Text

Proposed Changes to Tech Specs,Incorporating Improvements Re RCS Pressure Temp Operating Limits
	ML20210C033
Person / Time
Site:	McGuire, Mcguire
Issue date:	09/08/1986
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20210C012	List: ML20210C009; ML20210C033;
References
	TAC-62779, TAC-62780, NUDOCS 8609180220
	Download: ML20210C033 (50)
	v • d • e

{{#Wiki_filter:P' PROPOSED REVISION TO TECHNICAL SPECIFICATIONS AND MENDMENT TO FACILITY OPERATING LICENSE 0609100220 0609009 ADOCK 0500 PDR P

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS itEA_CTORCOOLANTSYSTEM 3.4.9.1 The Reactor Coolant system (except the pressurizer) pressure and temperature shall be maintained within the limits determined by analysis performed in accordance with the requirements of 10CFR 50, Appendix G. TheReactorCoolantSysteg/ hour.(except the Pressurizer) heatup rate and cooldown rate shall not exceed 100 F APPLICA8ILITY: Modes 1,2,3,4,5 and 6 with reactor vessel head installed. ACTION: With any of the above pressure-temperature limits exceeded, restore the pressure and/or temperature to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition l on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STAN08Y within the next 6 hours and reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours. l l SURVEILLANCE REQUIREMENTS l l 4.4.9.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 12 hours and at least l once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. I McGUIRE - UNITS 1 and 2 3/4 4-30 l

EXI3 TING TECHNICAL SPECIFICATION AND BASES " MARK UP" l I l l I l l

ww-.i & A )R COOLANT SYSTEM y [g L u 9 PRESSURE / TEMPERATURE LIMITS g,,, g A WLg 0 t e e A MS ob 1R COOLANT SYSTEM q o C P!t. k'9, A ff u & f>. NG CONDITION FOR OPERATIO i 1 The Reac Coolant System (except the pressurizer) temperature and ire shall be ':fi:2 '- -~ - ": r: :'it th: 't :: :t:r- :- "~'es (,!d' [.~I 222 2 'I 2- ^ '22 !cu,h r G,_ee W $ p./s.m.(.e preeg, & pre sso re ter).ke ddjo ,_ q _ g' '"' Q Q JW "i K'W cd 'L'g T too **/h. w e+ 4.< c.. e - f _; :::,g-.- , innoe 1; y,, n. ^ , -a i-um w,,,p ..... :t: 7: c' ' :: t': : 7 :' '- 1^"" ry 1 ':r-7., '-- i--- h,;.....'.;;..,; '.., '.. r..; t' :,: am. +w. w..,nm ..a s f 7 -- ,1 1 g -.; - ABILITY: Mmr$t=tTR%. pre.ssare.. k m pac k l E N ny cf the aboveg imits exceeded, restore the temperature and/or pressure l hin the limit within 30 minutes; perform an engineering evaluation to Ine th] effects of the out-of-limit condition on the structural integrity R: actor Coolant System; determine that the Reactor Coolant System remains ablo for continued operation or be in at least HOT STANDBY within the hours and reduce the RCS T and pressure to less than 200*F and ig,respectively,withinthP9o110 wing 30 hours, l ** Sb k w [ al: f'" W . LANCE REQUIREMENTS The Reactor Coolant Sy em temperature and pressure shall be ned to be within the limits at least once per 30 minutes during system cooldown, and inservice leak and hydrostatic testing operations. ._"__ 7,%...........;'. - t:ri:' ',---"'-*4aa - u=411*ar= =a-'i-- .T_ n:..'.=.':-~ ':.* nu~s a n -'In, acco raansu g 6ys'~""'"-'""",',.Y'7",'""* n 7 rsyy 6uw 7 Amendment No. Unit 2) - UNITS 1 and 2 3/4 4-30 Amendment No. Unit 1)

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^ e ? E% TABLE 4.4-5 m g REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE Q m a E VESSEL LEAD NUMBER LOCATION FACTOR WITHDRAWAL TIME (EFPY)* N 1. U 4.86 First Refueling 2. V 58.5* 4.05 8 i 3. W 124' 4.86 Standby j 4. X 236* 4.8 4 i R j 5. Y 238.5* 4.05 15 l h 6. Z 304* 4.86 Standby I

Withdrpwal time may be modified to coincide with those refueling outages.or plan hutdowns most closely approaching the withdrawal schedule.

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I i REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) Reducing T,yg tolessthan500'Fpreventsthereieaseofactivityshould a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective ACTION. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. + yk#[ 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Th emperature and pressure changes during heatup and co dow are limited to consistent with the requirements given in the SME Boiler and pe Pressure Vesse ode, Section III, Appendix G: 1. The reactor coo temperature and pressure a system heatup and yd QJ L (Jn cooldown rates (with e exception of the pr surizer) shall be li;ited Mb accc-dence _ith fi;;;..V 2. 4 2 eiE2. AM {.- (gg ;gyu{gg ;;7{gd_ Lfc 96psp::tfiedthmemn Wo 4wof-a. Allowable combinations of p re and temperature for specific / temperature change rates e belo and to the right of the limit lines. d own. Limit I s for cooldo rates between those presented t.u e e A / C WO; may be obtained by terpolation; and (,. 7tG P / b. f!;er:: 2. S2 ..d 2.4 2 defin limits 4 assur revention of 9 / non-ductil allure only. For normal operation, o r inherent { # plant c acteristics, e.g., pump heat addition and p surizer l a 4,* heat capacity, may limit the heatup and cooldcwn rates at can be g ,J<. a eved over certain pressure-temperature ranges. y 2. se limit lines shall be calculated periodically using methods provi ed A b

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e bL L 3. The secondary side of the steam generator must not be pressurized above h 200 psig if the temperature of the steam generator is below 70*F, u.f-4. The pressurizer heatup and cooldown rates shall not exceed 100"t/nr anu Q 200*F/hr, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than rek 320*F, and 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. \\ N unl N k %0.f ~J i McGUIRE - UNITS 1 and 2 8 3/4 4-7 ~~ t ~ t

0 6L Y g.V hy 'y Q REACTOR COOLANT SYSTEM cd c.M BASES ( PRESSURE / TEMPERATURE LIMITS (Continued) j The fracture toughness properties of the ferr,ftic materials in the reactor vessel are determined in accordance with the NRC tandard Review Plan, ASTM E185-73, and in accordance with additional react 6r vessel requirements. These properties are then evaluated in accordance wit 6 Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and, Pressure Vessel Code and the calculation methods described in WCAP-7924-A " Basis for Heatup and Cooldown Limit Curves, April 1975." Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temp,e'rature, RTNOT, at the end of 10 effec-tive full power years (EFPY) of service life. The 10 EFPY service life period is chosen such that the limiting RTNOT,a't the 1/4T location in the core region is greater than the RT f the limit.ing unirradiated material. The selection NOT of such a limiting RT assures tha!all components in the Reactor Coolant NOT System will be operated conservativ ly in accordance with applicable Code requirements. g.g,/ Thereactorvesselmaterials[havebeente,Sted-te-determi heir initial RTNOT; the results of these testjs are shown iQ] :t h " 2 ' O Reactor .g,9, y operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, NOT based upon the fluence, coppe con n, anu p;,v 8'b'Ob Figure S 2.'M' ' 'e coca-t of we mater 1'aT inquestion,canbepredictedusi 7 the largest value of ART cmputedbyeitherRpulato ice 1.w, nevision 1, "Ef fects of NOT Residual Elements on Predigted Radiation Damage to Reactor "-" al Materi E* V8 or the Westinghouse Copper Trend Curves shown in Figur63/4.4-2. and cooldown limit curves /of Fi;;sa 3.4 ^ m,d 0.?g incruce preuicted e heatup ments for this shift in RTNOT #tth;;"d"#QOCPYaswellasadjustmentsfor J possible errors in the pressure and temperature sensing instruments. Values of ART determined in this manner may be used until the results NOT from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in 'ata 4.4-5. The lead factor represents the relationship between the fast neutr n flux density at the location of the capsule and the inner wall of the pres ure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the pressure vess91 m terial by using the lead factor and the withdrawal time of the capsule. The eatup and cooldown curves must be recalculated when the ART determined fr the surveillance capsule exceeds the calculattd ART NOT NDT for the equivalent psule radiation exposure. Allowable pres ure-temperature relationships for various heatup and cool-down rates are calc, lated using methods derived from Appendix G in Section !!I of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A. k te od. na h u/ w e ck. Ts. [6 le. S ~, L McGUIRE - UNITS 1 an R /4 4-8 S

