ML20005E790

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Proposed Tech Spec Section 3/4.4.6, Pressure/Temp Limits & Associated Bases
ML20005E790
Person / Time
Site: Limerick Constellation icon.png
Issue date: 12/29/1989
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20005E787 List:
References
NUDOCS 9001110068
Download: ML20005E790 (5)
v • d • e


Text

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- l 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in ls F

accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, j cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve a C for operations with a critical core other than-low power PHYSICS TESTS, with:

i

a. A maximum heatup of 100 F in any 1-hour period,
b. U A maximum cooldown of 100 F in any 1-hour period, f
c. 0 A maximum temperature change of less than or equal to 20 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and
d. The reactor0vessel flange and head flange temperature greater than or equal to 80 F when reactor vessel head bolting studs are under l tension. I W

APPLICABILITY: At all times. 7 ACTION: j, With any of the above limits exceeded, restore the temperature and/or pressure I to within the limits within 30 minutes; perform an engineering evaluation to I determine.the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; deterruine that the reactor coolant system remains  ;

acceptable for continued operations or be in at least HOT SHUTDOWN within 12 '

hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS

  • f 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curve A, B, or C as applicable, at least once per 30 minutes.

9001110068 891229 PDR P

ADOCK 05000352 PDC LIMERICK - UNIT 1 3/4 4-18

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[ - REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)'

t 4.4.6.1.2- The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve C within 15 minutes prior to the withdrawal of control rods to bring the l L reactor to criticality and at least once per 30 minutes during system heatup.

4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel. material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1. The results of'these examinations shall be used to update the curves of Figure 3.4.6.1-1. .

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4.4.6.1.4- The reactor flux wire specimens shall be removed at the first refueling outage and examined to determine reactor pressure vessel fluence as a function of time and power level and used to modify Figure B 3/4 4.6-1. The results of these fluence determinations shall be used.to adjust the curves of Figure 3.4.6.1-1, as required.

4.4.6.1.5 Thereactorvesselflangeandhgadflangetemperatureshallbe verified to be greater than or equal to 80 F:

a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
1. 5 1000 F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2. 5 900F, at least once per.30 minutes, s
b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

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LIMERICK - UNIT 1 3/4 4-19  :

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c h , -

P MIN RPV METAL TEMP VS. RX VESSEL PRESS FIG URE 3.4.6.1-1 i i

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(. P.

A- SYSTE M HYDROTEST LIMIT A B C R- WITH FUEL IN VE8SEL

'880 [ f i f

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j E s. N ON.N U,CL E AR H E ATIN G '

f S 1300 A N D-V E S S E L-C O O L D OW N- LI M I T .

S 3250 ..C _.N UALE AR_(CdR E CHillCAL) 1

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,. LIMIT BASED ON GE BWR fl l f i -j 1200 1:lC E N 8 t N G TO PICA L-R E P O RT j g  ;

{ N E DO- 2177 8 A /

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d 1100 t

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l NOTEi CURVES A,B, AC / i

' 1050 AN E~P~A'E DTCYE 070~A PVT.Y

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l .l' 4 N 1000 A S -T H E - L I M I T S F O R U P TO -- - - - --- - - - - - - -

12 E F P Y O F O P E R ATIO N

'8 -~ ~ ~ ~ ~ ' ' ~ ~ ~ ~ " ' ~ ' - ' - ~ ~ - ' ~ ~ -

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)( 900 - - - - - - - - - - - - - - - - - -- - - - - - - -

y 850 4 - - - - - - - - - - - - - - - - - - - - - - - -

L- 800 4- - -

E - - - - - - - - - - - - - - - -

s S 750 ~ ~ ~ ~ - - ----- - -- - - - - - - - - -

S 7on - . - - . - - . - _ - _ . . . . . - - - _ - - .-

E ,,, . _ _ . _ _ . . _ . . . . _ .. . _ . _ ___ _ __

! L Si0 o.io ,

I ggn .- _ _ , - . - - . ~. .

