ML18010B051
| ML18010B051 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/26/1993 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18010B050 | List: |
| References | |
| NUDOCS 9303090318 | |
| Download: ML18010B051 (31) | |
Text
ENCLOSURE 5 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS TECHNICAL SPECIFICATION PAGES 9303090318 930226 PDR ADQCK 05000400 P PDR
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4 ~ 4~9 PRESSURE/TEMPERATURE LIMITS Reactor Coo1anc Syscem..........................,.....,... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO II FPY.................................... 3/4 4-35 I FIGURE 3.4 "3 REACTOR COOL SYSTEM HEATUP LIMITATIONS-APPLICABLE UP TO II FPY...........................,...,... 3/4 4-36 I TABLE 4.4"5 DELETED............................................... 3/4 4-37 TABLE 4 ~ 4-6 MAXIMUM COOLDOWN AND HEATUP RATES FOR MODES 4, 5 AND 6 (WITH REACTOR VESSEL HEAD ON)....................... 3/4 4-38 P res suri zer ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-39 Overpressure Protection Systems........................... 3/4 4-40 fIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE SYSTEM................,.....,...,,......,..., 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY ~ ~.................................... 3/4 4-43 3/4 ~ 4~ 11 REACTOR COOLANT SYSTEM VENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-44 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO F ~ ~ ~ ~ ~ 3/4 5-3 3/4.5 ' SUBSYSTEMS - T LESS THAN 350'Fo.............
350'CCS 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 5-9 SHEARON HARRIS - UNIT 1 V111 Amendmenc No.~
280 w MQLAw<$ c /AT Lg t ~ c I>TPT<.'QqI$
~ ning
.APP': '2i = 9 3 5 -"= Y 2600 NIATER!AL PERT;Y 3AS ES Controilinc oct ai Plate 94197-2 2 '.00 Cooper Con'tent 0.107'.SOFAS Nickel Content Regulatory Guide 1.99, Rev. 2 2200 RT qpT <nttlal 86'F RTqpT at 'z 4 T 1 75'F RTqpT at 3/4 T 90 F
~
2000 Instrument Error Aoplied =~10,. and -60 psig l800
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0 I
,1200 I Cl C3 I C5 Z:
i I I I
800 I 50'/HR I
>0'/HR i
600 5'/HR I 20'R 400 2oo 100 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F R<pJCLQC. Ldi44 RGURE 3.4-2 REACTOR COOIANT SYSTEM pic.
COOLDOWN LIMITATIONS APPLICABLE UP EFPY TOP SHEARON HARRIS UNIT 1 3/4 4-35 Amendment No. ~'
2800 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 11 EFPY 2600 ELH MATERIAL PROPERTY BASES:
Controlling Material = Plate A9153-1 2400 Copper Content 0.09K Nickel Content 0.45K Regulatory Guide 1.99, Rev. 2 RTNpT Initial 60'F 2200 RTNpT at 1/4 T = 148'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1600 1400 1200 Ci LLI ABOVE 2 20'F, SIN GLE CURVE FO L RATES. I 50'F /HR 1000 o C5 20'F/HR 800 IO'F/HR SF/ HR 600 400 200 100 140 180 220 260 300 340 , 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOIfN UMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-35 Amendment No.
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"-'. 00 Controj inc va ia. Plate 84197-2 Copper Content 0.107e
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,'Iickej Content 0.507'.99, Regujatory Guide Rev. 2 2200 RTNpT:nitial 86F RTNpT at '>~ 4 T 175'F
! RTNpT at 3/4 T 9F ,2000 I i fnstrurnent Error Applied +1 and -op psig .
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'00 l 10'/HR I I
0'/HR l 400 I
'200 I
100 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES F pepjaca, Wj(pj FIGURE 3.4-3 REACTOR COOLANT SYSTEM Nej!I pge HEATUP LIMITATIONS APPUCABLE UP TO'FPY I SHEARON HARRIS UNIT 1 3/4 4-36 Amendment No. ~
2800 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS ISLH APPLICABLE UP TO 11 EFPY 2600 10'F/HR MATERIAL PROPERTY BASES:
Controlling Material = Plate A9153-1 00'F/H 2400 Copper Content 0.097e Nickel Content 0.45K Regulator Guide = 1.99, Rev. 2 2200 RTNpT Initial 60'F RTNpT at I/4 T = 148'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program. 1800 Plant Procedure, PLP-106.
