ML20217D635

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Proposed Tech Specs Sections 3/4.4.9 & B 3/4.4.9,revising Heatup & Cooldown Curves
ML20217D635
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 10/08/1999
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20217D630 List:
References
NUDOCS 9910180029
Download: ML20217D635 (29)
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Text

.-

4 INDEX BASES SECTION PAGE 6~

3/4.4./ STEAM GENERATORS.......................................... 8 3/4 4-3 l 3/4.4]REACTORCOOLANTSYSTEMLEAKAGE............. B 3/4 4-4 3/4.4./ CHEMISTRY.................................................

B 3/4 4-5 3/4.4 f. SPECIFIC ACTIVITY........................................ B 3/4 4-5 3/4.4.

JO PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 3/4.4.)sSTRUCTURALINTEGRITY...................................... B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SU85YSTEMS............................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK (RWST)........ .............. B 3/4 5-2 8 3/4.6 CONTAIhMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYS11MS...................... B 3/4 6-3 3/4.6.3 PARTICULATE IODINE CLEANUP SYSTEM......................... B 3/4 6-4 3/4.6.4 CONTAINMENT ISO LATION VA LVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-5 3/4.6.5 COMBUSTIBLE GAS CONTR0L............................ ...... B 3/4 6-5 9910180029 991008 PDR ADOCK 05000395 P PDR SUMMER-UNIT 1 4III Amendment No. a

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INDEX i

' BASES I

SECTION PAGE I

3/4.4.5 STEAM G E N E RATO R S . . . . . . . . . .. . . . . . . ... . . . ... . .. . . . .. . .. . . . . . .. . . . .. . . . . . .. . .. .. . ... . . . B 3/4 4 -3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE......................................... B 3/4 4-4 3/4.4.7 C H E M I ST R Y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4 -5 3/4.4.8 SP ECI FI C ACTIVITY . . . .. . . . . . . . . . . . . . . . . .. . . . .. . . .. .. .... . . . . ... . . .. . . . . . ..... . . . .. . . ... .. .. . B 3/4 4-5 3/4.4.9- PRESSURE /TEMPERATU RE LIMITS .................................................. B 3/4 4-6 3/4.4.10 STRUCTU RAL INTEG RITY....... ..................... .................................... B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 3/4.5.1 REACTOR TRIP SYSTEM INSTRUMENTATION ................... .... ...... B 3/4 5-1 1 l

2/4.5.2 and 3/4. 5.3 ECC S SU B SYSTEM S .. .. .. . ... .. . . . .... .. ... . ... ... .. . . . .. . . . . . . ... . . . .. .. . . .... . . B 3/4 5- 1 3/4.5.4 REFUELING WATER STORAGE TANK (RWST.................... ............. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 1

I 3/4.6.1 PRI MARY CONTAINMENT . .. ... ........... . ................ .. ..... ... .... .. . ............. B 3/4 6-1 l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS....................... ...... B 3/4 6-3 3/4.6.3 PARTICULATE LODINE CLEANU P SYSTEM....................................... B 3/4 6-4 3/4.6.4 CONTAINMENT ISOLATION VALVES.............. .......... ........... ........... B 3/4 6-5 3/4.6.5 COMBU STIBLE GAS CONTROL ...................................................... .. B 3/4 6-5 ,

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SUMMER - UNIT 1 Xill Amendment No. 44 1 i

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REACTOR COOLANT SYSTEM MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1, -2 LIMITING ART VALUES AT 32 EFPY: 1/4T, 107 F 3/4T, 94 F 2500 ,

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Figure 3.4-2 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to 50 and 100 F/hr) Applicable for the First 32 EFPY (Without Margins for Instrumentation Errors)

SUMMER - UNIT 1 3/4 4-31 Amendment No.53,-119,-193,

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SUPMER - UNIT 1 3/4 4-32 AmendmentNo.53.///

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4 REACTOR COOLANT SYSTEM MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1, -2 LIMITING ART VALUES AT 32 EFPY: 1/4T, 107 F l 3/4T,' 94 F l

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SUMMER - UNIT 1 3/4 4-32 Amendment No. 53r1497133,

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) -

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/gran 00SE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible lodine spiking phenomenon which may occur following changes in THERMAL POWER.

