ML20082S828
| ML20082S828 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/06/1991 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20082S825 | List: |
| References | |
| NUDOCS 9109170311 | |
| Download: ML20082S828 (30) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _
e o ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR Pl. ANT UNITS 1 AND 2 DOCKET NCS. 50-327 AND 50-328 (TVA-SQN-TS-90-01. RI)
LIST OF AFFECTED PAGES Unit 1 3/4 4-24 3/4 4-25 B 3/4 4-6 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-13 Unit 2 3/4 4-30 3/4 4-31 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-14 9109170311 910006 PDR ADOC.K O*JOOO327 P PDR
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s- o. CURVES APPLICABLE FOR HE'ATUP RATES UP-TO 60*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY. MARGINS OF 60 PSIG AND 10* F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR.
' MATERIAL PROPERTY BASIS - SON UNIT I CONTROLLING MATERIAL: LOWER SHELL FORGING 3000 COPPER CONTENT: 0.I3 WT%
, NICKEL CONTENT: 0.76 WT% INITI AL RTNDT: 73*F - j RTror AFTER 16 EFPY: t/4T, 195'F i 3/4T, 166*F 2500 -
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0 0 v CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY, MARGINS OF 60 PSIG AND 10*F ARE INCLUDED FOR POSSIBLE INSTRUMENT ERROR. MATERIAL PROPERTY BASIS - SON UNIT i CONTROLLING MATERIAL: LOWER SHELL FORGING 3000 - COPPER CONTENT: 0.13 WT% L __ NICKEt CONTENT: 0.76 WT% INITIAL RTNoT: 73*F "~ RTNui AFTER 16 EFPYi 1/4T, 195* F _ 3/4T, I66*F
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i 1 J i . 50 - j l l l i ; i l . i 0 25 50 100 200 300 400 500 AVERAGE REACTOR COOLANT SYSTEM TEMPERATURE (*F) FIGURE 3,4-3 SEOUOYAH UNIT I REACTOR COOLANT SYSTEM COOLDOWN LIMITATICNS APPICICABLE UP TO 16 EFPY , i SEOUOYAH - UNIT I 3/4 4-25
i o REACTOR COOLANT SYSTEM BASES l 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G.
- 1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with-the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the first full power service period.
a) A110weble combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may ba obtained by interpolation, b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g. , pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2) These liniit lines shall be calculated periodically using methods provided below.
4.
- 3) The secondary side of the steam generator must not be pressurized above I 200 psig if the temperature of the steam generator is belov 70 F.
- 4) The pressurizer heatup and cooldown rates shall not exceed 100 F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the sprhy fluid is greater than 560 F.
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- 5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
DSER.T 1 The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, I h and ASTMTIEDiP, and in accordance with additional reactor vessel require-ments These properties are then evaluated in accordance ,th(Appendix G of thD107; kmcr Addend; tc Section !!! cf the ASME Boiler and Pressure Vessei Code,and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975." i
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_ - ~ June 25, 1985 SEQUOYAH - UNIT 1 B 3/4 4-6 Amendment No. 4C \
8 0 INSERT 1 10 CFR 50, Appendix G, addresses metal temperature of the closure head flange and vessel regions. Appendix G states that the minimum metal temperature of the closure flange region should be at least 120 degrees Fahrenheit (F) higher than the limiting RTNDr for this region when the , pressure exceeds 20 percent of the preservice hydrostatic test pressure (561 pounds per square inch gauge (psig) for Westinghouse Electric Ccrporation plants). For SQN. Unit 1, the minimum temperature of the closure flange and vessel flange regions is 90 degrees F since the ; limiting initial RTuor for the closure head flange is -40 degrees F (see Table B 3/4.4-1). These numbers (561 psig and 90 degrees F) include a margin for instrumentation error of 10 degrees F and 60 psig. The SQN Unit I heat up and cooldown curves shown in Figures 3.4-2 and 3.4-3 are not impacted by this regulation. i I l l i
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o o REACTOR COOLANT SYSTEM RASES Heatup and cooldcwn limit rurves are calculated usino the most limitino value of the nil-ductility reference temperature, PTNDT, at the end of N effective full power years of service life. The N EFPY service life period is chosen such that the limitino RTNDT at the 1/4T location in the core rmion is creater than the PTNDT of the limitino unirradiated naterial. The selection of such a limitino PTNDT assures that all components in the Peactor Coolant System will be operated conservatively in accordance with applicable Code recuirements. The reactor vessel materials have been tested to detemine their initial PT NDT; the results of these tests are shown in Table B 3/4.4-1. Peactor operation and resultant fast neutron (E areater than 1 FEV) irradiation Therefore, an adjusted reference can cause an increase in the PTNDT. tenperature, based upon the fluence of the material in cuestion bn M fredicted = ine " ; = 3/'.' ', The heatup and cooldown limit curves
- $v of Floures 3.4-2 and 3.4-3 include predicted adjustnents for this shif t in PT at the end of}rf'EFPY, as well as adjustnents for possible NDT errors in the pressure and temoerature sensino instruments.