g lABLE B 3/4.4-1 E g REACTOR VESSEL TOUGHNESS (UNIT 1) 50 FT-LB/35 MIN. UPPER SHELF g MIL TEMP (*F) (FT-LB) g PARALLEL TO NORMAL TO PARALLEL TO NORMAL TO RT m CU P NOTT MAJOR WORKING MAJOR WORKING NOT MAJOR WORKING MAJOR WORKING COMPONENT (% ) {%1 (*F) DIRECTION OIRECTION (*F) DIRECTION DIRECTION . +20 77 138* 78 100(b) C1. Hd. Dome j C1. Hd. Seg. +10 27 76* 16 135 88** N C1. Hd. Flange +48$ -50 -5* 48 154 100**' Vessel Flange +40$ -42 -13* 40 153 99"* Inlet Nozzle +60$ +22 44* 60 129 84** i Inlet Nozzle +60$ 41 56* 60 132 86** Inlet Nozzle +60$ 30 61* 60 121 79"* Inlet Nozzle +60$ 40 69* 60 115 75** Outlet Nozzle +60$ -15 24* 60 124 81** Outlet Nozzle +60$ 18 5?* 60 114 74** I Outlet Nozzle +60$ 4 l 46* 60 133 86** i Outlet Nozzle +60$ 72 101* 60 122 79** Upper Shell 0.14 +10 55 108* 48 109(c) (c) L'ppar Shell 0.10 +10 67 106* 46 102 J66 Upper Shell 0.13 0 40 76* 16 141 92** Inter. Shell 0.13 0.010 -30 36 94 34 136 98 l Intar. Shell Q.14 0,011 0 33 60 0 137 102 Ir.ter. Shell 0.11 0.013 -20 12 47 -13 153 103 Lower Shell 0.14 0.009 -1G 36 60 0 125 93 Lowar S he'.1 G.10 0.010 -10 32 90 30 136 113 Lower She' 4 0.10 0.010 0 42 75 15 128 100 Be. Peel Seg. -70 14 45* -15 137 89** 1 Bot. F.rel Seg. -30 25 58* -2 145 94** 5 -- l Bot. Peel Sag. -20 42 87* 27 123 80** vi Bot. Hd. Dome $f 0 50 101* 41 123 80** r H n Y 4 3 h,-{ h i J o ~LT e e

J j FF E TA8tE B 3/4.4-1 (Continued) REACTOR VESSEL TOUGHNESS (UNIT 1) 50 FT-L8/35 i gg MIN. UPPER SHELF MIL TEMP (*F) (FT-LB)

q PARALLEL TO NORMAL TO PARALLEL TO NORMAL TO us CU P

NDTT MAJOR WORKING MAJOR WORKING NOT MAJOR WORKING MAJOR WORKIE RT COMPONENT {%_1 {%) (*F) DIRECTION DIRECTION (*F) OIRECTION DIRECTION >= I l Weld 0.30 ") -60 -6 -50 ^2 Haz 110 -50 28 -32 80 i I

Estimated (77 ft-Ib/54 mil temp. for long data)
- Estimated (65% if longitudinal upper shelf)

(a) Conservative Estimate (b) 95% Shear $ Estimated (60*F or 100 ft-lb temp., whichever is less) (c) 90% Shear ,I t' s. 56 ) O j l i l ,l I -s s -l i m

^ Nn TABLE B 3/4.4-1 (Continued) Eg , REACT 03 VESSEL TOUGHNESS (UNIT 2) MINIMUM 50 FT-L8/ AVENAGE UPPER SHELF c i 5 35 MIL TEMP (*F) (FT-LB) PARALLEL TO NORMAL TO PARALLEL TO NORMAL 10 CU P NOTT MAJOR WORKING MAJOR WORKING NOT MAJOR WORKING MAJOR WORKING RT [ COMPONENT M g) (*F) DIRECTION DIRECTION (b) (*F). DIRECTION DIRECTION (b) i C1. Hd. Dome -31 52 72 12 132 86 C1. Hd. Ring 16 3 23 16 156 101.5 Hd. Flange -13 41 61 1 155 100.5 Vessel Flange -4 21 41 -4 174 113 Inlet Nozzle -13 -4 16 -13 141 92 Inlet Nozzle -31 3 23 -31 114.5 74.5 Inlet Nozzle -22 -8 12 -22 129 84 a> Inlet Nozzle -40 -6 14 -40 132 86 Outlet Nozzle i w -13 33 53 -7 124 81 A Cutlet Nozzle -40 16 36 -24 103(c) 67(c) !t Outlet Nozzle -49 24 1 44 -16 116(c) 75.5(c) l - Outlet Nozzle -40 10 30 -30 121(c) 78.5 (c) Upper Shell -4 41 61 1 157.5 102.5 Inter. Shell 0.16 0.012 -4 7 37(a) - 4(a) # 147 96(a) Lower Shell 0.15 0.004 -30 -22 7(a) -30(a) 152 141(a) j 80t. Hd. Ring -4 55 75 15 109(c) 71(c) Bot. Hd. Peel -49 38 58 -2 136 Bot. Hd. Peel 88.5 -40 -24 -4 -40 131 85 Bot. Hd. Peel -13 -18 2 -13 142 92 Bot. Hd. Peel -13 -8 12 -13 132 86 Bot. Hd. Dome -40 -12 8 -40 127 82.5 Weld (Inter / Lower) 0.05 0.010 -76 NA - 8(a) -68(a) N.A. 128(a) HAZ 1 -76 NA -58(a) -76(a) N.A. 125.5(a) 4 T,,, p (a) Based on Actual Data h g A C, f 7 (b) Estimated Per Branch Technical Position MTE8 5-2 !l \\ (c) 100% Shead not reached; Upper Si. elf Toughness will be greater than that listed. @I A y gtg t

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40 v-REACTOR COOLANT SYSTEM b N 0 ( BASES PRESSURE / TEMPERATURE LIMITS (Continued) The general method for calculating heatup and cooldown limit curves is 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 dimensions of this postulated crack, referred to in Appendix G of ASME Section III 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 against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most -limiting value of the nil-ductility reference temperature, RTNOT, is used and this includes the radiation-induced shift, ARTNOT, corresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that-the total stress intensity factor, K ob,cooldown cannot be greater than the reference stress intensity fact for the metal temperatufe at that time. K is btained from the reference IR' IR fracture toughness curve, defined in Appendix G to the ASMEsCode. The K curve is given.by the equation: IR Kgg = 26.78 + 1.223 exp [0.0145(T-RTNOT + 160)] ' (1) Where: K is the reference stress intensity factor as a function of the metal temperatuf!Tandthemetalnil-ductilityreferencetemperatureRT

Thus, NOT.

the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKyg + kit 1 IR (2)- Where: KIM is the stress intensity factor caused by membrane (pressure}

stress, It is the stress intensity factor caused by the thermal gradients, K

K is provided by the code as a function of temperature relative tfRthe RT f the material, NOT C = 2.0 for level A and 8 service limits, and C = 1.5 for inservice hydrostatic and leak test operations. At any time during the heautp or cooldown transient, K is determined by_ the metal temperature at the tip of the postulated flaw, thhpappropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses f McGUIRE - UNITS 1 and 2 8 3/4 4-14

h.y y f ')

v REACTOR COOLANT SYSTEM

( BASES PRESSURE / TEMPERATURE LIMITS (Continued) resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K for the reference flaw is computed. FromEquation(2)thepressureshe,ssintensity factors are obtained and, from these, the allowable pressures are calculated. C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because-the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting _ pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the i' vessel ID. This condition, of course, is not true for the steady-state situation. It follows that at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value of K at the 1/4T IR location for finite cooldown rates than for steady-state operation. Further-more, if conditions exist such that the increase in K exceeds Kg, the IR calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated.if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. HEATUP Tnree separate calculations are required to d'etermine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions ) as well as finite heatup rate conditions assuming the presence of a 1/4T l defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pre,sure. The metal temperature at the. crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during heatup is lower than the K forthe1/4Tcrackdhingsteady-state IR conditions at the same coolant temperature. During heatup, especially at the McGUIRE - UNITS 1 and 2 B 3/4 4-15

w b f REACTOR COOLANT SYSTEM ( BASES PRESSURE / TEMPERATURE LIMITS (Continued) end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K 's for steady-state and finite heatup rates IR do not offset each other and the pressure-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 heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of 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 heatu;) 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. 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-f point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the j 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 from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the composite curves for the heatup rate data and the cooldown j rate data are adjusted for possible errors in the pressure and temperature f sensing instruments by the values indicated on the respective curves. S b Although the pressurizer operates in temperature ranges above those for h which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis W performed in accordance with the ASME Code requirements. N The OPERABILITY of two PORVs or an RCS vent opening of at least 4.5 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 A i McGUIRE - UNITS 1 and 2 B 3/4 4-16