Reducing T,,, k less h M pmenu W release of MW skuld a steam generator tube rupture since the saturation pressure of the primary 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 primary coolant will be detected in sufficient time to take corrective action. Infometion obtained on iodine spiking will be used to assess the parameters associated with spi. king phenomena. A reduction in .

frequency of isotopic analyses following power changes any be pemissible if .

justified by the data obtained.

c. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limitut- 5: xxt ':-t "t' : x ;1. --t: ;i n: in : . _ _- . -. _ . -

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1) The rector 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.

a)~ Allowable combinations of pressure and'tamperature for specific temperature change ratas are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

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=y REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY - (Continued)

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/

gram DOSE EQUlVALENT l-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Reducing Tm to less then 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary 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 primary coolant will be detected in sufficient time to take corrective action. Information obtained on l

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.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited by curves developed using the methodology from Westinghouse Topical Report, WCAP-14040-NP-A, updated to include the requirements of the 1996 ASME Boiler and Pressure Vessel Code,Section XI, Appendix G along with ASME Code Case N-640.

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.

'a) Allowable 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 be obtained byinterpolation.

SUMMER - UNIT 1 B 3/4 4-6 Amendment No. 53r64r

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) 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 limit lines shall be calculated periodically using methods p,,.-;u u w.ua.gh a.4a1L.raec.
3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F.
4) The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200*F/hr respectively. The sper.y shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 625'F.
5) System in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

fracture toughness properties of the ferritic materials in the reactor ves 1 are determined in accordance with the 1972 Summer Addenda to Secti III of the A5ME Boiler and Pressure Vessel Code.

Neatup and 1down limit curves are calculated using the mo t limiting I value of RT NDT eference nil-ductility temperature). The m<st limiting RT of the sater in the e region of the reactor vessel is deter-NDT mined by using the pre rea wr vessel material properties and j estimating the n'adiation- (T RT is designated as the T. HDT higher of either the drop wei uctility transition temperature l (NDTT) or the temperature at wh1 material exhibits at least 50 ft Ib of impact energy and 35-mil latera ansion (normal to the major work-ing direction) minus 60*F. l RT NOT increases as the material is exposed fast-neutron radiation.

Thus, to find the most limiting RT NDT at any ti eriod in the reactor's life, ART due to the radiation exposure associat with that time NOT period must be added to the original unirradiated RT The extent of NDT.

the shift in RT NOT is enhanced by certain chemical elements uch as ,

copper) present in reactor vessel steels. Design curves which s the effect of fluence and copper content on ART for reac r vessel s Is are shown in Figure B 3/4 4-2. NOT SUMMER - UNIT 1 B 3/4 4-7 Amendment No. 53

4:

REACTOR COOLANT SYSTEM

-BASES PRESSURErrEMPERATURE LIMITS (Continued) 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 limit lines shall be calculated periodically using methods acceptable to the NRC. l
3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 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 spray fluid is greater than 625*F.
5) System in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

SUMMER - UNIT 1 B 3/4 4-7 Amendment No. 6%

REACTON COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

' en the copper content of the most limitir.g material, the radiation-in ed ART NOT can be estimated from Figure B 3/4.4.2. Fast neutron fluenc E > 1 Mev) at the vessel inner surface, the 1/4 T (wall thick-ness), an /4 T (wall thickness) vessel locations are given as a func-tion of ful ower service life in Figure B 3/4.4.1. The data for all other ferritic terials in the reactor coolant pressure boundary are examined to '7sur hat no other component will be limiting with respect to RT NDT' The preirradiation fractu hn s properties of the V. C. Summer

, Unit I reactor vessel mater presented in Table B 3/4.4-1. The fracture-toughness properties critic material in the reactor coolant pressure boundary are de d in accordance with the NRC Regulatory Standard Review Plan.1 postirradiation fracture-toughness properties of the reactor vessel belt e material were obtained directly from the V. C. Summer Unit 1 Vessel Mate 1 Surveillance Program.

The ASME approach for calculating the allowab limit curves for various heatup and cooldown rates specifies that the tot stress intensity factor, Kg , for t$ttpombined thermal and pressure stresses any time during heatup or coold8wg cannot be greater than the referenc tress intensity factor, KIR, f r the metal temperature at that time. K IR btained from 4

the reference fracture t' dug'hness curve, defined in Appendix f the ASME Code.2 The KIR """** 15 9I 'O uby the equation:

.z K

IR=26.78+1.223exp[0.0145(JRTNOT + 160)] Equation )

1 Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5. . ndard Revi lan or 'he Review of Safety Analysis Reports for Nuclear d ipn, NUREG-0800', 1981.

2ASME Boiler and Pressure Vessel Code, Se I Division 1 - Appendices,

" Rules for Construction of Nuclear Vessels," Appe " Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, Ame ciety of Mechanical Engineers, New York,1983.

