatory Cuide.

Position (1) deter ines the chemistry factor from the copper and nickel content of the matczial. Position (2) utilizes surveillance data sets vhich zclatc the shift in reference temperature of surveillance specimens to the fluence. The sclccti.on of Position (1) oz (2) i.s made based on the availability of credible surveillance data, ind the results achieved in applying the tvo Positions .

COOK NUC~~ PLANT - UNIT 2 33 3/4 4-6 NDMENT NO. Pg,Q

,J

'll

~

~

RZACTOR COOLANT SYST~

EASES The actual shift in the reference tempezature of surveillance specimens and neutron fluence is established periodically by removing and evaluating reactor vessel material izradiati.on surveillance specimens and dosimetry installed near the inside vali of the reactor vessel in the core area.

The heatup and cooldovn limit curves of Figures 3. -2 and 3.4-3 include predicted ad]ustments for this shift in RT.NDT at the cnd of 1 EFRY, as ve11 as ad]ustmcnts for possible errozs in the pressure and temperat sensing instruments.

IW e 1 tup and cooldovn curves developed based on the

~ ~

1. The pro)ected ce values establi.she specimen analysis.

2. Int iate shell plate C555 eing the limitin ezial as tezmined by Position 1 egulatory Cuide, Revision 2, vith a copper and nickc content of 0.15% 0.57't, zespectively.

The pzessure-tempezature li=it lines shovn on Figure 3.4-2 for reactor criticality and for inservice leak and hydrosratic testing have been provided to assure compliance vith the minimum temperature requirements of Appendix G to 10 CFR 50.

The nu=bez of reactoz vessel irradiation surveillance specimens and the frequencies for removing and testing these spec';..ens aze provided in Table 4.4-5 to assure compliance vith thc requirements of hp-endix H to 10 CFR Part 50, The li=i.rarions imposed on pressurizer heat p and cooldovn and spray vater te=perarure di.fferential are provided to assuze that the pressurizer is operated vithin rhc design criteria assumed foz the farigue analysis performed in accordance vith the ASME Code requi.rements.

The OPERABILITY of tvo PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 squazc inches censures that the RCS vill be protected from pressure transients vhi.ch could exceed the limits of Appendix C to 10 CFR Part 50 vhen one or more of the RCS cold legs aze less rhan or equal to 152 F. Either PORV oz RHR safety valve has adequate relieving capability to protccr. the RCS from overpzessuzisation vhcn the tzansi.ent is limired to ei,ther (1) the start of an idle RCP vtth the secondary vater temperature of the steam generator less than or equal ro 50 o F above rhe RCS cold leg temperat res or (2) the start of a charging pu=p and its in]ecti,on into a vatcr solid RCS.

3 4 4.10 STRUCTt:RAL INT GRZTY Codex'OOK The inspecr.ion and testing programs for AS~~ Code Class 1, 2 and 3 co=ponents ensure thar. the structural integrity of these co-ponents vill bc

=air'tai.ned at an acceptable level throughout the life of the plant.

extent applicable, the i.nspection program for these co ponents is compl'ance vith Section XI of thc AS'.iE Boiler and Pressure Vessel NUC~~ PiMHT - UNIT 2 B 3/4 4-10 ~~ND~NT NO. yP,Pfg

Insert No. 1 The 15 EFPY heatup and cooldown curves were developed based on the following:

rS The intermediate shellplate, C5556-2, hei;ag- the limiting material as determined by with position 1 of Regulatory Guide 1.99, Revision 2, a Cu and Ni content of .15% and .57%, respectively.

2 ~ The fluence values contained in Table 6-14 of Westinghouse WCAP-13515 report, "Analysis of Capsule U From the Indiana Michigan Power Company D.C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program," dated February 1993.

The RT-NDT shift of the reactor vessel material has been established by removing and evaluating the reactor material surveillance capsules in accordance with the removal schedule into Table 4.4-5. Per this schedule, Capsule U is the last capsule be removed until Capsule S is to be removed after 32 EFPY (EOL).

Capsules V, W, and Z will remain in the reactor vessel and will be removed to address industry reactor vessel embrittlement concerns, if required.

17

~ ~

Attachment 3 to AEP:NRC:0894U PROPOSED TECHNICAL SPECIFICATIONS FOR DONALD C. COOK NUCLEAR PLANT UNIT 2

~ ~

0 n 2600 REACTOR COOLANT SYSTEM HEATUP LIHITATIONS nED  :

APPLICABLE FOR FIRST 15 EFFECTIVE FULL POHER i;i I 2400 YEARS(MARGINS OF 60 PSIG AND 10OF ARE INCLUDED k.t j FOR POSSIBLE INSTRUMENTATION ERROR.)

2200 i ii

':-l"

~ ~ ~ ~ ~

2000 LEAK TEST LIHIT-:

.-'.. ~

7 t.'-.i j

~ I 1800

~

~ 4

~ ~ ~

1600

~

l.'.'. ""'.." '.."

~

'. '. UNACCEPTABLE "..i ACCEPTABLE X

Ill OPERATION i i 'i i i: i i OPERATION I

ttt 0

1400 ~

~ ~

~

I ~ ~

~

~ hhij..