detemined in this manner may be used until the Values of delta PTNOT results from the material surveillance prooram, wa evaluated accordino to ASTP E185, are available. The first capsule uP'sbe removed at the end of the first Core cycle. Successive capsules will be removed in accordance with the reouirements of ASTF ElP5-7J and 10 CFP 50, Appendix H. The heatuo and cooldown curves must be recalculated when the delta RT detemined from the surveillance capsule exceeds the cciculated FDT delta RT for the e uivalent capsule radiation exposure, NDT Re A% G;de. 1.99, G.;g, ,z b g-. has been fred.'e+<l ui.n I 94 ulo " g g Q c and a fe<Y. sve face tioence. WcnP s297o Ilo 2 ffec.febe. full Lk:Y fower ycars Uurves for (Reference ^Jerma l 6fwa h A " Hert- f~ and Ceelda u ~ ~ q - t SENIOYAH - 'R'I T 1 P 3/d 47 l
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,6 l-Parallel to Major Woritne Directina a-158. ear act reported c-Plateue reper shelf energy decreased to Si at a test 7-Nnaal to ka ter Hor 6 f ng Pirection twMinleum upper shelf emerites temperature of 3nn*f. This a m ly util be reevalmated
*f st feete fased on LEAIC Regulatory Standard Seview Plan, Section 5.3.7 PTfB d.ee the resvits of feneric tas6 A.11 are a*a814tle.
e -o REACTOR COOLANT SYSTEM
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BASES thermal stresses and different Ky 's for donotoffseteachotherandthehressuresteadystateandfiniteheatuprates 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. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite hedtop rates is cbtained. The second portion of the heatup analysis concerns the calculation of i pressure-temperature limitations for the case in which a 1/4T deep outside surface 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, of course, are dependent on both - the rate of heatup and the time (or coolant temperature) along the heatup ' ramp. Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. i Following the generation of pressure-temperature curves for both the - steady-state and finite heatup rate situations, the f'nal limit curves are produced as follows. A composite curve is constructed based on a point-by point puun, . comparison of the steady state and finite heatup rate data. At any given
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temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under. consideration. , r The use of the composite curve.is necessary to set conservative-heatup limitations because it is possible for conditions to exist such that 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. ,
] Instr 55 '" Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.
Although the pressurizer operates in temperature ranges above those for i which there is reason for concern of non-ductile failure, operating limits , are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. ,
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o ( SEQUOYAH UNIT - 1 B 3/4 4-13
- D
e: o INSERT 2 The leak test limit curve'shown in Figure 3.4-2 represents the minimwn temperature requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of Branch Technical-Position MTEB 5-2 and 10 CFR 50, Appendix G. The criticality limit cur *;e shown in Figure 3.4-2 speelfie. pressure-temperature limits for core operation to provide additional nurgin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) require the reactor vessel to be at a temperature equal to or higher than the minimum temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum pressure-temperature curve for heatup and cooldown. The maximum temperature for the in-service hydrostatic test for the SQN Unit I reactor vessel is 327 degrees F. A vertical line at 327 degrees F on the pressure-temperature curve, intersecting a curve 40 degrees F higher than the pressure-temperaturo limit curve, constitutes the limit for core operation for the reactor vessel, i l l l 6 l
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e O REACTOR COOLANT SYSTEM l
- BASES O di Alc. Mica meMs Nrser fe D' went 7914. fi,%ss's fer Hedef*and j) ,
PRESSURE /ffMPERATUR[ tIMITS (Con'inued) c.c lJ.,,3 n f, :4 Cyr.es , Ayll n yg,
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- 5) System preservice hydrotests and in service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Bot ter and Pressure Vessel Code, Section XI.