SUMMER - UNIT 1 8 3/4 4-8 Amendment No. 53

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REACTOR COOLANT SYSTEM BASES RESSURE/ TEMPERATURE LIMITS (Continued) are K gg is the reference stress intensity factor as a function of the 1 temperature T and the metal reference nil-ductility temperature RT Thus, the governing equation for the heatup-cooldown analysis is N0 defin in Appendix G of the ASME Code as follows:

CK y +K It # K IR Equation (2) where K gg is the ress intensity factor caused by membrane (pressure) stress K is the stres intensity factor caused by the thermal gradients It C = 2.0 for Level A ve B service limits C = 1.5 for hydrostatic test conditions during which the reactor core is not crit At any time during the heatup or oldown transient, K IR is determined by the metal temperature at the tip of postulated flaw, the appropriate value for RTNOT, and the reference fr ture toughness curve. The thermal stresses resulting from temperature gra nts through the vessel wall are calculated and then the corresponding (th 1) stress intensity factors, kit, for the reference flaw are computed, om Equation 2, the pressure stress intensity factors are obtained and, fr these, the allowable pressures are calculated.

COOLDOWN For the calculation of the allowable pressure-versus- lant temperature during cooldown, the Code reference flaw is assumed to exist t the inside of the vessel wall. During cooldown, the controlling location o he flaw is always at the inside of the wall because the thermal gradients duce tensile stresses at the inside, which increase with increasing cooldown r es. Allow-able pressure-temperature relations are generated for both steady-r te and finite cooldown rate situations. From these relations, composite li 't curves are constructed for each cooldown rate of interest.

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UMMER - UNIT 1 B 3/4 4-11 Amendment No.53

/

REACTOR COOLANT SYSTEM BASES NLDOWN(Continued) e use of the composite curve in the cooldown analysis is necessary be-cause c trol of the cooldown procedure is based on measurement of reactor coolant t

~

erature, whereas the limiting pressure is actually dependent on the material t erature at the tip of the assumed flaw. During cooldown, the 1/4 T vessel locati is at a higher temperature than the fluid adjacent to the vessel ID. This condi n, of course, is not true for the steady-state situation. It follows that, at Gi ven reactor coolant temperature, the AT develooed during cooldown results in IR at the 1/4 T location for finite cool-higher value of K down rates than for ste '-state operation. Furthermore, if conditions exist such that the increase in g exceeds kit, the calculated allowable pressure during cooldown will be grea r he steady-state value.

The above procedures are ne se there is no direct control on tem-perature at the 1/4 T location and, e, allowable pressures may unknowingly be violated if the rate g is decreased at varivus intervals along a cooldown ramp. The use of the site curve eliminates this problem and insures conservative operation of th stem for the entire cooldown period.

Three separate calculations are require o determine the limit curves for finite heatup rates. As is done in the coold analysis, allowable pressure-teeperature relationships are developed for sten state conditions as well as finite heatup rate conditions assuming the presence f a 1/4 T defect at the ,

inside of the vessel wall. The thermal gradients dur heatup produce com- l pressive stresses at the inside of the wall that allevi the tensile stresses  !

produced by internal pressure. The metal temperature at crack tip lags the coolant temperature; therefore, the K IR f r the 1/4 T crack ring heatup is

. lower than the K gg for the 1/4 T crack during steady-state co ions at the same coolant temperature. During heatup, especially at the end o he tran- )

sient, conditions may exist such that the effects of compressive the al i stresses and lower K IR 's do not offset each other, and the pressure-te erature curve based on steady-state conditions no longer represents a lower bound f l

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StMER - UNIT 1 B 3/4 4-12 Amendment No.53 l

)

REACTOR COOLANT SYSTEM BASES a

The similar curves for finite heatup rates when the 1/4 T flaw is considered.

ant t fore, both cases have to be analyzed in order to insure that at any cool-erature the lower value of the allowable pressure calculated for steady-s te and finite heatup rates is obtained.

The sec d portion of the heatup analysis concerns the calculation of pressure-tempe face flaw is ass ture limitations for the case in which a 1/4 T deep outside sur-

d. Unlike the situation at the vessel inside surface, the thermal gradients e ablished at the outside surface during heatup produce stresses which are te ile in nature and thus tend to reinforce any pressure stresses present. Thes thermal stresses are dependent on both the rate of heatup and the time (or c lant temperature) along the heatup ramp. Since the thermal stresses at the out e are te ile and increase with increasing heatup rates, each heatup rate must b a d on an individual basis.