1200

~ ~~ ~

~ ~

O l ~ ~

1000 I: PRESSUR E-TEMPERATURE t i i i i i l t

! LIMIT FOR HEATUP RATES ."".:-

i

CRITICALITY F/HR ='-.~

! l UP TO 60  : i LIHIT 800 ~

\

~ ~ ~

~ ~ ~ ~

600

'MATERIAL PROPERTY BASIS

{INTERMEDIATE PLATE, C5556"2 400 ~ ~ ~ ~

~

Cu .15 X. Ni .57 X

~ ~

~ \ ~

INITIAL RT DT 58 F m

200 .15 EFPY RT (1/4T) 178 F X

t3 m

L (3/4T) ~1500F 2: 05 0 100 150 200 250 300 350 400 450 K 0 o AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE ( F)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM, PRESSURE TEMPERATURE, LIMITS FOR 60 F/HR RATE, CRITICALITY LIMIT AND HYDROSTATIC TEST LIMIT

~ ~

o ~ ~

O 2600  : REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS 0o o  ; i APPLICABLE FOR FIRST 15 EFFECTIVE FULL POWER fE' I

2400 i YEARS (MARGINS OF 60 PSIG AND 10 F ARE INCLUDED M '. i FOR POSSIBLE INSTRUMENTATION ERROR 2200

~ ~

~ ~

2000 tu ~ ~

CC 1800 C(l UNACCEPTABLE '='. i '. i C(l t(l CK Ct. 1600 ACCEPTABLE 0)

K OPERATION Il(

6 l-Cll 1400 I .l PRESSURE-TEMPERATURE ~ ~

M Cll 0(

I- LIMITS z 1200 ~ ~ ~

oo ~ .r ~ ~

o 1000 ~ ~

~ r ~

CX o ~

~

(4 rr ~

~ ~ ~

~ ~

I CJ 800 ~ ~

tx( rw CZ COOLDOWN

~ ~

(I((!I(; ( (

~

(! ( I ( i, 600 ll RATE F/HR

~ ~ ~

MATERIAL PROPERTY BASIS INTERMEDIATE PLATE, C5556-2 400 Cu ~ r15 X. Ni ~ .57 X I

~ ~ ~

~

INITIAL RTNDT 58 F l 100 ~ ~

200 15 EFPY RT DT (1/4T) 178 F

((l oz ~ ~

~ rr ~ ~ ~

~ ~ ~

(3/4T) ~150 F (TI

'z 05 0 100 150 200 250 300 350 400 450 zo AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE ( F)

~ FIGURE 3.4-3 REACTOR COOLANT SYSTEM, PRESSURE TEMPERATURE, LIMITS FOR VARIOUS COOLDOWN RATES

REACTOR COOLANT SYST B SES 3 4.4.9 PRESSURE TEMPERATURE LIMITS 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, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided'n Section 4 ' 4 of the FSAR. During startup and shutdowns

~

the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

An ID or OD one-quarter thickness surface flaw is postulated at the location in the vessel which is found to be the limiting case. There are several factors which influence the postulated location. The thermal induced bending stress during heatup is compressive on the inner surface while tensile on the outer surface of the vessel wall. During cooldown, the bending stress profile is reversed. In addition, the material toughness is dependent upon irradiation and temperature and therefore, the fluence profile through the reactor vessel wall, the rate of heatup and also the rate of cooldown influence the postulated flaw location.

The heatup limit curve, Figure 3.4.2, is a composite curve which was prepared 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 of 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 on the most limiting value of the -.

predicted adjusted reference temperature at the end of 15 EFPY.

The reactor vessel materials have been tested to determine their initial RT~: The results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E ) 1 MeV) irradiation will cause an increase in the RT~. Therefore, an adjusted reference temperature must be predicted in accordance with Regulatory Guide 1.99, Revision 2. This prediction is based on the fluence and a chemistry factor determined from one of two Positions presented in the Regulatory Guide. Position (1) determines the chemistry factor from the copper and nickel content of the material. Position (2) utilizes surveillance data sets which relate the shift in reference temperature of surveillance specimens to the fluence. The selection of Position (1) or (2) is made based on the availability of credible surveillance data, and the results achieved in applying the two Positions.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 4-6 AMENDMENT NO, 6&~423

REACTOR COOLANT SYST BASES The actual shift in the reference temperature of surveillance specimens and neutron fluence is established periodically by removing and evaluating reactor vessel material irradiation surveillance specimens and dosimetry installed near the inside wall of the reactor vessel in the core area.

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT~ at the end of 15 EFPY, as well as adjustments for possible errors in the pressure and temperature sensing instruments.

The 15 EFPY heatup and cooldown curves were developed based on the following:

The intermediate shellplate, C5556-2, is the limiting material as determined by position 1 of Regulatory Guide 1.99, Revision 2, with a Cu and Ni content of 0.15% and 0.57%, respectively.

2 ~ The .fluence values contained in Table 6-14 of Westinghouse WCAP-13515 report, "Analysis of Capsule U From the Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program", dated February 1993.

The RTz shift of the reactor vessel material has been established by removing and evaluating the reactor material surveillance capsules in accordance with the removal schedule in Table 4.4-5. Per this schedule, Capsule U is the last capsule to be removed until Capsule S is to be removed after 32 EFPY (EOL).

Capsules Vg Wg and 2 will remain in the reactor vessel and will be removed to address industry reactor vessel embrittlement concerns, if required.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, one PORV and the RHR safety valve, or an RCS vent opening of greater than or equal to 2 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 152'F. Either PORV or RHR safety 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 temperature or (2) the start of a charging pump and its injection into a water solid RCS.

3 4.4.10 STRUCTURAL INTEGRITY The inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

COOK NUCLEAR PLANT UNIT 2 B 3/4 4-10 AMENDMENT NO. 8&,%RAN