REACTOR COOLANT SYSTEM l BASES p // PRESSURf7 TEMPERATURE LIMITS (Continued) ( N the RCS cold legs ard'less than or equal to 300*F. Either PORV has adequate- - J relieving capability to pro'tect-tite RCS from overpressurization when the transient is limited to either: (1).~the-sstart of an idle RCP with the secondary I water temperature of the steam generator less than-or_ equal to 50*F above the RCS cold leg temperatures, or (2) the start of a HPSI pump ~and.its injection into. a-water-solid RCS. 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 components will be maintained at an acceptable levsl 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 Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i). / Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME. I Boiler and Pressure Vessel Code, 1971 Edition and Addenda through Winter 1972. ? 6 y S. McGUIRE - UNITS 1 and 2 8 3/4 4-17

PROPOSED REVISION TO THE BASES OF TECHNICAL SPECIFICATIONS 1

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) Reducing T,yg to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure.of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance bquirements provide adequate assurance that excessive scocific activity levels in the reactor coolant will be detected in sufficient time to take corrective ACTION. Information obtained on iodine spiking will be used te assess the parameters associated with spiking phenomena. A reductien in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE-LIMITS REACTOR COOLANT SYSTEM LCO: The LC0 is provided to establish a licensing tie between the applicable regulation, 10 CFR 50, Appendix G fracture Toughness Requirements, and the NRC approved action statement that provides operational requirements in the event the regulatory requirements are violated. It is wholly acknowledged that this LC0 is redundant to the requirements of the operating license which requires compliance with applicable regulations. The presure-temperature limits are derived from analysis performed in accordance with 10 CFR 50, Appendix G. This analysis utilizes the results of examinations of reactor vessel material irradiation surveillance specimens that are removed from the reactor vessel cavity on a schedule, established pursuant to 10 CFR 50, Appendix H. The approved swveillance capsule withdrawal schedule is located in McGuire FSAR Chapter 5. System preservice hydrotests and inservice. leak and hydrotests are to be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The limit of 100 F/ hour is the rate of change of temperature and not the observed total changes over a 1 hour period. It is a value that has been assumed in the analysis to develop the pressure temperature curves. APPLICABILITY: Applicability assures compliance with regulations. The requirements must be met during all modes of-operation including heatup, cooldown, in service leak and hydrostatic test operations. Heatup, cooldown, in service leak and hydrostatic test operations are considered Condition I (Normal Operation and Operational Transients). Condition I occurrences are those which are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, Condition I occurrences are accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action. These pressure-temperature limits are also applicable during steady-state operation. Temperature and pressure shall be verified to be within their limits at least once per 12 hours. McGUIRE - UNITS 1 and 2 B 3/4 4-7

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The cooldown limit curves are not applicable to conditions of off-normal operation (e.g., small LOCA and extended loss of feedwater) where cooling is achieved for extended periods of time by circulation of water from ECCS through the core. If core cooling is restricted to meet the cooldown limits under other than normal operation, core integrity could be jeopardized. ACTION STATEMENT: The action statement is consistent with that in Standard Technical Specifications. The actions are based on reasonable engineering judgement as to appropriate action to be taken in the event a pressure-temperature limit is exceeded. SURVEILLANCE 4.4.9.1 This surveillance provides the means by which compliance with the LCO is achieved. REFERENCES - McGuire FSAR, Section 5.2 McGuire FSAR, Section 15.0.1.1 RV Capsule Report, McGuire 1, WCAP 10786, February 1985 RV Capsule Report, McGuire 2, WCAP 11029, January 1986 PRESSURIZER 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 spray fluid is greater than 320 F. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are providad to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. OVER PRESSURE PROTECTION SYSTEMS The OPERABILITY of two PORVs or an RCS vent opening of at least 4.5 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 300 F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when i De 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 temperatures, or (2) the start of a HPSI pump and its injection into a water-solid RCS. McGUIRE - UNITS 1 and 2 B 3/4 4-8

Pages B 3/4-9 to B 3/4-16 Deleted i,i McGUIRE - UNITS 1 and 2 8 3/4 4-17

McGUIRE FINAL SAFETY ANALYSIS REPORT UPDATE l i

d-Section III of the 1971 ASME Boiler and Pressure Vessel Code,.plus applicable Addenda and Code Cases to Winter 1971. The location and orientation of the impact specimens for-Units 1 and 2 are in accordance with paragraph NB-2300 of ASME Section III. In addition, the reactor vessel materials meet the fracture toughness requirements of 10CFR50, Appendix G, to the extent possible. The j pressure-temperature limitations on reactor operation, as.well as leak and 4 hydrostatic test. conditions.are determined in accordance with Appendix G to Section III of the ASME B&PV Code and Appendix G, 10CFR50. Since the fracture toughness testing performed on vessel material from Units 1 and 2 did not in-i clude all of the tests necessary to determine RT in the manner prescribed NDT in NB-2300 of ASME III, Summer 1972 Addenda, the necessary properties were estimated using the procedures provided in Branch Technical Position MTEB 5-2, " Fracture Toughness Requirements for Older Plants." A summary of the fracture toughness data for the Unit 1 and Unit 2 reactor j pressure vessel material are given in Tables 5.2.4-1 and 5.2.4-2. i 5.2.4.2 Acceptable Fracture Energy Levels 4 Initial upper shelf fracture energy levels for the materials of the reactor vessel beltline (including welds) as determined by Charpy V-notch tests on specimens oriented in the " weak" direction of the material will be established for the vessel irradiation surveillance test programs for Units 1 and 2. No initial upper shelf energy criteria has been established for the beltline l materials. The Charpy V-notch test data for Units 1 and 2 are in Tables 5.2.4-3 through 5.2.4-13. 5.2.4.3 Operating Limitations During Startup and Shutdown The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan,. ASTM i E185-73 and in accordance with other requirements discussed in Section i 5.2.4.1. These properties are then evaluated in accordance with appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and methods described in WCAP-7924-A (ref. 6) to_ derive the heatup 1 l and cooldown restrictions for the reactor pressure vessel. The calculation of ( allowable pressure temperature relationships for various temperature heatup and cooldown rates is discussed in detail in WCAP-7924-A. 2 i The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect j with a depth of one quarter of the wall thickness, T, and a length of 3/2 T is assumed to exist at the inside of the vessel wall as well as at the outside .of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current detection capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conser-i vative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, R1 is used and this includes NDT,crresponU!T,totheendoftheperiod j the radiation-induced shift, ART ag for which heatup and cooldown curves are generated. 5.2-32 j McGUIRE - UNITS I and 2 Update ..,.-n c.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, l K, for the combined thermal and pressure stresses at any time during heatup 7 or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. K is obtained from the reference fracturetoughnesscurve,definedinAppenkRtx G to the ASME Code. The KIR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.0145(T-RTET + 160)] (1) Where: K is the reference stress intensity factor as a function of the IR metal temperature T and the metal nil-ductility reference temperature RTET