Following the generation of p s state and finite heatup rate situati s -te erature curves for both the steady-constructing a composite curve based o inal limit curves are produced by r t-by point comparison of the steady,-state and finite heatup rate data.

able pressure is taken to be the lesser of aany given temperature, the allow-three values taken from the curves.,under consideration. The use of the c osite curve is necessary to set conservative heatup limitations because it is p sible for conditions to exist wherein, over the course of the heatup ramp, the trolling condition switches from the inside to the outside and the pressure lim must at all times be based on analysis of the most critical criterion. Then the aposite curves for the i

heatup rate data and the cooldown rate data are adjuste or possible errors in the pressure and temperature sensing instruments by the va es indicated on the f respective curves.

i Finally, the new 10 CFR 508 rule which addresses the metal erature of the closure head flange and vessel flange regions is considered. 10 CFR 50 rule states that the metal temperature of th closure flange regions a t exceed the material RTNOT by at least 120*F. for normal operation when the pres re 4

3Coae oi i.  :' "=anirtions, 10 CFR 50, Appendix G " Fracture Toughness Requirements," U.S. Nuedea ission, Washington, D.C.,

Amended May 17, 1983 (48 Federal Register 24 .

SUMMER - UNIT 1 B 3/4 4-13 Amendment No.53

REACTOR COOLANT SYSTEM BASES The following Pages have been deleted:

B 3/4 4-E.

B 3/4 4-9 B 3/4 4-10 B 3/4 4-10a B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 I

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l SUMMER - UNIT 1 8 3/4 4-8 Amendment No. 6%

C, 9 4 2. 3 ~ l,, -1. sec +k e. b m

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oll Mestof sad Cesidousr Cenos.s b /$e. ges ein" fed. THe se psynter/Als erhobst boroMf /AU Alvas of t o 7 *f at 1/4 T c.osd 94*F er 3/+ T -

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REACTOR CUULANT SYS BASES A41ssmnx l'TE AeJE447%tt z s.e.iTs (Qnt,aua.d) ,

emee:2 20 ;;r::nt c' th: pr:ivh: hydr::ttth i::t pr:::er: (521 ;;f; ';r V. C. 3_ _, G, ; t 1). T. Lie " 0/4.4.1 indi;etee thet the if ritii.; "TNDT "I 10*T ;;;;r: 'n th: 5::d '!: ;; :" '!. O. 5 :r '.'ait 1, :nd th; ch'- -

1h;;-

bh t:-.;:r:tur: c' th h 7 ;'On h 130*f :t pr:;;;r ; er.:ter then 021 p;ie.

_ .(Leit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated" :h; th: =th:2 dh::;nd. Th: t rh:t ha ;'

th: ' h't :;ru n h prc: rted '- th: """ ";;;ht:r; St:nd: d ":r' : " hr

  • Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been..obtained.directly. from.the reactor pressur

. vessel surveillance program. Charpy tes ecimens from Capsule Jainaicar.e ,

that i:05 th: c'^r;i'h :: =1d =te? rr_ re region '-t r--dirt shell plate _ ---

code no, .^.?'"' ' rh'5f ted th' *u ' "'  :' 20*F at affluence of M G"#CI i 10 % cm - --$: ht h =1' d th'-  :;;n;rhh 2:i; Curr

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=r, th :::!dr cur: *= 'f;rn 3.'-3 h '- rted by th h 10 C"" 50 re' .

Allowable combinations of temperature and pressure for specific temperature I change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 3.4-2. This is in addition to her criteria'wMeh must h t before the reactor is made critica as <>c.,sst_ 4 1%e 4s #e w,q <aaraf Ard /4/ qa The leak test limit curve shown in Figure 3.4-2 represents minimum tempera-ture requirements at the leak test pressure specified by applicable codes. The leak test limit curve was determined by methods of ";"n n:n 2 :nd ' (--- --

res 3.4-2 and 3.4-3 define limits for insuring prevention of nonductile_ _

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-NtHIEG-0000--t90t. I SUDMER - UNIT 1- B 3/4 4-14 Amendment No.53

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i insert A l

' The reactor must not be made critical until pressure - temperature combinations are to the right of the criticality line shown in figure 3.4-2. The criticality limit curve specifies pressure - temperature limits for core operation to provide additional margin during  ;

actual power production as'specified in Appendix G to 10 CFR 50. The pressure -

temperature limits for core operation (except for low power physics tests) are that the

' reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40 F higher than the minimum pem11ssible temperature in the corresponding pressure - temperature curve  ;

for heatup and cooldown, calculated as described in this technical basis. The vertical line drawn from these points on the pressure - temperature curve, intersecting a curve 40 F higher than the pressure - temperature limit curve, constitutes the limit for core  !

operation for the reactor vessel.  !