"6" The fracture toughness properties df the ferritic materials in the reactor
- _ vessel are determined in accordance with the NRC Standard Review Plan, ,
2, "~~~ ASTM E1857 R and in accordance with additional reactor vessel require- . ments. Thet, crocerties are then evaluated in accordance with Appendix G to 10 CFR 50 and EA6e-44jt4}on-of-Scotton-141--of-the ASME ' ter and g Pressure Vessel Cod p %g, ,, 'lj , g ,, .f-
~ , $rdenK D;,;s,o. t Heatup and cooldown limit curves ar) calalated using the most limiting %% value of the nil-ductility reference temperature, RTNDT, at the end of 16 ef fective tull power years of service life. The 16 EFPY service life period is chosan such that the limiting RT NDT at the 1/4T location in the core region is greater than the RT NOT of the limiting unirradiated material. The selection of such a limiting RT NOT assures nat all g components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. .
The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tesu are shown in Table B 3/4.4-1 Reactor operation and resJ1 tant fast neutron (E greater than 1 MEV) irradiation San cause an increase in the RT NDT. Therefore, a.1 adjuste') reference temperature, based upon the fluence of the material in question,h-be- )
-pwelcted a4ng 549 uW/4 The heatup and '.coldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 16 EFPY, as well as adjustments for possible NDT errors in the pressur e ad temocratur2 sensing instrumen',.
- -% v - %_ v ~m/ %
he5 been pre d ie h d 65ea-keg.later}Sule- I'99, d '** 2-asd 4. fra k sur (see loe st s. e f 0. 944 x 10 n /cm'~ fe.r-llo e M e c 4 : < C. II p ,wac yearg ( L'< { cence w ee p u2.g .,,; ora,s , , and. Cee,ldven L.~d Cures fue rJer., I cycraf,'s rr . 7ve,e I99 t, __.~^-x.- , _ .s 5EQUOYAH - UNIT 2 B 3/4 4-7 , L
e o INSERT 3 1 10 CIR 50. Appendix G addresses metal temperatute of the clohure head flange and vessel regions. Appendix G states tha *. the minimum net al I temperature of the closure flange region should bi at least 120 degrees Fahnenheit (F) higher than the limiting k7933 for this region whon the l pressure exceeds 20 percent of the preservice Sydrostatle test pressure ; (561 pounds per square inch gauge (psig) for Westinghouse Electric l Corporation plants). For SQN, Unit 2, the minimum temperature of the closure flange and vessel flange regions is 117 degrees F sinen the i limiting initial RTygt f or the closure head flange is -13 degrees F (see l Table B 3/4.4-1). These numbers (56'. psig and 117 degrees F) include a trargin f or instrumentation error of 10 degrees F and 60 :>tig. The SQN l'ntt 2 heat up and cooldown curves shown in Figures 3.4-2 and 3.4-3 are not irrpacted by this regulation. !
r e s i REACTOR COOLANT SYSTEM ( j
-s ,
- 0. ASIS.
PRESSURE / TEMPERATURE LlHITS (Continued) Values of ART determined in this manner may be used until the results NOT < f rom the material surveillance program, evaluated according to ASTM E185, [ was are available. The first capsule b: removed at the end of the first core cycle. Successive capsules will be removed in accordance with the { requirements of ASTM E185-JTc%2 and 10 CFR 50, Appendix H. The heatup and t cooldown curves mi t be recalculated when the ARTNOT d't*I'd I"
- th' surveillance capst ,. Jxceeds the calculated ART '
NOT for the equivalent capsule radiation exposure, i i Allowatle pressure
- temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in i Section III of the ASME Boiler and Pressure Vessel Code as required by i Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-7924 A.