Thus, the governing equation for the heatup-cooldown analysis is defined in

- Appendix G of the ASME Code as follows: CK73 + kit I IR (2) Where: K is the stress intensity factor caused by membrane (pressure) yg

stress, K

is the stress intensity factor caused by the thermal gradients, It i K is provided by the code as a function of temperature relative to tk!RT f the material, ET C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations. At any time during the heatup or cooldown transient, K is determined by the metaltemperatureatthetipofthepostulatedflaw,tkpe appropriate value for RT and the reference fracture toughness curve. The thermal stresses resulth,from temperature gradients through the vessel wall.are calculated and then the corresponding thermal stress intensity factor, K FromEquation(2)thepressureskT,frthe reference flaw is computed. ress intensity factors are obtained and, from these, the allowable pressures are calculated. l The heatup and cooldown curves for Unit 1 and Unit 2 are based on the material properties given in Tables 5.2.4-1 and 5.2.4-2. These curves define the allowable pressure at the actual indicated temperature as a function of the rate of temperature change. Allowances for instrument error in measurement of temperature and pressure are incorporated into the curves in Figures 5.2.4-1, 5.2.4-2, 5.2.4-3, and 5.2.4-4. Initially the Reference Nil Ductility Temperatures (RTPredictba)RT for McGuire Unit 1 and Unit 2 reactor vessels were 0*F and -4'F. values were derived for the 1/4 T and 3/4 T (one quarter and three quart b of vessel thickness) locations in the limiting core region material by using Figures 5.2.4-5 and 5.2.4-6. A service period of 10 Effective Full Power Years (EFPY) was selected for unit I reactor vessel and 8 EFPY for the unit to reactor vessel. 5.2-32a McGUIRE - UNITS 1 and 2 Update

The limiting material for unit I reactor vessel at the 1/4 T and 3/4 T locations is the core region weld metal which has initial RT f 0 F and ET an estimated copper content of 0.30 percent. The limiting material for unit 2 at 1/4 T and 3/4 T location is the intermediate shell which has an initial RT f -4 F and an estimated copper content of ET -0.16 percent. i i l 1 i i i I 4 l 5.2-32b i McGUIRE - UNITS 1 and 2 Update l

_ _ _ _. _ _ ~. _ _ _ A conservative margin of 10*F and 60 psig is used in heatup and cooldown curves for possible instrumentation error. The heatup and cooldown curves in Figures 5.2.4-1 though 5.2.4-4 include the estimated, ART from Figures 5.2.4-6 for the fluence that corresponds to the -selected servik period. 4 J The'results of the reactor vessel radiation surveillance program are used to verify that the ART Predicted from Figure 5.2.4-4 is appropriate, ET and to make any changes necessary to correct Figure 5.2.4-4 if the ART l determinedfromthe'surveillancecapsulesisdifferentfromthepredicN i ART Analysis of the reactor vessel radiation surveillance data as well-i as M a.ils of the heatup cooldown calculations are included in appendix j ' reference 7 of this chapter. i The use of an RT that includes a ART to account for radiation effects j onthecoreregigTmaterial, automatica h provides additional conservatism l lL for the non-irradiated regions. Therefore, the steam generators, pressurizer, flanges, nozzles, and other regions not affected by radiation are favored by additional conservatism approximately equal to the assumed ART ET

5.2.4.4 Compliance with Reactor Vessel Material Surveillance Program Requirements Changes in fracture toughness of the core region plates, forgings, weldments, and associated heat affected zones due to radiation damage are monitored by a surveillance program which conforms with ASTM E-185-73, Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels. The elevation of the ra-4 diation damage in this surveillance program is based on pre-irradiation and i

post-irradiation' testing by Charpy V-notch, and tensile specimens and post-irradiation testing of 1/2T compact tension specimens carried out during the lifetime of the reactor vessel. Specimens are irradiated in capsules located near the core midheight and removable from the vessel at specified intervals. For additional details of the irradiation surveillance program refer to j Subdivision 5.4.3.7 and reference 7 of this chapter. 5.2.4.5 Reactor Vessel Annealing 4 See Subdivision 5.4.3.8 for a discussion of reactor vessel annealing. 5.2.5 AUSTENITIC STAINLESS STEEL The unstabilized austenitic stainless steel material specifications used for the (1) Reactor Coolant System Boundary, (2) systems required for reactor shutdown, and (3) system required for emergency core cooling are listed in i Tables 5.2.3-1 and 5.2.3-2. l The unstabilized austenitic stainless steel material specifications used for the reactor vessel internals which are required for emergency core cooling for any mode of normal operation or under postulated accident conditions, and for l core structural load bearing members are listed in Table 5.2.3-3. All of the above tabulated materials are procured in accordance with the L specification requirements and include special requirements of the applicable ASME Code Rules. 5.2-33 Update

5.

2.9 REFERENCES

1. ~ Cloud, R. L., Pipe Breaks for the LOCA Analysis of the Westinghouse [ Primary Coolant Loop, WCAP-8172, July,-1973. i 2. Flachsbart, B. B., and Logcher, R. D.,," ICES STRUDL-II, The Structural Design Language Frame Analysis," MIT-ICES-R68-91, Vol. 1, November, 1968. 3.

Bordelon, F., and Nahavandl, R., A Space-Dependent Loss of Coolant i

Accident and Transient Analysis for PWR System (SATAN Digital Computer Code), WCAP-7845, January, 1972. 4. Cooper, K., Miselis, V., and Starek, R. M., Overpressure Protection for Westinghouse Pressurized Water Reactors, WCAP-7769, Revision 1, June 1, 1972. 5. Nay, J. A., Process Instrumentation for Westinghouse Nuclear Steam Supply Systems, WCAP-7671, April, 1971. 6. Anderson, S. L., Hazelton, W. S., and Yanichko, E. E., Basis for Heatup and Cooldown Limit Curves, WCAP-7924, August, 1972. Basis for Heatup and Cooldown Limit Curves, WCAP-7924-A, April 1975. i-7. S. E. Yanichko, T. V. Congedo, W. T.' Kaiser, Analysis of Capsule U 'i from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation l Surveillance Program, WCAP-10786, February 1985. S. E. Yanichko, T. V. Congedo, W. T. Kaiser, Analysis of Capsule V from Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance i Program, WCAP-11029, January 1986, t 8. Shabbits, W. O., Dynamic Fracture Toughness Properties of Heavy Section A 533 Grade B Class 1 Steel Plate, WCAP-7623, December, 1970. i 9. Szy Slow Ski, J. J., and Salvatori, R., Determination of Design Pipe l Breaks for the Westinghouse Reactor Coolant System, WCAP-7503, Rev. 1, February, 1972. 10. Enrietlo, J. F., Control of Delta Ferrite in Austenitic Stainless Steel Weldsents, WCAP-8324, May, 1974. 11. Golik, M. A., Sensitized Stainless Steel in Westinghouse PWR Nuclear l Steam Supply Systems, WCAP-7477-L, (proprietary), March, 1970. I 12. Hazelton, W. S., Addendum 1 to Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems, WCAP-7477-L, Add. 1, (proprietary), j May, 1971. 13. Hazelton, W. S., Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems, WCAP-7735, August, 1971. 14. McGuire Nuclear Station Inservice Inspection Plan, Revision 6, March 6, l 1984. I 5.2-49 l Update

Specimens oriented parallel to the principal rolling direction of plate for Unit 1 or major working direction for forging 'cr Unit 2.

- Specimens oriented normal to the principal rolling direction of plate for Unit 1 or major working direction of forging for. Unit 2.

The following dosimeters and thermal monitors are included in each of the six capsules: Dosimeters Iron Copper Nickel Cobalt-Aluminum (0.15% Co) Cobalt-Aluminum (Cadmium shielded) U-238 (Cadmium shielded) Np-237 (Cadmium shielded) Thermal Monitors 97.5% Pb, 2.5% Ag (579*F Melting Point) 97.5% Pb, 1.75% Ag, 0.75% Sn (590*F Melting Point) The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall with the specimens being located between the core and the vessel. Since these specimens experience accelerated exposure and are actual samples from the materials used in the vessel, the transition temperature shift measurements are representative of the vessel at a later time in life. Data from CT fracture toughness specimens are expected to pro-vide additional information for use in determining allowable stresses for irradiated material. 19 The calculated maximum fast neutron exposure at the vessel wall is 2.1 x 10 2 n/cm >1 Mev. The reactor vessel surveillance capsules are located at 56* and 58.5 as shown in Figure 5.4.3-3. The relative exposures of the capsule and the vessel maximums are listed below: Lead Vessel Maximum Capsules at By a multiplying factor of 56* (u,w,c,z) 4.86 58.5* (v,y) 4.05 Correlations between the calculations and the measueements on the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Subdivision 5.4.3.7.1 of this FSAR and have indicated good. agreement. The calculations of the integrated flux at the vessel wall are conservative. The anticipated degree to which the specimens perturb the fast neutron flux and energy distribution is considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure is made of use of data on all capsules withdrawn. The tentative schedule developed prior to licensing for removal of the capsules for post-irradiation testing was as follows: 5.4-9 Update