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REACTOR COOLANT SYSTEM BASES I

PRESSURE / TEMPERATURE LIMITS (Continued)

I Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated as described in Westinghouse Topical Heport, WCAP-15102, Rev. 2, "V. C. Summer Unit 1 Heatup and Cooldown Curves for Normal l Operation".

Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program. Charpy test specimens from Capsule W indicate that the core region lower shell plate code no. C9923-1, -2 are the limiting beltline materials for all heatup and cooldown curves to be generated. These materials exhibit limiting ART )

values of 107 F at 1/4T and 94 F at 3/4T at a calculated inner surface fluence of 3.84 x  !

2 '

10" n/cm at 32 EFPY.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and  !

cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 3.4-2. This is in addition to other criteria which must be met before the reactor is made critical, as I discussed in the following paragraphs.

i The leak test limit curve shown in Figure 3.4-2 represents minimum temperature l requirements at the leak test pressure specified by applicable codes. The leak test limit .,

curve was determined by methods of the Standard Review Plan, Chaptor 5.3.2 and Appendix G of the ASME Code,Section XI. l The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 3.4-2. The criticality limit curve specifies pressure - temperature limits for core operation to provide additional margin during actual power production as specified in Appendix G to 10 CFR 50. The pressure

- temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40 F higher than the minimum permissible temperature in the corresponding pressure - temperature curve for heatup and cooldown calculated as described in this technical basis. The vertical line drawn from these points on the pressure - temperature curve, intersecting a curve 40 F higher than the pressure - temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 3.4-2 cnd 3.4-3 define limits for insuring prevention of nonductile failure.

Thedostrument uncertainties, effects of forced flow from the reactor coolant pumps, arid the elevation effect of the pressure sensors are incorporated into the curves located in the plant operating procedures.

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SUMMER - UNIT 1 B 3/4 4-14 Amendment No. 53r l l

a Docum:nt Control D:sk Attachment 11 TSP 980009 RC-99-0154 l Page 1 of 5 l

L SAFETY EVALUATION FOR REVISING THE PRESSURE / TEMPERATURE LIMITS FOR THE VIRGIL C. SUMMER NUCLEAR STATION '

TECHNICAL SPECIFICATICNS I

Description of Amendment Reauest l The Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) are being I revised to incorporate the new Pressure / Temperature Limits Curves consistent with the  ;

analysis results of reactor vessel specimen W. These figures are contained in Section 1 3/4.4.9 and are presented as figures 3.4-2 and 3.4-3. These figures were developed using the methodology included in WCAP 14040-NP-A, Methodoloav Used to Develoo Cold Overoressure Mitiaatina System Setooints and RCS Heatuo and Cooldown Limit Curves, as well as Code Case N-640, Alternative Reference Fracture Touahness for Development of P-T Limit Curves for Section XI. Division 1. Additionally, the Bases section for the Pressure / Temperature Limits is being revised to accurately reflect current l industry standards and regulations. A significant portion of this Bases section is being deleted due to the information also being located in WCAP 15102, Revision 2.

Safety Evaluation The proposed changes to the PT curves reflect the results of the analysis performed on specimen W as part of the reactor vessel material irradiation surveillance specimen program. The analysia was performad and the calculations prepared using guidance contained within Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials and Appendix G to 10 CFR 50. Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels used for light-water-cooled reactor vessels.

The results of this analysis were provided to the NRC October 9,1998 as required by Appendix H to 10 CFR 50. The results are included in Westinghouse WCAP-15101, Revision 0, Analysis of Capsule W from the South Carolina Electric & Gas Company V.

C. Summer Unit 1 Reactor Vessel Radiation Surveillance Proaram. Also provided at that time was WCAP-15103, Revision 0, Evaluation of Pressurized Thermal Shock for V.

C. Summer Unit 1.

The new PT curves for normal heatup and cooldown of the primary reactor coolant system have been developed using the methods discussed in Section 3 and 4 of WCAP-15102, Revision 2. Unlike the current PT curves, the new curves 60 not account for instrument uncertainties. Instrument uncertainties will, however, be included within Plant Operating Procedures. The operational limit curves will include the following effects:

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  • Attachment 11 TSP 980009 RC-99-0154 Page 2 of 5 1
1. Instrument uncertainties associated with the pressure and temperature measurements.