The general c.ethod for calculating heatup and cooldown limit curves is i based upon the principles of the linear elastic fracture mechanics (LEFM) , technology. In the calculation procedures a semi-elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimenstions nf this postulated crack, referred to in Appendix G of ASME !!! as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference ; crack are conservative and provide sufficient safety margins for protection ; I against non-ductile failure. To assure that the radiation emerittlement effects are accounted for in the calculation of the limit curves, the i most limiting value of the nil ductility reference temperature, RTNOT' I' ! used and this includes the radiation induced shift, aRTNOT, corresponding {i to the end of the period for which heatup and cooldown curves are generated, t N iEQUOYAH - UNIT 2 8 3/4 4-8 1 r
, , , . - - > - , - . - - , - . - , - - - - , , , - - _ , -wwm.- , - _ ,. - ~ - , -
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[ 50 FT-LB/35 AVERAGE UPPER SHELF HEA) MATERIAL CU '# NDIT MIL TEMP *F FIL8 COMPONENT NO. GRADE % % 'F PMWD' tMa'02 ND1 PR D' MWF CL Hd. Dome 52899-1 A533BCLI - -
-13 28 48* -12 75" CL Hd. Ring -
A508CL2 - - 5 34 54* 5 125.g* Hd Flange 4890 A50BCL2 - -
-13 <-67* <-67 -13 141. '
Vessel Flange 4832 A508Cl2 - -
-22 -47 -27 -22 155 5 3
Inlet Nozzle 4868 A50BCL2 - -
-22 41 61* 1 79 Inlet Nozzle 4872 A508CL2 - -
-22 12 32* -22 108*a Inlet Nozzle 4877 A508CL2 - -
-31 1 21* -31 113 Inlet Nozzle 4886 A508CL2 - -
-31 -52 -32* 28 138 8
Outlet Nozzle 4867 A508CL2 - -
-31 19 39" -21 ' 85*
cn Outlet Nozzle 4873 A508CL2 - -
-22 21 41* -19 76 ta Outlet Nozzle 4878 A508CL2 - -
-40 -6 14* 105*
8 2 Outlet Nozzle 4887 A50eCL2 - -
-22 -11 9* c 143*5 p
e Upper Shell Inter Shell 4885 4853 A508CL2 A508CL2
-M S 0.13 A+t+ -22 25 19 45*
70 5 10 104 138 93 Lower Shell 4994 A508Cl2 0.14 .- -40 8 38 -22 140 5 3 100 Trans. Ring 4879 A508CL2 - 7., 5 27 47" 5 98 Bot. Hd. Rim 52835-1B A533BCL1 - -
-4 48 68* 8 81*8 Bot. Hd. Rim 52835-18 A533BCLI - -
-22 25 45* -15 81 Bot. Hd. Rim $2899-2 A533BCLI - -
-13 39 59* -1 62* 8 Bot. Hd. 5297-1 A533BCLI - -
-31 14 34* -26 99.5 Weld -
Weld 0.13 M -4 - 14 -4 - 101 liAZ - HA2 - -
-13 -
17 -13 - 120 c.si 5 Paralled to Major Woriing Direction 2 Normal to Major Working Direction
- Estimate based on USAEC Regulatory Standard Review Plan, Section 5.3.2 MIEB 5-2
* % $ hear not reported
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a o REACTOR COOLANT SYSTEM f, BASES __ _ PRESSURE / TEMPERATURE LIMITS (Continued) 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 fini.te heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values 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 such that over the course of the heatup ramp the controlling condition switches f rom the ir. side to the outside and the ,;,ressure limit must at all times be based on analysis of the most critical criterion. 8 Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductfle failure, operating limits are provided to assure compatibility of operation with the fatirjue analysis performed in accordance with the ASME Code requirements. h-3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these compontnts will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Pa-t 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1). . Components of the reactor coolant system were designed prior to the issuance of Section XI of the ASME Boiler and Pressure vessel Code. These components will be tested to the extent practical within the limitations of ; the original plant design, geometry and materials of construction. m ? e b SEQUOYAfi - UNIT 2 8 3/4 4-14
O % INSERT 4 The leak test limit curve shown in Figure 3.4-2 repiesents the minimum temperature requirements at the leak test pressure hpecified t)y applicable codes. The leak test limit curve was determitied by methods of Branch
'lechnical Position MTE!! 5-2 and 10 CFR $0, Appendix L;. ;
The criticality limit curve shown in figure 3.4-2 specifles pressure-temperature lim.(s for core operation to provide additional . margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) require the ; reactor vessel to be at a temperature equal to or higher than the minimum I temperature required for the in-service hydrostatic test, and at least 40 degrees F higher than the minimum pressure-temperature curve for hentup and cooldown. The 'naximum temperature f or the in-service hydrostatic test for the SQN l' nit 2 reactor ve sel is 274 degrees F. A vertical line at 274 degrees F on the pressure-temperature curve, intersecting a curve 40 degrees F higher than the pressure-temperature limit curve, constitutes , the limit for core operation for the reactor vessel.