Capsule U-First Refueling Capsule X-4 effective full power years (EFPY) Capsule V-8 EFPY-Capsule Y-15 EFPY Extra capsules for complimentary testing or additional exposure The present schedules for capsule withdrawal are provided in Table 5.4.3-1. 5.4.3.7.1 Measurement of Integrated Fast Neutron (E >'1.0 MEV) Flux at the Irradiation Samples Information on the spectrum of neutron fluxes at the location of the irradiation samples is obtained from the multigroup diffusion code PIMG(1) Dosimeters including U-238, Np-237, Co-A1, Cu, N1, Cd shielded Co-A1, and Fe are contained in the capsule assemblies. Fe (n,p) 54Mn reaction 54 The procedure for measuring fast neutron flux by the is described below. The measurement technique for the other dosin.eters, which are sensitive to different portions of the neutron spectrum, is similar. The 54Mn product of this reaction has a half life of 314 days and_ emitis gamma rays of 0.84 Mev energy which are easily detected using a Nal scintillator. 54 In irradiated steel samples, chemical separation of the Mn may be performed to ensure freedom from interfering activities. This separation is simple and 54 very effective, yielding sources of very pure Mn activity. In some samples all the interferences may_be corrected for by the gamma spectrometric methods without any chemical separation. The count data is used to give the specific 54 activity of Mn per gram of iron. Because of the relatively long half life of 54Mn, the flux may be calculated for irradiation periods up to about two years. Beyond this time the dosimeter begins to reflect the later integrated 54Mn power output. The burnout of the neutrons em" produced is not significant until the 3 ~1 thermal flux is about 1014 sec The analysis of the sample requires that two steps are completed: first the measurement of 54Mn disintegration rate per unit mass of sample and second the measurement of iron content of the sample. Having completed these analyses, the calculation of the flux is as follows: For an irradiation the activity of any activation product (A) is given: A = $ a N (1 - e 'd) a Where $ = the neutron flux, n/cm -see o = effective cross-section, barns N = number of target atoms ,1-A = decay constant of produce, sec tg = irradiation time, see td = decay time from end of irradiation to counting time, see Then for a power reactor operating at various power levels over scme period, we allow for flux changes by dividing the exposure period into several parts and normalizing the flux in each part as that fraction of full power represented. Then for t periods: 5.4-10 Update

t -At. Atd

)'

A = $,ON I (1-e n e 9p 1 2 Where $,= flux at maximum power, n/cm -sec th 'i, = cooling time for end of n period, see i D th d = cooling time for end of n period, see Fn = flux normalizing factor which is O actual power output in n period maximum possible in n period If now we write 55 $,ON = c I $pggg (E, r)

o, (E) p 1

When E is the neutron energy and r is the radial distance from core center line. The right hand side of the above equation is g sum of the products of PIMG fluxes and the 54Fe (n,p) 54Mn cross section averaged over the PIMG energy groups, then the measured neutron flux (E > 1 Mev) is given by: 10 Mev $ (E > 1 Mev) = C I $PIMG ( '#) E=1.0 where C is a constan: The error involved in the measurement of the specific activity of the dosimeter after irradiation is estimated to be 15 percent. Recently Westinghouse reanalysed the integrated fast neutron flux calculations for the McGuire Units 1 and 2. The reanalysis uses discrete ordinates S transport analysis to determine the fast (E > 1.0 Mev) neutron flux and" fluence as well as the neutron energy spectra within the reactor vessel and the surveillance capsules. The spectrum-averaged reaction cross sections derived from the reanalysis are used to analyze the surveillance capsules. Details of this analysis are contained in reference 7. The only difference in the methodology is to use multigroup transport computer codes to more accurately model the neutron environment at the capsule location. 5.4.3.7.2 Calculation of Integrated Fast Neutron (E > 1.0 MEV) Flux at the Irradiation Samples The method to be described herein is an approximation to the ideal three-dimensional neutron transport solution, but correlations between its predictions and measurements on samples irradiated in Yankee and Saxton cores indicate good agreement. 5.4-11 McGUIRE - UNITS I and 2 Update

The spectrum of neutro.1 fluxes at the capsule onedimensionalmultigroupdiffusioncodePIMGgationisobtainedfromthe for the array of annular shields surrounding a cylindrical core of infinite height. The cylindrical core has a cross-sectional area equal to that of the actt.a1 core. The radial source distribution chosen for the core represents the expected average over the life of the station. The magnitude of the neutron fluxes generated by the PIMG code, which does not treat transport effects, is adjusted by application of a spatial correction factor. T factor is the ratio of the fast neutron doseratecalculatedbytheSPIC-1gry code for an all water medium surrounding a typical Westinghouse PWR to the fast neutron dose rate obtained by PIMG in the identical geometry. The SPIC-1 fast neutron dose rate calculation uses an empirical fast neutron attenuation kernel in the form of a linear combination of a single exponentials which are fitted to the experimental fast neutron dose rate distribution in pure water. The axial and aximuthal variations of neutron flux at the capsule location are determined separately. Theaxialdistributionisexgssedastheratioof the normalized results of two calculations using PDQ , a two dimensional 4 group (r,z) diffusion code. In the first of these an infinitely high equiva-lent cylindrical core with a fission neutron source strength S, per unit i height is surrounded by an all water medium containing the capsule location. In the second, the finite height is surrounded by an all water medium. The fixed source option of the PDQ4 code is selected so that the axial variation of source strength in the core represents a good approximation to the average over the core life. The radial distribution is identical to that chosen for PIMG. The ratio, $(E,r,z)F x 8 1 S $(E,r); p Where subscripts F and 1 denote finite and infinite core representations respectively, is the required axial correction term. The aximuthal distributions of neutron fluxes at the same location are derived PDQ3gcomparisonoftheresultsofthetwodimensional4groupg,y) code from and the one dimensional 4 group diffusion program AIM-5 In the PDQ3 calculation the core, whose shape can be specified exactly, is surrounded by an all water medium. The radial and aximuthal source distributions in the core are both reasonable approximations to the averages expected during the core life. The radial source distribution in the AIM-5 calculation, in which the equivalent cylindrical corn is surrounded by an all water medium, is identical to that chosen for rigg, The product of, 1. The spatially corrected PIMG results, 2. Axial correction ter.n, and 3. Azimuthal correctica term, defines the three dimensional verification of neutron flux at the sample locations. 5.4-12 Update

The technique indicated above overpredicts Saxton measurements by 30 percent and the Yankee measured values by 14 percent. In both reactors the measured results are averages for a set of specimens in a capsule located outside the thermal shield opposite a core corner. More recently, resitlts from SELNI specimens were overpredicted by 10 percent. The reported technique also gives excellent agreement with measured data reported for the PM2A reactor. Based on the above evidence, it is concluded that the PIMG calculation, corrected as described, conservative by approxi-mately 20 percent. Therefore, calculated fluxes at the vessel wall are reliable. To accurately determine the neutron environment at the capsule locations Westinghouse reanalyzed the calculations for neutron flux and energy spectra using a two dimentional transport computer code. The effect of the surveil-lance capsule structures has also been modeled in these new calculations. DOT a two dimentional discrete ordinate transport code was utilized for this analysis. The cross sections used in this analysis were obtained from SAILOR cross section library which was developed specifically for light water reactor applications. For this new analysis, power distributions representative of time averaged conditions, derived from statistical studies of long term opera-tion of Westinghouse 4-loop plants were employed. The use of these generic core power distributions and SAILOR cross section data yields conservative results with calculations exceeding measurements by 10 to 25%. Details of the l calculations along with results of testing can be found in Reference 7. 1 ] i 4 5.4-12a Update i ~ ..-...._.---.._m._-_._ .e_.-. -m .___-._v- _n,

1. Shop ultrasonic examinations are performed on all internally clad surfaces to_an acceptance and repair standards to assure an adequate cladding bond to allow later ultrasonic testing of the base metal from inside. surface. The size of cladding bonding defect allowed is 3/4 of an inch in diameter. 2. The design of the reactor vessel shell in the core is a clean, uncluttered cylindrical surface to permit future positioning of the test equipment without obstruction. 3. After the shop hydrostatic testing, selected areas of the reactor vessel are ultrasonic tested and mapped to facilitate the inservice inspection program. Vessel design data are in Table 5.4.2-1. Transients and anticipated number of cycles are in Table 5.2.1-1. The vessel fabricator quality surveillance information is in Table 5.4.4-1. 5.