2.- Pressure increases due to the elevation head differences between the pressJre measurement and the reactor vessel beltline region.

3. Pressure ircreases between the pressure measurement and the reactor vessel beltline region due to RCS flow (i.e., form, friction, and velocity head effects resulting from RC Pump operation).

The instrument uncertainties are calculated using methods consistent with ISA-S67.04.

Based on use of current wide range pressure and temperature measurements, the uncertainties range from 138 to 166 psi and from 14 to 26 F for readings taken from the plant computer or from the main control board analog indicators respectively. The pressure increases, based on 102% of best estimate flow and an elevation head conservatively referenced to the bottom of the reactor vessel, are 12,40,52 and 74 psi for 0,1,2, and 3 RCPs running, respectively. With these effects included, the operational limit curves will be substantially below and to the right of the new TS PT curves.

Overpressure Protection Section 5.2.2 of the Standard Review Plant (SRP) specifies that the low temperature overpressure protection system be designed in accordance with the guidance of Reactor System Dranch Technical Position (BTP RSB) 5-2. The BTP RSB 5-2 guidance specifies that the system be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to prevent RCS pressure from exceeding the Appendix G limits of 10 CFR Part 50 while operating at low temperatures. The existing low temperature overpressure system for VCSNS is provided by the relief valves in the suction lines of the residual heat removal (RHR) system as described in the original licensing submittal for the V. C. Summer Nuclear Station dated June 29, 1983.

The current Limiting Conditions for Operation (LCO) in TS 3.4.9.3 requires than the overpressure protection system be operable with two RHR relief valves with a lift setting of .5 504 psig and the associated RHR relief valve isolation valves open or the RCS depressurized with an RCS vent of > 2.7 inches. This LCO is applicable when any RCS cold leg is s 300 F when the head is on the reactor vessel. Also, TS 3.5.3 provides restrictions for a maximum of one operable centrifugal charging pump when any RCS cold leg temperature is less than or equal to 300 F, and TS 3.4.1.3 provides restrictions which preclude starting an Reactor Coolant pump when one or more of the RCS cold leg temperatures are less than or equal to 300 F unless 1) the pressurizer water volume is less than 1288 f t and/or 2) the secondary water temperature of each steam 4

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. TSP 980009 RC-99-0154 Page 3 of 5 i I

generator is less than 50 F above each of the RCS cold leg temperatures. These restrictions preserve design basis assumptions made within the overpressure analyses ,

(discussed below) and are supplemented by additional normal plant operating '

procedures which maximize the use of a pressurizer cushion (steam /nitropen bubble) during periods of low pressure, low temperature operation. Collectively, tnese administrative limits minimize the severity of potential overpressure transients and provide reasonable assurance that most transients can be terminated by operator action before the RHR relief valves are challenged. No change to the above Technical Specifications (TS) associated with the lift setting of the RHR relief valves, the minimum required vent size, or the TS enabling temperature of 300 F are proposed or are required to support the new TS PT curves.

Overpressure Analyses Using RHR Relief Valves Current overpressure analyses show that the maximum opening pressure and capacity of a single RHR relief valve is adequate to prevent the RV pressure from exceeding the l new PT limits. Consistent with the original design basis for the low temperature overpressure protection system at VCSNS, two types of overpressure transients are considered:

1. A mass addition transient resulting from operation of one centrifugal charging pump with letdown isolated. i
2. A heat addition transient caused by an inadvertent startup of one inactive i Reactor Coolant Pump with a 50 F mismatch between the prin:ary and secondary side.

The calculations assume that the transients occur while the pressurizer is in a water solid condition and cover RCS temperatures ranging from 60 F to 350 F.

l Table 1 shows the calculated peak pressures within the RV and the associated Appendix G limit assuming operation of a single RHR relief valve. Consistent with WCAP-14040-NP-A, the design credits the fact that overpressure events most likely occur during isothermal conditions in the RCS and therefore utilizes the steady-state Appendix G limit in judging the adequacy of the RHR relief valve capacity and lift setting.

Within the overpressure analyses, the opening setpoint of the RHR relief valve is assumed at 468 psig (i.e.,104% of the maximum TS limit) to account for instrument uncertainties (1%) associated with verification of the lift setting and potential variations (3%) in the opening setpoint between required surveillances (i.e., per TS 4.4.9.3.1.c).

