ML20086M125

Amend 199 to License NPF-3,revising TS 3/4.4.9.1 pressure- Temp Limits,Rcs,Figures 3.4-2,3.4-3 & 3.4-4,Bases Section 3/4.4.9 & License Condition 2.C(3)(d)

ML20086M125
Person / Time
Site:	Davis Besse
Issue date:	07/20/1995
From:	Gundrum L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20086M128	List: ML20086M125 ML20086M132
References
NUDOCS 9507240136
Download: ML20086M125 (8)

Text

-

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.g e yi UNITED STATES l

j NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 20b5!W001 o

%.....}

TOLEDO EDIS0N COMPANY CENTERIOR SERVICE COMPANY t

8!!D l

THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET N0. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.199 License No. NPF-3

.l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company (the licensees) dated January 30, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; i

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C of Facility Operating License No. NPF-3 is hereby amended to read as follows:

9507240136 950720 i

PDR ADOCK 05000346 P

PDR

4 2.C.(2)

. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment _No. 199, are hereby incorporated'in the license.

i The Toledo Edison Company shall operate the facility in accordance with the Technical Specifications.

2.C.(3)(d) Prior to operation beyond 21 Effective Full Power Years, the.

Toledo Edison company shall provide to the NRC a reanalysis and proposed modifications, as necessary, to ensure continued means of protection against low temperature reactor coolant system overpressure events.

l 3.

This license amendment is effective as of its date of issuance and shall be implemented not later than 90 days after issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h,,&

Mbt w,n Linda L. Gundrum, Project Manager Project Directorate 111-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: July 20, 1995

ATTACHMENT TO LICENSE AMENDMENT NO.190 E&CILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.' The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Insert 3/4 4-25' 3/4 4-25 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 B 3/4 4-10 B 3/4 4-10 B 3/4 4-11 B 3/4 4-11

r cp 3C.

m.

A, Ficure 3.4-2 v

v, Reactor Coolant System Pressure-Temperature Limits g

for Heatup and Core Criticality for the First 21 EFPY a

2600 i

i j

plH i

pK w 2400 noie.:

vi cn 2200

'l 8

. uiowabie n eup rai. is so rn,inamp3. s,re,ed by. t5 F step change I

followed by an 18 minute hold.

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i S 2000 y

' 2. When decay heat system (DH) Is opera 6ng without RC pump opera $ng. Indicated DH re' urn i

i i

i j

n-1800 temperature to the reactor vessei she5 be used.

w 6

3. The acceptable pressure and temperature combinatione are below and to he dght of the limit

Critical.ity R

E I600 j

P i'"I' !

R m

4. insvymoni erer is noi.ecounied ror in mese amin.

r'o 1400 i

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~

_E.

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i Point Temo Press O

1200 d

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! 4(J !

G A

70 325 1000 B

8s 327 -

u 9

C 115 330 i

M 800 o

178 477 --

j j

j j

l l

l l

l j

i E

230 477 m

600

"!' stW Limi '.

F 238 810 -

t y

u

.]

G 280 1211

.S 400 en E

ct H

345 2500 -

--- - -T a

73

^

B, 1

339 0

5 200 9

J 339 1130 _

y:

K 402 2500 I

i i

l i

l i

l i_ l

.o 0

i 50 100 150 200 250 300 350 400 450

.Ci Indicated Reactor Coolant Inlet Temperature, F e

_=._

a.

e D<:

~,

v Im O,

Flouro 3.4-3 vm Reactor Coolant System Pressure-Temperature Limits 5

for Cooldown for the First 21 EFPY 2600 i

i

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p m

%=:

1-

]

y }fQQ

. 1. Allowable cooldown role above 270 F le 100 Fair (Anmp). Smited by -

j

_j

....q........g........a.........s........y..

..p...

........g_._

u>

a 15 F step change followed by an 9 minute hold.

......g...... _ 4....... 4...

...p

.....p...H..:.

8 a

p....g........g..._

-g

2. Allowable cooldown rate batow 270 F is 50 F/hr (Ramp). limited by G.

,o,, 2000 a 15 F step change followed by en 18 minute hold.

i

- ------- ---- - :- --- -- r. ----- :i --- -- t -- ---i v1 8

3. A maulmum step semperature change of 15 F is allowabio when mmoving i

i i Point Temp Press i i

C1.

I8QQ as RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR meum tamp :j----- -{i

-- --p --- -- y -

--}..- ~ j....._

g

~

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A^

70 156 i

i 1600 to the mactor coolant sysisen prior to stopping all RC pumps.

w o

-.-------.------.----.:-- e 120 30s

~

us X

4. when the decay heat removal system (DH) la operadne without any l

l l

C 170 477 j

j-1400 RC pumps opereeng. Indicated DH ruum temperature to the F-- i------i----- - l-D.

195 477 L---i- --- -!-- --

ru c

re&ctor vessel sheli be used.

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.i E

-198.

640, i i

es o

1200

'1419.l......l....._

5. The acce,tsbio,,es.u,e and._

combin..on,e,, -i ---i---

---i-----y F

270-l-

d below and to the right et the smet curve.

i i.

G 315 2219. j j

1000

- - - ----i -- --- i ---- - i------- i........ J.