4.5 REFERENCES

1.

Bohl, H., Jr., et al., PIMG--A One-Dimensional Multigroup P Code for the y

IBM-704, WAPD-TM-135, 1959. 2.

Shure, K., " Radiation Damage Exposure and Embrittlement of Reactor Pressure Vessels," Nucl. Appl, 2, 106-115 (April 1966).

3.

Gillis, P., SPIC-1, An IBM-704 code to calculate the Neutron Distribution outside a right-circular cylindrical source. WAPD-TM-196 (1959).

4. Cadwell, W. R., PDQ-4--A Program for the Solution of the Neutron-Diffusion Equations in Two Dimensions on the Philco-2000, WAPD-TM-230 (1961). 5. Cadwell, W. R., PDQ-3--A Program for the Solution of the Neutron-Diffusion Equations in Two Dimensions on the IBM-704, WAPD-TM-179 (May 1960). 6. Flatt, H. P., and Baller, D. C., AIM-5--A Multigroup, One-Dimensional Diffusion Equation Code, NAA-SR-4694, March, 1960. 7. S. E. Yanichko, T. V. Congedo, W. T. Kaiser, Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-10786, February 1985. S. E. Yanichko, T. V. Congedo, W. T. Kaiser, Analysis of Capsule V from Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program, WCAP-11029, January 1986. 4 5.4-15 Update

TABLE 5.4.3.1 SURVEILLANCE CAPSULE REMOVAL SCHEDULE (Ref. A and B) The following removal schedule is recommended for future capsules to be removed from the McGuire Unit 1 and Unit 2 reactor vessel. i Lead Removal Estimated Fluence Capsule Factor Time [a] (n/cm2 x 1919) Unit 1 U 4.76 Removed (1.06) 0.414 (Actual) X 4.76 4 1.80[b] V 4.06 8

3. 07.I Cl Y

4.06 15 5.75 I W 4.76 Standby Z 4.76 Standby I Unit 2 V 4.06 Removed (1.03) 0.306 (Actual) X 4.76 4 1.72 I ICl U 4.76 7 3.01 Y 4.06 15 5.50[d] W 4.76 Standby Z 4.76 Standby a. EFPY from plant startup l b. Approximates vessel end of life 1/4 thickness wall location fluence c. Approximates vessel end of life inner wall location fluence d Approximates vessel inner wall location fluence for plant life extension to 60 EFPY. Should be adjusted accordingly if plant life is extended to another t'ime period. References. A. WCAP-10786, Analysis of Capsule U from the Duke Power Company McGuire Unit 1 Reactor Vessel Radiation Surveillance Program, February, 1985 by S. E. Yanichko, T. V. Congelo and W. T. Kaiser. B. WCAP-11029, Analysis of Capsule V from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program, January, 1986 by S. E. Yanichko, T. V. Congelo and W. T. Kaiser. l l New Table j Update - _ =.

~I mus.s. 2..p --i I -.n,- MCGulRE UNil I NE ACIOR VESSEL 100GINES$ I ABLE ( W u> c df - 1078g ) Matertal Spec 6fIcatton Code Cu P Hl T Ri USE HDI gg Component Pksmber Number (%) (%) (%) (*F) (*F) (ft-lb) Closure head dome A5338CL.I D5006-1 0.81 0.080 0.48 20 37 ICI ICI 65 Closure head segment 5 A5338CL.I B5087 O.1I O.000 0.62 . 10 10[cj gg[c) ICI ECI ICI Closure head flange A508CL.2 85002 0.0l0 0.75 40 40 10l ECI 29[cl ICI vessel flange A508CL.2 84708 0.010 0.73 2G IOl ICI ICI ICI Inlet rx2zzle A508CL 2 05003-1 0.42 0.010 0.68 60 60 89 ICI ICI ICI Inlet ruiz z le A508CL.2 B5003-2 0.10 0.012 0.78 60 60 88 ICI 60[c] 79[c] Inlet rm zle A508CL.2 05003-3 0.10 0.009 0.69 60 ICI ICI 77[cl Inlet nozzle A508CL.2 05003-4 0.10 0.010 0.69 60 60 ICI ICI ICI Outlet s wiz z le A508CL.2 05004-1 0.005 0.74 60 60 8'2 !CI ICI ICI butlet swizz le A500CL 2 05004-2 0.007 0.74 60 60 75 ICI ICI ICI Outlet nozzle A508CL.2 D5004-3 0.005 0.75 60 60 90 ICI 60[c) 81[c] Outlet swizzle A508CL.2 05004-4 0.006 0.79 60 ICI ICI upper shell A533BCL.I 85453-2 O.14 0.011 0.58 10 15 73 Upper shell A533BCL.t 05041-2 O.10 0.011 0.54 IO 27 ICI ICI 68 ICI ICI Upper shell A533BCL.I 850ll-3 0.13 0.010 0.56 0 O ,95 Intermediate sheII A5338CL.I B5012-1 0.13 0.010 0.60 -30 34 101 Intermediate shelI A533BCL.t 05082-2 O.83 0.011 0.62 O O 104.5 Interme.liate shell A5330CL.I 85012-3 0.10 0.013 0.66 -20 -13 109 Lower shell A533BCL.1 05013-1 0.14 0.009 0.56 -10 O 94 tower shell A533BCL.I D5083-2 0.10 0.010 0.52 -10 30 315 tower shell A533BCL.I H5013-3 0.10 0.040 0.55 O 15 104 Dottom swad segment A533ttCL.1 0545t1 - I O.14 0.014 0.60 -70 -26 tc] ICI 90 tk2t tom head segment A5330CL.I b5458-2 O.IS 0.014 0.54 -30 -15lc] ICI D6 ICI ECI Dottom head segment A533BCL.1 05458-3 0.13 0.012 0.56 -20 2 82 ICI ICI t$ottom head dome A5330CL.I B5085-1 O.13 O.010 0.53 0 10 79 Intermediate shell longi t ud ina l weld seams Mt.22lal O.21 O.Oli 0.88 -60 -50 >llO Intermeal. ate shell to lower shell weld GI.39 0.05 0.006 -70 -70 >I26 ICI ICI tower shell longitudinal weld seams Mt.32 0.20 0.015 O O 90 IDI ICI ICI tower sheril longitudinal weld seams MI.33 O.21 0.016 0.66 O O th] ICI ICI tower shell longitudinal weld seams M t. :44 O.30 0.013 0.64 O O a. Used reactor vessel surveillance weldment is. Used in weld root region only IstHnated per U.S. NRC Standard Review Plan!II C. ~ __ f f

'~' = TA S L E )~ 2 4 .'2, Taste =e=* flCGul8E UNii 2 SEACIOS VE5SEL TOUGHNESS TASU ~ Material ___ Uooer shelf EneCgy_ SpectfIcatton Heat Cu P Ml Tway 8TNDT 88d8 NMW8 Component Number Number (5) (5) (5) (*f) (*F)(a) (ft-lb) (ft-lb) Closure head done A5338. C1. 1 55154-1 .01) .63 -31 12 132 Closure head ring A508 Cl. 2 001055 .006 .86 16 16 156 Closure head flange A500 C1. 2 526916 .012 .82 -13 1 155 vessel flange A508 Cl. 2 218512 .016 .82 -4 -4 114 Inlet nozzle A500 C1. 2 526341-1 .04 .006 .16 -13 -13 141 Inlet nozzle A508 C1. 2 526395-1 .05 .009 .13 -31 -31 114.5 Inlet nerale A508 Cl. 2 526531 .06 .009 .16 -22 -22 129 Inlet narzle A508 C1. 2 52653) .06 ,009 .18 -40 -40 132 Outlet nozzle A508 Cl. 2 526341 .04 .001 .11 -13 -1 124 Outlet nozzle A508 C1. 2 525189 .05 .011 .83 -40 -24 103 Outlet nozzle A508 C1. 2 525189 .05 .010 .84 -49 -16 lib Outlet nozzle A508 C1. 2 526395-2 .03 .010 .14 -40 -30 121 Nozzle shell A508 Cl. 2 411085 .006 .89 -4 1 151.5 (c) Ir.ter. St. ell A500 C1. 2 526840 .16 .012 .85 -4 - 4(h) 141 96(b) ~ 141(b) Lower snell A508 Cl. 2 411331-11 .15 .004 .88 -30 -30(b) 152(C) Sottom head ring. ADS Cl. 2 521428 .06 .013 .11 -4 15 109 Bottcme head segment A5338, C1. 1 55126-2 .001 .59 -49 -2 136 Bottom heae segment A5334. C1. 1 55126-2 .001 .59 -40 -40 131 Bottom head segment A5338, C1. 1 55292-2 .006 .58 -13 -13 142 Bottom hee 4 segment A5334. Cl. 1 55292-2 .006 .58 -13 -13 132 Sottom head domme A5338, C1. 1 55292-3 .006 .58 -40 -40 121 --12;(b) laterinediate to lower shell weld (d) .05 .010 .10 -16 -68(b) 125.5(b) -16 -16(b) Ideld HAZ Jede - hajor Idorking Strection lead 0 - Normal to Mejor Idorking Otraction al Estimated per NuafG-0600, USN8C Standard Soview Plan, Branch Tech. Position - MIES 5-2. b) Based on actual data, t) 1005 slicar not reached, upper shelf energy is greater than listed. d) Submerged arc.seld (weld wire heat 895015 and Grau Lo Flux Lot No. H 6). h