As shown, the calculated peak pressures for both the design basis mass and heat addition transients remain substantially less that the new PT limit. Although not required to meet the new PT limits, Table 1 results for the mass addition transient reflect current administrative limits which preclude RC Pump operation below an indicated RCS temperature of 100 F and which limits running more than one RCP untilindicated RCS temperature is > 160 F. Current overpressure analyses thus demonstrate that the

n- ,

Docum:nt Control Disk r Att: chm:nt 11 TSP 980009 RC-99-0154 Page 4 of 5 capacity and lift setting of a single RHR relief valve remains adequate for low temperature overpressure protection.

Overpressure Analyses Using 2.7in" Vent in the event one of the RHR relief valves becomes inoperable, the RCS will be depressurized with an RCS vent of greater than or equal to 2.7 in" to provide overpressure protection. Potential vents include opening two or more of the pressurizer

. PORVs, removal of one or more or the pressurizer safety valves, or removal of the pressurizer manway. With the RCS depressurized, the RCPs will be stopped with power removed thereby making the mass addition transient the only credible transient.

Assuming the minimum required vent at the top of the pressurizer, peak RV pressures for the design basis mass addition event are < 250 psig; this is well below the new PT limits. Thus, the current minimum vent size remains adequate for low temperature overpressure protection.

Enable Temperature Branch Technical Position Reactor Systems Branch (BTP RSB) 5-2 requires the low temperature overpressure system to be in operation during startup and shutdown conditions below the enable temperature, defined as the coolant temperatures corresponding to a reactor vessel metal temperature of at least RTwor + 90 *F at the

beltline location (1/4T or 3/4T) that is controlling in the appendix G limit calculations.

)

Use of the'BTP RSB 5-2 definition requires an enable temperature of 235.2 F based on

- the following:

1. RTwor equal to 107 F, which corresponds to the limiting ART value for the 1/4T location @ 32 EFPY for the RV lower shell plates (per Table 4-12 of WCAP-15102, Revision 2).

- 2. RTwor equal to 94 F, which corresponds to the hmiting' ART value for the 3/4T l location @ 32 EFPY for the RV lower shell plates (per Table 413 of WCAP- j 15102, Revision 2).  !

l

3. - A metal to coolant temperature difference of 24.6 F for the 1/4T location and  !

51.2 F for the 3/4T location based on Westinghouse calculations supporting l WCAP 15102, Revision 2 for the maximum allowed heatup rate of 100 F/hr. l Since the current enable temperature within the Technical Specifications (300 F) is greater than the required enable temperature of 235.2 F, it remains conservative and {

need not be changed to support use of the new PT curves applicable to 32 EFPY.

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. Docum:nt Control D:sk Attachm:nt il TSP 980009 RC-99-0154 Page 5 of 5 The existing enable temperature margin (64.8 F) also exceeds the maximum uncertainty associated with the wide range temperature measurement (26 F), thus allowing the use of 300 *F as an operational limit with no adjustment for instrumentation error.

Table 1 Overpressure Analysis Results RCS Max RV Pressure For Max RV Pressure For Appendix G Temperature Mass Addition Heat Addition SS Limit (F) Transient Transient (psig) (psig) 60 528.3 504.8 621 (Note 3) .

74 (Note 1) 536.6 508.5 621 (Note 3) 100 536.1 514.9 621 (Note 3) 134 (Note 2) 574.6 ---

1315.8 150 574.3 525.7 1571 200 573.4 536.1 >2341 250 572.4 548.3 >2341 300 571.2 561.7 >2341

)

j 350 569.9 578.9 >2341 Notes:

1. Minimum actual temp below which RCP operation is not allowed.
2. Minimum actual temp above which 2 or more RCPs may be operated.
3. Flange Temperature Requirement l

+ Docum::nt Control D:sk o Attachment 111 TSP 980009 RC-99-0154 Page 1 of 3 NO SIGNIFICANT HAZARDS EVALUATION FOR REVISING THE PRESSURE / TEMPERATURE LIMITS IN THE VIRGIL C. SUMMER NUCLEAR STATION TECHNICAL SPECIFICATIONS Description of Amendment Reauest The Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS) are being revised to incorporate the new Pressure / Temperature Limits Curves consistent with the analysis results of reactor vessel specimen W. These figures are contained in Section 3/4.4.9 and are presented as figures 3.4-2 and 3.4-3. These figures were developed using the methodology included in WCAP 14040-NP-A, Methodoloav Used to Develoo Cold Overoressure Mitiaatina System Setooints and RCS Heatuo and Cooldown Limit Curves, as well as Code Case N-640, Alternative Reference Fracture Touahness for Development of P-T Limit Curves for Section XI. Division 1. Additionally, the Bases section for the Pressure / Temperature Limits is being revised to accurately reflect current industry standards and regulations. A significant portion of this Bases section is being deleted due to the information also being located in WCAP 15102, Revision 2.