1600 I 1400 1200 C) 50'F/ HR 1000 o C5 30'F /HR Z 800 20'F/ HR 10'F /HR 600 400 200 100, 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP UMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-36 Amendment Na.
TABLE 4.4-6 MAXIMUM COOLDOWH AND HEATUP RATES FOR MODES 4, 5. AND 6 (WITH REACTOR VESSEL HEAD ON)
'COOLDOWN RATES TEMPERATURE"" COOLDOWN IN ANY 1 HOUR PERIOD g5O-(55'F 50'F isa-i35'F 50-175'75-1
'F 20'F J35-H5 P 14 30'F 10'F
< II5op. 5'F Ao'F" HEATUP RATES 130'EMPERATURE'""
HEATUP IN ANY 1 HOUR PERIOD
<ISA'P < 140" F 10'F 135-(50 F 140 1 "F 20'F 155 90"F 30'F l'15-350'F 1 3 50 o
50'F emperacure used should be based on lowest RCS cold leg value except when no I RCP is in operation; then use an operacing RHR heat exchanger outlet tern erature.
lo'6/HR Cooldo+n ra4e rnoybc, cased iP less Ann ehreeRQI s ore, oisercdiag.
SHEARON HARRIS UNIT 1 3/4 4-38 Amendment No. ~
0 0
400 I
0 LL I
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lL MEASURED RCS TEMPERATURE PF)
RCS TEMP LOW PORV '~ HIGH PORV OF PSIC 0 Q~ Q
""'SIC 90 370 380 100 370 380 125 400 410 250 400 410 300 427 437 325 440 450
- VALUES BASED ON 5 EFPY REACTOR VESSEL DATAf INSTRUMENT ERRO S ARE coHTRygggD gY PCClFICATIo> EAU(PMENTLIS7 PRO~PAML, Pl
~E T&gpuicAt.
AWAIT PROCEDuRE PLP>>4 ~
FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendmenr. No. ~'
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) disc inccion becwe n the radionuclides above and below a haLE-life of 15 minutes.
For these reasons che radionucl.ides char. are excluded from conside.acion are expecced to decay to very Low Level.s beEore they could be transpor"ed :rom che reactor cool.ant co che SITE BOUNDARY under any accident condition.
Based upon the above considerations Eor excluding certain radionuclides from the sample analysis, the all.owable time oE 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample caking and complecing che initial anaLysis is based upon a typical. rime necessary co per-Eorm the sampling, transport che sample, and perform che anal.ysis oE about 90 minutest AEter 90 minuces, the gross count shoul.d be made in a reproducibLe geomecry of sample and counter having reproducible beta or gamma ~elf-shielding properties. The counter should be resec co a reproducibl.e efficiency versus energy. 'It is not necessary to identify specific nuclides. The radiochemicai determination oE nuclides should be based on multiple councing oE the sample within typical counting basis Eollowing sampling of Less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.
Reducing T to Less than 500'F prevents the release oE activity should a steam generator tube rupture occur, since the saturacion pressure of the reactor cool-ant is below che 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 cime to take corrective action. A reduction in frequency oE isotopic analyses EolLowing power changes may be permissibLe if justified by che data obtained.
3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are Limited co be c nsxstenc with the requirements given in the ASME Boiler and Pressure Vessel.
Code, Section III, Appendix G, and 10 CFR 50 Appendix G and H. 10 CFR 50, Appendix G also addresses the metal temperature of the closure head El.ange and vesseL ELange regions. The minimum metal temperature oE the closure Elange region should be at Least 120'F higher than the limiting RT NDT Eor these regions when the pressure exceeds 20X (621 psig Eor Westinghouse plants) oE che preservice hydrostatic test pressure. For Shearon Harris Unic 1, the minimum temperature of the closure ELange and vessel flange regions is 120'F because che limiting RT NDT is O'F (see Table B 3/4 4-1). The Shearon Harris Unic cooLdown and heatup Limitations shown in Figures 3.4-2 and 3.4-3 and Table 4.4-6 are not impacted by the 120'F Limic.
- 1. The reactor cooLant temperature and pressure and system cooldown and heatup rates (with the excepcion of che pressurizer) shall be Limiced in accordance with Figures 3.4-2 and 3.4-3 and Table 4.4-6 Eor the service period specified thereon:
a ~ Allowable combinations of pressure and temperature for specific temperature change rates are below and to che right of che Limic lines shown. Limit L ines Eor cooldown rates becween those pre-sented may be obtained by interpolation', and SHEARON HARRIS - UNIT 1 B 3/4 4-6 Amendmenc No. A
REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE L.'MITS (Continued)
- b. , Figures 3.4-2 and 3.4-3 define limits to assure prevention oi non-ductile failure only. For normal operation, other inherent plant charactezistics, e.g., pump heat addition and pzessuri er hearez capaciry, may limit tne heatup and cooldown zates that can be achieved over certain pressure-temperature ranges.
