L-82-480, Forwards Addl Info Requested Re Confirmatory Piping Analysis Open Items
| ML17213A630 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 10/29/1982 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| L-82-480, NUDOCS 8211040306 | |
| Download: ML17213A630 (11) | |
Text
REGULATORY FO TION DISTRIBUTION SY M
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ACCESSION NBR e 8211'040306 DOC ~ DATE: 82/10/29 NOTARIZED!
NO FACIL:50 389 St ~ Lucie Planti Unit 2~ Florida Power 8 Light Co>
AUTH~ NAME AUTHOR AFFILIATION UHRIG~R ~ E.
Flor ida Power 8 Light Co ~
REC I P ~ NAME RECIPIENT AF F ILIATION EI'SENHUTiD>G., Division of Licensing
SUBJECT:
Forwards addi info requested re confirmatory piping analysis open
- items, DISTRIBUTION CODE:
BOOIS COPIES RECEIVED:LTR
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ENCL J SIZE:
TITLE: Licensing Submittal:
PSAR/FSAR Amdts 8, Related Correspondence NOTES:
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'00, JUNO BEACH, FL 3340B FLORIDA POWER 8( LIGHTCOMPANY October 29, 1982 L-82-480 Office of Nuclear Reactor Regulations Attention:
ter. Darrell G. Eisenhut, Director Division of Licensing U. S.
Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Nr. Eisenhut:
Re:
St. Lucie Unit No.
2 Docket No. 50-389 Confirmator Pi in Anal sis Attached please find the additional information your staff requested to provide confirmation of the piping analysis open'tems.
If you have any questions regarding this submittal, please contact us accordingly.
Very truly yours, Robert E. Uhrig Vice President Advanced Systems and Technology REU/RJS/JES/jea Attachment cc:
J.
P. O'Reilly, Region II Harold R. Reis, Esquire 82ii04030b 82i029 PDR ADOCK 05000389 A
.PDR PEOPLE
~
~. SERVING PEOPLE
ATTACHMENT Confirmator Pi in Anal sis 0 en Items An investigation was performed for all Safety Class 2
& 3 Systems (irre-spective of operating temperature) to demonstrate that the number of equiva-lent thermal cycles, as defined in ASME subsection NC 3611.2, was sufficiently low to confirm the conservatism of the existing stress analyses.
In accordance with the agreement reached at a meeting with the NRC and Florida Power
& Light Company on October 14, 1982 an acceptance criteria of 1000 "Realistic" cycles was employed.
In conducting this analysis, the following Safety Class 2 and 3 systems were reviewed:
Charging Safety Injection Main Steam Main Feedwater Component Cooling Water Letdown Auxiliary Feedwater Containment Spray Intake Cooling Water A sample calculation specifying methodology and a summary of results is provided on Table I.
Using realistic values of cycle frequencies, all systems were shown to exhibit approximately 700 equivalent cycles.
Using all the thermal trans-ients that appear in the Safety Class 1 specification (Refer to Tables II and III), which are conservative both in frequency and temperature varia-tion, all systems were shown to have less than 1000 equivalent thermal cycles.
Therefore, the above results confirm the conservatism of the ex-isting stress analyses for Class 2
& 3 systems employing weldolets.
TABLE I SAMPLE CALCULATION Thermal Transient 5% Ramp Up 5%
Ramp Down 10% Step Up 10% Step Down Loss of RCP Flow Reactor Trip Loss of Load Loss of Sec ss Normal Var.
Hydro Leak Test Loss of Chrgr
.Loss of Letdown Regen H-X Iso Max Purification Max Dilution Low VCT Norm Start Aux FW Inj Charging ATE Q TI 455 70 455 208 455 47 455 145 455 204 455 187 455 204 455 330 455 6
455 335 455 335 455 335 455 400 455 349 455
- 128, 455 126 455 45 455 455 NI20 9000 8000 2000 32 16 14 500
~- 500 15000 15000 299 2000 2000 40 400 40 5
10 10 200 43 20 50 26 1.
Charging T max = 520 F
0 2.
Ambient Temp (Tamb) 65 F
3-ATE = T max Tamb = 455 F
- 4. N~ of cycles = NI 5.
Temp change of transient =QTX 6.
Ecpxiv 5 of cycles at d TI N = NI (4 TI/d TE)
.Totals 955
TABLE IX
~ 0 SL2-FSAR TABLE 3.9-2 TRANSIENTS USED IN DESIGN AND FATIGUE ANALYSIS Normal Conditions (a) 500 heatup ana cooldown cycles during the design life of the components with heating and cooling at a rate of 100 F/hr between 70 F and 532 F (653 F for the pressurizer).