Basis for No Sionificance Hazards Consideration Determination South Carolina Electric & Gas Company (SCE&G) has evaluated the proposed changes to the VCSNS TS described above against the Significant Hazards Criteria of 10 CFR 50.92 and has determined that the changes do not involve any significant hazard. The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequences of an accident prev:ously evaluated?

The proposed changes revise the Pressure / Temperature Limits Curves to provide curves that reflect the results of the analysis performed on reactor vessel surveillance specimen W. This analysis was performed using NRC approved methodology as documented in WCAP 14040-NP-A, utilizing the 1996 ASME Boiler and Pressure Vessel Code,Section XI, Appendix G requirements, along with ASME Code Case N-640. These curves provide the limits for operation of the Reactor Coolant System during heat up, cool down, criticality, and hydrotesting. These curves are provided without instrument uncertainties included, however, the uncertainties are included in the curves provided in the operating procedures. The limits protect the reactor vessel from brittle fracture by separating the region of acceptable operation from the region where brittle fracture is postulated to occur. Failure of the reactor vesselis not a VCSNS design basis accident, and, in general, reactor vessel failure has a low probability of occurrence and is not considered in the safety analysis.

R Docum:nt Control Disk . l J ' Att: chm:nt til

~ TSP 980009 RC-99-0154 Page 2 of 3 Therefore, the change does not involve a significant increase in the probability or i consequences of an accident previously. evaluated.

2. - Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes revise the Pressure / Temperature Limits Curves, Section 3/4.4.9, to incorporate the results of the analysis performed on reactor vessel specimen W. There are no plant design changes or significant changes in any operating procedures. This change adjusts the heatup and cooldown curves to reflect the shift in nil-ductility reference temperature of the reactor vessel as a result of neutron embrittlement, and alternate methodology utilized to generate l the curves. Therefore, the change does not create the possibility of a new or l different kind of accident from any accident previously evaluated. l

3. Does this change involve a significant reduction in margin of safety?

The proposed changes revise the Pressure / Temperature Limits Curves, Section 3/4.4.9, to incorporate the results of the analysis performed on reactor vessel .

specimen W. The new PT curves ensure that the 10 CFR 50 Appendix G, I requirements are not exceeded during normal operation including Reactor l Coolant System transients during heat up, cool down, criticality, and I hydrotesting. The new PT curves were prepared, using accepted industry methodology, for a projected reactor vessel neutron exposure of 32 EFPY. )

The new curves will serve as the basis for operating limitations, to provide margin against non-ductile fractures. The uncertainties introduced by l instrumentation, forced flow and elevation differences are not reflected in the TS l

curves. These uncertainties will be factored into the curves presented in the '

operating procedures.. Since administrative limits remain in place to ensure that p

10 CFR 50 Appendix G limits are not challenged, the margin of safety described ,

in the TS Bases is not reduced by the proposed change. Therefore, the change does not involve a significant reduction in a margin of safety.

Pursuant to 10 CFR 50.91, the preceding analyses provides a determination that the i proposed Technical Specifications change poses no significant hazard as delineated by 10 CFR 50.92.

Environmental Assessment '

This proposed Technical Specification change has been evaluated against criteria for and identification of licensing and regulatory actions requiring environmental

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TSP 980009-RC-99-0154 Page 3 of 3 i

' assessment in accordance with 10 CFR 51.21. It has been determined that the proposed change meets the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). The following is a discussion of how the proposed Technical Specification change meets the criteria for categorical exclusion.

.10 CFR 51.22(c)(9): Although the proposed change involves change to requirements

~

with respect to inspection or Surveillance Requirements, (i), the proposed change involves No Significance Hazards Consideration (refer to the No Significance Hazards Consideration Determination section of this Technical Specification Change Request); .

(ii) . there are no significant changes in the types or significant increase in the amounts of any effluents that may be released offsite since the proposed change does not affect the generation'of any radioactive effluents nor does it affect any of the permitted release paths; and.

(iii) there is no significant increase in individual or cumulative occupational radiation

. cxposure.

- Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Based on the aforementioned and pursuant to 10 CFR 51.22 (b), no environmental assessment or environmental impact statement need be prepared in connection with issuance of an amendment to the Technical Specifications incorporating the proposed change.

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