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i a e . t i I i ENCLOSURE 2 j PROPOSED TECliN! CAL SPECIFICATION CHANGE i h SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 j DOCKET NOS. 50-327 AND 50-328 ! (TVA-SQN-TS-90-01, RI) i DESCRIPTION AND JUST1FICAT10N FOR NEW f REACTOR COOLA'iT SYSTEM PRESSURE-TEMPERATURE LIMITS t i t L h a t 5 i, j h-l l 4 ! I
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I e a ENCL 05URE 2 t t Desrript ion _of Chang [ [ Tennessee Valley Authority proposes to modif y the Sequoyah Nuclear Plant l (SQN) Units 1 and 2 Techni:a1 Specifications (TSs) to incorporate new I reactor coolant system pressure-temperature (P-T) limits (TS Figures 3.4.2 ' and 3.4.3) that are based on Revision 2 to NRC Regulatory Guide (RG) 1.99, ;
" Radiation farbrittlement c 9eactor Vessel Materials." Tb P.ases 3/4.4.9, j Pressure / Temperature Limit has been revised to reflect the aptlication l of RG 1.99, Revision 2. mewnodology. The new P-T limits are applicable up i to 16 effective full power years (ETPY) for both units. f Reason for Change NRC issued Generic Letter (GL) 88-11. "NRC Position on Radiation j Embrittlement of Reactor Vessel Materials and its Impact on Plant t Operations." on July 12 1988. The purpose of GL 88-11 was to call i attention to Revision 2 to RG 1.99. The GL required licensees to perform (
a technical analysis for neutron embrittlement using the revised RG 1.99 i methedology to ensure continued compliante with Appendix G of Title 10 of f the Code of Federal Regulations. Chapter 50 (10 CFR 50). This analysis i has been completed and has resulted in the need to update the P-T limits ! presented in the TSs. ! Justification for__Chan g { t TVA's proposed change utilites the methodology provided in RC 1.99 ; Revision 2 to provide new P-T limits for SQN Units 1 and 2. SQN's new . P-T limits were developed-for TVA by Westinghouse Electric Corporation. [ The P-T calculations for each unit are provided in Enclosure 4 (WC/p--12970 ; for Unit 1 and WCAP-12971 for Unit 2). The calculations are basen on , available surveillance capsule data. 1ne moat limiting value of reference ! nil-ductility temperature (RTNDT) was determined for each material { (intermediate shell torging, lower shell larging, and welds) in the ! reactor vessel beltline region. New P-T.11mits for heatup and cooldown ! were then developed using the most limiting RTNDT (i.e.. highest adjusted j reference' temperature). For Unit I che limiting material was determined ' to be the lower shell forging. For Unit 2 the limiting material was determined to be the weld metal. Samplo calculations for each unit are provided in Table 4 of the respective WCAPs. It should be noted that the new TS figures (Figures 3.4-2 and 3.4-3) provided in Enclosure 1 do include a margin of 60 pounds per square inch and 10 degrees Fahrenheit for instrument -Inaccuracy. Figures that are provided in the Westinghouse WCAPs do not include this margin for instrumentatien errors. Additional changes to the TS bases have beeri included to reflect new values and methods used to upgrado SQN's P-T limits, Information associated with the 10 CFR 50. Appendix C., requirements for evaluating minimum temperatures of SQN's vessel flange and clesure head flange . regions was.added as historical information. This information does :ot l af fect SQN's proposed heatup and couldown liraits.