H 405 2326....i........ i... -

uo

6. Instrument error Is not semunted for in Wiese Emits.

i i

i i

I 483 2500

.i i

a 800

- - - - - - -.$----*-----*------4-------+-------->---

->------*-----------*------i--------+-----i------>-------*------*---'-----

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4 cr j

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- - - i -- -- !- -- - :i- ---- i---- -- l- ---- - E--- !------- !- ----- i -- -- !----- - i--- - i-- - i------ 4 ------ i -- - ! ---- i--- --

.o 600 o

i

-c

- - ---- :-- - - :. --- g:: ---- p:--- -- - :-- ---- : --- -- : --- --- :-- ----- :---

3

,o 400 r

c g'.

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B : -- :- - -- >: ------- : --- -- 3 -- --- :- ---- : ------ :- --- - :- ----- i-- ---- : ------ 2 ---- :------- :- -- :- -- -

200 n>

A:

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1

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=>

O 3;

50' 100 150 200 250 300 350 400 450 500-i 5

Indicated Reactor Coolant Inlet Temperature, F :

b

es 23 Fiauro 3.4-4 v.i m

Reactor Coolant System Pressurc-Temperature IIcatup and 3:

Cooldown Limits for Inservice Leak and Ilydrostatic Tests for the First 21 EFPY.

r 2600

~

l l

g

==

2400 b

~

.9

1. A!Iowablo hentup rate is 50 F/hr (Ramp), Emited by a 15 F stnp change

[

j j

J j

l a

toRowed by an is minuta hou.

' 2200 2

2. Allowable moldown rato above 270 F is 100 FAr (Ramp). Em!tod by i

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E 2000 a 15 F step change followed by an 9 minute hoM.

m 2

3. Allowable cooldown rete below 270 F le 50 F/hr (Remp). Ilmited by l

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A 1800

. is r siep change ioriowed by is minvie hoid.

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g gg

4. A maximum step temperature change of 15 F is allowable when removing i

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at RC pumps from operation with he DHR system oparn6ng. The stop j

a ey temperature change is dorced as RC temp minus the DHR retum temp i

i i Point Temo Press i w1 1400 te ihe coactor cooiani ersiem prior is stoppino ='i ne pumps.

Ei i

I A

70 254 i

5""*"'"*****Y**"***'''*'")"'"'""""'""'

h l200

~

RC pumps operating. Indented DH retum Inmperature to the iF i

i B

90 364 m

y reactor vesser shan be used.

C 125 477

[

}-

1000 o

205-477 l

u 9

s. The acceptable pressure and temperature combinations are gl l

i i

E 215 935 i

M 800 6*'*"""d'****'*'**'*"*""*-

F 250 1242 l

l N.

7. Instrument error is not acmuntad for in these Rmits, j

j i

i G

318 2500 ;

I m

600

l E

i i

i i

i i

i i

i i

i fj 400 4

B.:

i

=

w 200 a

a A:.

I I

I I

I I

I I

I I

I 0

g 50 100 150 200 250 300 350

.400 b

Indicated Reactor Coolant inlet Temperature, F

.t

@W

I REACTOR COOLANT SYSTEM BASES The closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt pre-load).

This region largely controls the pressure-temperature limitations of the first several service periods.

The outlet nozzles of the reactor vessel also affect the pressure-temperature limit curves of the first several service t

periods.

This is due to the high local stresses at the inside corner of

.l the nozzle which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RT, temperature of the beltline region materials will be high enough so that the beltline region of the reactor vessel will start to control the pressure-temperature limitations of the reactor coolant pressure boundary.

For the service period for which the limit curves, are established, the maximum allowable pressure as a function of fluid temperature is obtained j

through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region. The maximum allowable pressure is taken to be the lower pressure of the three calculated oressures.

The pressure limit is adjusted for the pressure differential betwee1 the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The limit curves were prepared based upon the most limiting adjusted reference temperature of all the beltline region materials at the end of twenty-one effective full power years.

The actual shift in RT of the beltline region material will be established m

periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, 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 vessel inside the radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The limit curves must be recalculated when the ART determined from the surveillance capsule is different from the calculated k for the equivalent capsule radiation exposure.

DAVIS-BESSE, UNIT 1 B 3/4 4-10 Amendment 4% 199

REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available. The adjusted reference temperatures are calculated by adding the predicted radiation-induced ART,31 and the unirradiated RT,31 The procedures described in Regulatory Guide 1.99, Rev. 2, were used for predicting the radiation induced ART,37 as a function of the material's copper and nickel content and neutron fluence.

Figure 3.4-2 presents the pressure-temperature limit curve for normal heatup. This figure also presents the core criticality limits as required by Appendix G to 10 CFR 50.

Figure 3.4-3 presents the pressure-temper-ature limit curve for normal cooldown.

Figure 3.4-4 presents the pressure-temperature limit curves for heatup and cooldown for inservice leak and hydrostatic testing.

All pressure-temperature limit curve are applicable up to twenty-one ef fective full power years.

The protection against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.4-2, 3.4-3 and 3.4-4.

l l

DAVIS-BESSE, UNIT 1 B 3/4 4-11 Amendment No. 444,199

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