12216 20 l I l J ff lll l'N i f l' lil lli I I l l l11 I:i i i ~ l 'lil li, li l fl ji lll lil l<, I f Jl l lil l ' / ) ) l lIll LEA'K TEST

/

f [ l llll / LIMI,T f f f ll ljj i l l j f (- llilill I 200a0 ,j ij l i-f l f I li C i / I/ i IIi 5 / l/ Illi i i h. I f f Il I I = / / I! I I / If ll l 5 1 / / - illi i I / c llill i[ 1 ~ ij l l l l~ l l I f I l i ll/ il !I'll' / i !I I' I ,I HEATUP-( CURVE 4 f l l jj l, l V I l'! I llI 1 i 1 s[l l lll llI ,I i W I l l l l j ill il CRITICALITY LIMIT SASED j ll l llil l 1 ON INSERVICE HYDROSTATIC 1-TEST TEMPERATURE (323_F) f I I I III'! 8 FOR THE SERVICE PERIOD l lll}l UP TO 10 EFPY ~ l l s I' O.0 O.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPERATURE (OF) MATEAf AL PROPERTY 'tASIS CONTROLLING MATERIAL

WELD METAL COPPfR CONTENT 0.30 wt%

PHOSPHORUS CONTENT 0.013evi% 8 i RTNOT INITIAL

0F RTNOTAFTER 10 EW 1/4T.I T F 3/4. 63*F CURVE APPLle.ABLE FOR NEATUP RATES UP TO 80*F/HR FOR THE SERVICE f ERIOD UP TO 10 EFPY'AND CONTAl'88 MARGINS CF 10 F AND80 PSIG FOR POSSIBLE INSTRUMENT ERRORS.

8 Figure 72 4-l WF McGu.'re Unit i Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY $cycxcoct : G)CH P ! 7M \\

3 G 3.0 I II I I I I ll I l ll l i l l I j 1 I l I lli i I IIl i I I I Il I II! l l l l lf 1 I ii! I I Ill I l} I l { l I ll / \\ [ l 11 I I I f 5 I B !I l 1 I I 1 l Il E !i ,l j / I 2 il l ll f l Ill E II i l ] i l I a _',Ii _ll I/ I I I ]- ill Ill il I fl Il l I I I I I I I l/ l i i i l l I I I III 'I I Ill I l l f illl il I I i 11! I l i l l' l I I I I Ill \\ .COOLDOWN RATES Mpf ll 1 IlI l 'F'hr

fff I

I I l' f I! 'O M7

0 e'A/

) l T-I I II 4 i .ani s-i ii l ' T'}9 ~ l ' l l l l l I ll 0.0 s 0.0 100.0 200.0 300,0 400.0 500.0 j INOICATED TEMPERATURE ( OF ) M ATE #f AL PHO*(RTY BA$tS CONTROLLING MATERIAL WELD M ETAL COPPER CO*dTENT 0.30 wt% 'HOSPHOR'JS CONTENT

0.013 wt%

RTNOT INITIAL O'F 1/4.1730 RTNOTAFTER 10 E FPY 3/4, s3a,F CUPVf1 APPtlCA8LE FOR COOLDOWN RATES UP TO 100 F/HR FOR THE SERVICE PERIOD UP 0 TO 13 ErPY Ai40 CC,NTAINS MARGINS OF 10 F ANO 40 PSIG FOR POSSIBLE INSTRUMENT ERRORS. 0 F*ngare fl'h-A 6 McGuire Unit 1 Re.ictor Coolant System CoOldown Limitations Applicable for the First 10 EFPY f CM AfA)c.t.. G)C W-l0 786 s m

l w l l l 2600 2 INSERVICE LEAK TEST 2400 - MINIMUM TEMPERATURE l 2200 2 gl 2 CURVE APPLICABLE FOR 1 - THE SERVICE PERIOD UP TO lIl 2000 ; 8 EFPY AND CONTAINS _ MARGINS OF IO'F AND l ! - 60 PSIG FOR POSSIBLE l 3 I800 - INSTRUMENT ERRORS !l l [ I 1600 2 ~ m 1400 [ HEATUP RATES TO m g 60*F/HR MATERIALS BASIS Q. 1200 2 CONTROLLING MATERIAL -REACTOR VESSEL g INTERMEDIATE SHELL 05 H 1000 COPPER CONTENT O.16 .h RT AFTER 8 EFPY 800 Z RT INITIAL-4*F NOT ET 600 l/4T.8S'F 2 3/47.36*F 400 2 CRITICALITY LIMIT 200 - 0 -IIIII1I I 1 I I I 1II II i i ei;;,,,,,,,, 50 i00 150 200 250 300 350 400 INDICATED TEMPERATURE (*F) Fugare f*2 4-) ISHEgliESEE McGu;re Umt 2 Reactor Coolant System Heatup Limitations Applicable for the First 8 EFPY Ef(At</C(,: N Yl ~ llUI-f T

r ::. 2600 j 2400 CURVE APPLICABLE FOR THE SERVICE f PERIOD UP TO O EFPY ANO l CONTAINS MARGINS OF IO*F 2200 AND 60 PSIG FOR POSSISLE Z INSTRUMENT ERROR l Z 2000 Z 1800 E. Z ~ 1600 Z 1400 l 1200 O Z MATERIAL BASIS N ~ i000 CONTROLLING MATERIAL 6 2 -REACTOR VESSEL COOLDOWN RATE 800 INTERMEDIATE SHELL 05 'F/HR COPPER CONTENT O.16 600 Z O RT

INITIAL,

-4'F 20 NT Z 60 RT AFTER 8 EFPY 400 T 100 1/4T,85'F 3/4T,36'F 200 Z II ' II'IIIII IIIlIIII'II'I I O O 50 000 150 200 250 300 350 INDICATED TEMPERATURE (*F) FIG UR E F 2* 9-4 W McGuire Unit No. 2 Reactor Coolant System Cooldown Limitations Applicable for the First 8 EFPY fepiane t.: ox# - !!alf w '\\

ARTNDY = [40 + 1000 (% Cu - 0.08) + 5000 (% P - 0.008)l [f/1019j1/2 UPPEB F 300 a E b 200 b e#' se s $i g iOO =. $ F u O Ww 0.10% Cu 0.35 0.30 0.25 0.20 % Cu 0.15% Cu 4 40 %P = 0.012 % Cu = 0.08 o % P = 0.008 O PLATE METAL (B 50121) 30 A WELD METAL o S !E l l IIIl! l I I IIll l 1, 1 20 2 x 1017 4 6 8 1018 2 4 6 8 1019 2 4 6 2 NEUTRON FLUENCE (E >1 Mev)(n/cm ) 5 Fu)are 52. ef g fqHFFK:1 Pre e ed Adjustment of Reference Temperature as a Function of Fluence, Copper, and Phosphorus Contents a g_ Q, u a, wca r-lo 7N. G[. y* t-a-L uc.rm dF'aiusiJrWirinn vruri-Ioto4 Hew 1asitI j