The heat-up and cooldown rate of the system is'dministratively limited to 75 F/hr to assure that these limits will not be exceeded.
This is basea on a normal plant cycle of one heatup and cool-down per month rounaed to the next highest hundred.
(b) 15,000 power change cycles over the. range of 15 percent to 100 percent of full load at 5 percent of full load per minute in-creasing ana decreasing.
This is based on a normal plant oper-ation involving one cycle per day for 40 years rounded to the next highest 1000.
(c) 2,000 cycles of step power changes of 10 percent of full load, increasing in the 15 percent to 100 percent of full load range ana decreasing in the 100 percent to 25 percent of full load range.
This is based on a normal plant operation involving one cycle per week for 50 weeks'of the year.
(d) 1 x 10 cycles of normal variations of - 100 psi and -
6 F 6
when at operating temperature and pressure.
This was selected based on 1 x 10 cycles being equivalent to infinite cycles ana thus the limiting stress is the endurance limit. - 100 psi is the maximum pressure fluctuation above the setpoint (2235 psig) before backup heaters come on or spray valves open.
For conservatism, the temperature cycle developed for the pressuri-zer is usea for all components.
U set Conaitions (a) 40 cycles of complete loss of reactor coolant flow when at 100 percent power.
This is based on one reactor trip per year for the life of the plant resulting from failure of electrical sup-ply to the reactor co'olant pumps.
(b) 400 reactor trips from full load.
This is based on one reactor trip per month for the life of the plant and includes trips due to operator error and equipment failure.
(c) 40 cycles of turbine trip from 100 percent power with delayed reactor trip.
This is based on one reactor trip per year for the life of the plant considering failure of the turbine trip/
reactor trip circuit as credible.
3.9-63 Amendment No. 5, (8/81)
~ ~
TABLE II SL2-FSAR TABLE 3.9-2 (Cont'd) 3 ~
Emergenc Conditions 5 cycles of complete loss of secondary pressure.
This transient would follow a steam line break.
A steam line break is not con-sidered credible in forming the basis for design of the Reactor Coolant System.
- However, system components will not fail struc-turally in the unlikely event that it does happen.
Faulted Conditions The loading combination resulting from the combined effects of the design basis earthquake and normal operation at full power are categorized as faulted condition.
The loading combinations resulting from the design basis earth-
- quake, normal operation at full power and pipe rupture conditions are categorized as faulted condition.
Design basis earthquake and pipe rupture loadings are combined by the SRSS method.
5 ~
Test Conditions 10 cycles of system hydrostatic testing at 3110 psig and at a tem-perature not less than 60 F above the highest component reference temperature (R'I> T) or 100 F above the highest component section (RT,
,) value.
Phis is based on one initial hydrostatic test plus NDT a mayor repair every four years for 36 years which includes equip-ment failure and normal, plant cycles.
200 cycles of leak testing at 2235 psig and at a temperature not less than 60 F above the highest component reference temperature (RT> T) or 100 F above the highest pipe section RT T.
This is baseh on normal plant operation involving five shuNowns for head removal or valve repair per year for 40 years.
3.9-64 Amendment
?Ao.
6.
(9/81)
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TABLE IXI SL2-FSAR TABLE 3.9-3A A/E SUPPLIED UALITY GROUP A TRANSIENTS PLANT EVENT LIFETIME OCCURRENCES COMPONENT*
CONDITION Plant Cooldown Plant Heatup
'Power Operation Loading/Unloading Ramp 5X per Min Step 10X Reactor Trip Hydro Static Tests, (3125 psia) 50G 500 15,000 2,000 400 10 Leak Test, (2250 psia)
Normal Pressure Variation
(+ 100 psi,
+7 F) 200 10'oss of Primary Flow 40 Loss of Secondary Pressure Loss of '1'urbine-Gen.
Load Purification, Boron Dilution (CVCS) 40 11,000 Loss of Charging Flow (CVCS) 20 Loss of Letdown 4CVCS)
Isolation Check Valve Leaks 50 40
- Definitions of the events Normal (N), Upset (U), Emergency (E),
Faulted (F) and Test (T) are given in ASME III, Para.
NB-3113, 3,9-65 Amendment No. l, (4/Sl)
TABLE III SL2-F SAR TABLE 3 ~ 9-3A (Cont'd)
PLANT EVENT LOCA (Safety Injection)
LOCA (Hot Leg Injection)
LIFETIME OCCURRENCES COMPONENT*
CONDITION F
- Definitions of the events Normal (N), Upset (U), Emergency (E),
Faulted (F) and Test (T) are given in ASME III, Para NB-3113.
- 3. 9-66 Amendment No. I, (4/81)