O 500 ~ ~ - ~ ~ - - - - - - - - - ~

p 450 --.10 C F R S DIA P P._G ..__, . . _ _ . _ - . . .. - . . _ _ _ _ . _

H 400 LIMIT 10 F M, _ , _ , _ _ _ _ _. _ _ , , _ _ _

E 3,, . _ . _ . . . . _ _ . . . . _ . _ . . _ _ _ . _ _

300. .12.5.{p sig L . . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

.. _ _7 250 - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

P 200 7-_-_ _ -

S ,,,

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/

G 70 paio! /:

50 - - - - --"

40 pelg 0

0 25 50 72 100 125 150 175 200 225 250 275 000 325 050 8 0'F MIN RX VESSEL METAL TEMP (T)

LDERICK - UNIT 1 3/4 4-20

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CRF 50 Appendix.G and ASME Code Section III, Appendix ,

G. The curves are based on the RTNDT and stress intensity factor information j for the reactor vessel components. Fracture toughness limits and the basis for j compliance are more fully discussed in FSAR Chapter 5, Paragraph 5.3.1.5,. l

" Fracture Toughness." i The reactor vessel materials have been tested to determine their initial l RTNDT. The-results of these tests are shown in Table B 3/4.4.6-1. Reactor l operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause  :

an increase in the RTNDT. Therefore, an adjusted reference temperature, based l upon the fluence, nickel content and copper content of the material in question, j can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations of  !

Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel

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Materials." The pressure / temperature limit curve, Figure 3.4.6.1-1, curves A, B l and C,.are based on the non-beltline, discontinuity areas of the RPV which do not '

receive significant neutron fluence and the RTNDTS will, therefore, not shift. 1 These limit curves are predicted'to be bounding for all areas of the RPV until 12- l EFPY when the beltline material's RTNDT will shift due to neutron fluence and I the beltline curves will intersect the non-beltline discontinuity curves. The '

non-limiting beltline curves are not shown on Figure 3.4.6.1-1, but are included j on FSAR Figure 5.3-4.

The actual shift in RTNDT of the vessel material will.be established .f periodically during operation by removing and evaluating, in accordance with 10  ;

CFR Part 50, Appendix H, irradiated reactor vessel flux wire and Charpy specimens  !

installed near the inside wall of the reactor vessel in the core area. Since the  !

neutron spectra at the flux wires, Charpy specimens and vessel inside radius are essentially identical, the irradiated Charpy specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of .

I the-flux wire and Charpy specimen data and recommendations of Regulatory Guide 1.99, Revision 2. This would include showing the beltline (versus non-beltline discontinuity) limits when they become bounding.

The pressure temperature limit lines shown in Figures 3.4.6.1-1, curves C, and  ;

A, 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 Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are provided in Table 4.4.6.1.3-1 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

LIMERICK - UNIT 1 B 3/4 4-5 I

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7 BASES TABLE B 3/4.4.6-1

  • REACTOR VESSEL-TOUGHNESS t

HEAT / SLAB MIN BELTLINE WELD SEAM I.D. OR RT STARTING A MAX.

  • SHELF. UPPER RpX.

COMPONENT OR MAT't TYPE HEAT / LOT CU(%) Ni(%) NDT ( F) RTN nT (*F) -(LFT-LBS) NDT ('F)

Plate SA-533 Gr B CL.1 C 7677-1 .11 .5 +20 +66 ,

NA 486 Weld SFA 5.5, 662A746/ .03 .88 -20 +25 NA +5 (E 80T8-G) H013A27A NOTE:

  • These values are given only for the benefit of calculating the end-of-life (EOL) RTNDT NON-BELTLINE MT'l TYPE OR HIGHEST STARTING HEAT / SLAB OR COMPONENT WELD SEAM I.D. RT HEAT / LOT NDT { F) {

Shell Ring SA 533, Gr. 8, CL. 1 C7711-1 +20 Bottom Head Dome "

C7973-1 +12 Bottom Head Torus "

C7973-1 Top Head Dome " +12 A6834-1 +10 Top Head Torus "

B1993-1 +10 Top Head Flange SA-508, CL. 2 1238195-289- 0 Vessel Flange "

2V1924-302 -30 feedwater Nozzle "

Q2022W-412- Weld Non-Beltline All 0 LPCI Nozzle ** SA-508, CL. 2 Closure Studs SA-540, Gr. B-24 Q2Q25W -6 l' All Meet requirements of 45 f t-lbs and 25 mils Lat. Exp. at +10 F

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Note: ** The design of the LPCI nozzles results in.their experiencing an EOL fluence in excess of 17 2 10 N/Cm which predicts an EOL RTNDTo 'f +39 F.

LIMERlCK - UNIT I B'3/4 4-7

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