I 1 1 1 1020 j i ( SURFACE, t 2.211 X 10 p 19' j 10 e 1/4 f e-- gg f p 1.747 X 10 N / s / /-/'2 h 7 U f f f ~ s z 3/4 T $ 1018 g' m 310" ) d Z~ .-._/ 7 1 z h, O ~r y I N N ~ f ,i ) 1017 r, j j j 1016 l 0 2 4 6 ~ 8 10 12 14 16 18 20 22 24 26 28' 30 32 SERVICE LIFE (Effective Full Power Years) i F.9wrof1 f-f tamna-E2. Fast Neutron Fluence (E >1 Mev) as a Function of Full Power Service Life (EFPY) 6 fa nO ' lA!Cf} f - ]O ]$f', i 1

SIGNIFICANT HAZARDS CONSIDERATION 1 ) (RLG19)

ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION Pursuant to the requirements of 10CFR50.91, this analysis provides a determination that the proposed amendment of the Technical Specification does not involve any significant hazards consideration, as defined by 10CFR50.52. A description of the amendment request has been provided La an earlier section of this submittal package. In order to facilitate staff review of this no signifi-cant hazards consideration, each element of the proposed change is being discussed ceparately. The following elements makeup this proposed technical specification revision: Element 1: Limiting Condition for Operation 3.4.9.1 The Reactor Coolant system (except the pressurizer) pressure and temperature shall be maintained within the limits determined by analysis performed in accordance with the requirements of 10CFR50, Appendix G. The Reactor Coolant Sgstem (except the Pressurizer) heatup rate and cooldown rate shall not exceed 100 F/ hour. Element 2:' Applicability Modes 1, 2, 3, 4, 5, and 6 with the reactor vessel head installed. Element 3: Action With any of the above pressure-temperature limits exceeded, restore the pressure cnd/or temperature to within the limit within 30 minutes; perform an engineering cvaluation to determine the effects of the out-of-limit condition on the struc-tural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS Tavg and pressure to less than 200 degrees F and 500 psig, respectively, within the following 30 hours. Element 4: Significant Hazards Consideration regarding Renumbered let surveillance, Relocated 2nd surveillance, Relocated Figures /- Tables, added once per 12 hours surveillance. Element 5: Bases as revised and contained on the proposed change pages ELEMENT 1 - Proposed change to LCO 3.4.9.1 involves both an administrative change and a technical change. The proposed administrative change is the rewrite of the LCO itself. The technical changes are the revised pressure-temperature limits presented in the capsule reports. This proposed change does not involve a significant hazards consideration because cperation of McGuire Nuclear Station in accordance with this change would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The administrative aspect of this change simply makes the LCO a licensing tie between the regulatory requirement for pressure-temperature limits (Appendix G) and the ACTION statement. McGuire is still required to meet the regulatory requirement. The details of when the LCO is applicable that had been in the LCO are now in the "APPLICA-

-~ l Page 2 BILITY" section. The heatup and cooldown rates are provided on the respec-tive heatup and cooldown curves, the assumptions used in the analysis are included in the capsule reporcs which now become part of the McGuire FSAR. The maximum heatup and cooldown rates are specified. Station procedures for heatup, cooldown, inservice Jeak and hydrostatic test are based on the curves contained in the capsule reports. These curves have been developed using methodologies consistent with the requirements of 10CFR50, Appendix G. Accordingly, the curves are technically justified. (2) Create the possibility of a new or different kind of accident from any previously analyzed. In consideration of the statements made above, it has been determined that a new or different kind of accident will not be possible due to this change. (3) Involve a significant reduction in a margin of safety. The margin of safety is relevant to the technical adequacy of the pressure-temperature limits. Inasmuch as the limits have been developed in a manner that is consistent with 10CFR50, Appendix G, the margin of safety is maintained. Thus, for Element 1. no significant hazards is involved. ELEMENT 2 - Proposed change to APPLICABILITY section. This proposed change seeks to clarify the applicability of this LCO. It is an administrative change. This proposed change does not involve a significant hazards consideration because operation of McGuire Nuclear Station in accordance with this change would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. This change simply makes the proposed LCO consistent with the proposed surveillance applicability. The existing applicability states "at all times" which is incorrect in that the spec-ification is not applicable during Mode 6 operation with the vessel head removed. The proposed change has no effect on the probability or conse-quences of an accident previously evaluated. (2) Create the possibility of a new or different kind of accident. The appli-cability is consistent with regulatory basis. This proposed change cannot create the possibility of a new or different kind of accident. (3) Involve a significant reduction in a margin of safety. The proposed change is a clarif f tation of an existing applicability. It has no impact on safety margin. Thus, for Element 2, no significant hazard is involved. l

~ I' i~ Page 3 e ) ELEMENT 3 - Proposed change to ACTION section. This proposed change transposes the words " pressure" and " temperature" to make it consistent with the title of the i specification, the LOO itself, the surveillance requirement and the Bases. It is simply an administrative change. i l This proposed change does not involve a significant hazards consideration because j operation of McGuire Nuclear Station in accordance with this change would not: (1) Involve a significant increase in the probability or consequence of an accident previously evaluated. This change has no impact on probability or consequences of accidents previously evaluated. (2) Create the possibility of a new or different kind of accident. This proposed change cannot create the possibility of a new or different kind of accident. I (3) Involve a significant reduction in a margin of safety. This proposed change is editorial and has no impact on safety margin. Thus, for Element 3, no significant hazard is involved. i ELEMENT 4 - Several proposed changes of an administrative nature and one tech-i nical: I a) Existing surveillance 4.4.9.1.1 is renumbered to 4.4.9.1. b) Existing surveillance 4.4.9.1.2 is relocated from technical specifi-cations to FSAR section 5.4 and Table 5.4.3-1. c) Existing technical specification figures 3.4-2a, 3.4-2b, 3.4-3a, and 3.4-3b are relocated to FSAR figures 5.2.4-1, 5.2.4-3, 5.2.4-2, and 5.2.4-4 respectively. Existing Table 4.4-5 is relocated to FSAR Table 5.4.3-1. The table has been revised in accordance with recommendations in the capsule report for Unit 2. d) Added once per 12 hour surveillance. This proposed change does not involve a significant hazards consideration because l operation of McGuire Nuclear Station in accordance with this change would not: i (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The renumbering of a surveillance and the relocation of certain requirements to the FSAR are administrative. Inasmuch as the figures and tables contain information that reflects regulatory requirements, compliance by McGuire is still required. Thus, as the plant will still be in compliance with regulatory specified requirements the probability or consequences of an accident previously evaluated are not i affected. The change in the table for withdrawal of surveillance capsules reflects the inadvertent removal of Capsule V during the first Unit 2 re-i fueling outage. The next capsule to be removed X, will be evaluated in 1 l I i ~.

i Page 4 4 4 sufficient time to revise the pressure-temperature limit curves prior to 8 EFPY, the present effective period. The remainder of the withdrawal schedule is consistent with the recommendations of ASTM E185. Thus, as stated pre-viously, as the plant will still be in compliance with regulatory specified requirements, the probability or consequences of an accident previously evaluated are not affected. l (2) Create the possibility of a new or different kind of accident. The changes discussed in this element do not in any way change the manner in which the plant will be operated. This proposed change cannot create the possibility of a new or different kind of accident. (3) Involve a significant reduction in a margin of safety. The proposed changes j discussed in this element do not in any way change the manner in which the plant will be operated. They have no impact on safety margin. l I Thus, for Element 4, no significant hazard is involved. ELEMENT 5 - Bases - The Bases of Technical Specifications are provided but are not part of technical specifications (50.36(a)) nor part of the Operating License. I Significant hazards considerations are necessary only for amendments to the operating license. Accordingly, no significant hazards review is necessary for the proposed revision to the Bases included along with this proposed license amendment. l Therefore, based on the considerations presented above, as well'as the discussions presented in the preceding sections of this submittal, Duke has determined that this change does not involve a significant hazards consideration. 4 1 l d ,-,,--,,,_-,.~----n_, ,,m_.,_n,__, , ~,., ,,,--m,--.,,--~_,--,,_,.,---,--}}

v t e Letter Sequence Request
TAC:62779, (Open) TAC:62780, (Open)
Initiation Request, Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20100H434, ML20100H458, ML20140A557, ML20141C845, ML20141C857, ML20210U330, ML20215K962, ML20235W092, ML20236B613
MONTHYEAR1986-10-21 [Table View]

ML20210C033

Text

2.9 REFERENCES

4.5 REFERENCES

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