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1 ( i In conclusion. TVA's proposed TS change establishes new P-T limit curves for i SQN Units 1 and 2. These P-T limits are applicable for the first 16 EfPY f or i both units. The proposed P-T limits were based on RG 1.99 Revidion 2, l methodology in accordance with GL 88-11. l t Envir_opental Ingnet Evaluation f The proposed change request does not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change ; would not: ! i
- 1. Result in a significant increase in any adverse environmental itnpact j previously evaluated in the Final Environniental Statement (FES) as ;
modified by the staft's testimony to the Atomic Safety and Licensing ! Board, supplements to the FES, environmental impact appraisala, or decisions of the Atomic Safety and Licensin$ Borrd. . t
- 2. Result in a significant change in effluent or power levels. l
- 3. Result in matters not previously reviewed in the licensing basis for SQN {
that may have a significant environmental impact. ; e i t f
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t 1 t h l t i f l l t t t
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e o I i ENCLOSURE 3 PROPOSED TECilNICAL SI'ECIFICATION CilANCE ! SEQUOYAll NUCLEAR Pl. ANT UNITS 1 AND 2 f DOCKET N05. $0-327 AND $0-328 ' (IVA-SQN-TS-90-01, R1 ) DETERMINATION OF NO SIGNIFICANT llAZARD CONSIDERATION
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. e ENCLOSURE 3 l
Significant flazard Evaluation } IVA has evaluated the proposed technical specification (TS) change and has (' deternined that it does not represent a significant hazard consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah ! Nuclear Plant (SQN) in accordance with the proposed amendment will nott
- 1. Involve a significant increase in the probability or consequences of !
an accident previously evaluated. The present pressure-ternperature l (P-T) limits for SQN (TS Figures 3.4-2 and 3.4-3) are based on the i methodology described in Westinghouse Electric Corporation l WCAP-7924-A, " Basis f or Heatup and Cooldown Limit Curves," and i American Society of Mechanical Engineers (ASME) Section III, , Appendix 0. These P-1 limits were computed using analytical { projections of neutron embrittlernent to the reactor vessel. The ! current TS P-T limits are applicable for the first 9.2 effective power ! years (ETPY) for Unit 1 and 16 EFPY for Unit 2. i f TVA's revised P-T limits for SQN were computed using the methodology ) described in NRC Regulatory Guide (kG) 1.99 Revision 2, " Radiation l Embrittlement of Reactor . Vessel Materials." TVA's application of the ! Revision 2 methodology resulted in P-T limits applicable to 16 EFPY- i f or both units. The increase in the projected EFPY on Unit 1 la ! because of the calculated decrease in the irradiation damage. There are two primary reasons why the reactor vessel irradiation is t projected to be less than originally predicted by Westinghouse. The ! first reason is the change in criteria associated with the chemistry of the reactor vessel material. The second reason is SQN's j low-leakage core configuration, which reduces the total neutron dose j to SQN's reactor vessel-(this was evidenced by the amount of damage ; measured by SQN's surveillance capsule samples).- ; TVA's modification to SQN's P-T limits complies with the calculative [ procedures and criteria contained la Revision 2 of RG 1.99. The new ; P-T limits for SQN continue to ensure prevention of nonductile reactor l vessel failure. Accordingly, the proposed change does not involve a- r significant increase in the probability er consequences of an accident ! previously evaluated. l
- 2. Create the possibility of a new or different kind of accident from any !
previously analyzed. TVA's proposed change to SQN's P-T limits complies with Generic Letter 88-11. "NRC Position on Radiation i Embrittlement of Reactor Vessel Materials and Its impact on Plant ! Operations," for utilizing the methodology provided in NRC RG 1.99, [ Revision 2. The new P-T limits do not result in a change to the plant' ' configuration. Consequently, the proposed change does not create the l possibility _ of a new or dif ferent kind of accident f rom any previously ; analyzed. } I 3. Involve a significant reduction in a margin of safety. IVA's proposed i TS change to incorporate new P-T limits for SQN remains consistent ! with the methodology provided in RG 1.99, Revision 2. Improvements in ! predicting radiation embrittlement of reactor vessel materials provide f a quantitative basis for detarmining margin of safety. The proposed i change does not reduce the margin of safety. ! I
e o j i l LNCL0st'RE 4 HEATUP AND C00LDOWN LIMIT cl'RVES ; FOR NORMAL Orl'kATION ; SQN l' NITS 1 AND 2 j WCAP-12970 AND 12971 , i r I ( t I 4 P l t I I t l b e o i h e t . , . , . -- - ..n.. , ~, .-,,n. -n.. ,, - , ~.. . . , - , - , - - , . , _ . , - - , - , . . - - . , - - . . ., c.-, .. ,- . . . -, - ,.-- ..,-. . . , _ ,, , . ..n,,}}