- 2. These Limit Lines shall be calculated peziodically using methods pro-
.vided below,
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam gene'razor is below 70'F, 4 ~ The pressurizer heatup and cool. down rates shaLL not exceed 100'F/h and 200'F/h, respectively. The spray shall not, be used if the tem-perature difference between the pressuri'ter and the spray fluid is gzeatez chan 62S'F, and 5..System preservice hydzotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements oz ASME Boiler and Pressure Vessel Code,Section XI.
The fzacture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1971 Winter Addenda to Section III oi the ASME Boiler and Pressure Vessel Code. These properties are then evaluateo i'n accordance with the NRC Standard Review Plan.
Heatup and cooldown Limit curves are calculated using the most, limiting value of the nil-ductiLity reference temperatuze, RTN>T, at the end of~S~ffective I ll full power years (EFPY) of service Life. The service life period zs chosen such that the limiting RTNOT at the 1/4T location in the core region is greater chan the RTNOT of the Limiting unirradiated material. The selection of such a Limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance wizh applicable Code requirements ~
The reacroz'essel matez ials have been tested to determine theiz initial RTN0T:
the results of these cests are shown in Table B 3/4.4-1. Reactoz operation ana resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon che fluence, copper content, and nickel content of the marerial in question, can oe predicced using Figure B 3/4.4-1 and the vaLue of aRTN>T computed by Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Matez ials."
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,include SHEARON HARRIS UNIT 1 B 3/4 4"7 Amendment No.
rABLE 8 3/4.4-1 REACTOR VESSEL TOUGHNESS CHARPY uPPER
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HEAT 5m HtM T N)T WD Jl COHPONENT CRADE NO ~WX too.X) ("F) IT-0 t'r-i n Closure Hd. Dome A533,B,CLl A9213-1 -10 114 Head Flange A508, CL2 5302-V2 l35 Vessel Flange 5302-Vl -10 llo inlet II Nozzle II 4388-4 "20 -20 lb9 438B-5 0 0 128 II II 4388-6 -20 -20 149 Outlet Nozzle 4398-4 -10 -10 151 II II 4398-5 -10 -10 152 4398-6 -10 -10 l50 Nozzle Shell A533B,CLl C0224-1 -20 -1 >>J 0 II II C0123-1 0 42 84 Inter. Shell II A9153-1 .09 -10 60 II 84197-2 . 10 .SO -10 91 Lower Shel 1 C9924-1 . 08>>~~ A&5 -10 54 C9924-2 . 08- .d4 A&5 -20 57 Bottom Hd. Torus A9249-2 -40 14 94 DoII)e A9213-2 -40 -8 l25 Weld (inter & Lower Sh<<l l Vert ical WI.ld Seams) ~>>)4 Wold ( IIII<<r. t u l.uwe r Sll << I I L I I t II builfll) .o4 .9> -'u "20
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REACTOR COOLANT SYSTEM BASES al-e, Ease J upon an ad'usted RTNDT PRESSURE/TEMPERATURE LIMITS (Conc inued)
The cooidown and heacup Limits oE Figures 3.4-2 and 3.4-3 predicted adjustmencs for this shiEt in RTNDT plus marjln)
Xn accord~co ~ith ~lee g Gula< l.99> Revision 2, the resul.ts from the materiaL surveil.lance program, evaluated according to ASTM E185, are available. Capsules wilL be removed and evaluated in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H. The results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel materiaL by using che Lead factor and che wich-drawal time of the capsuLe. The cooldown and heatup curves must be recalculated when the 4RTNDT determined Erom the surveiLLance capsule exceeds che calculated 4RTNDT Eor the equivalent capsule radi.ation exposure.
Allowable pressure-temperature relationships Eor various cooLdown and heacup rates are caLcuLated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure VesseL Code as required by Appendix G to 10 CFR Part 50.
The general method Eor calcuLating heatup and cooldown Limit curves is based upon the principles of the linear elastic fracture mechanics (LEFH) technology.
In the calculation procedures a semielliptical surface deEect with a depth of one-quarter oE 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 oE this postulated crack, referred to in Appendix G oE 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 sufEicient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of che Limit curves, the most limiting value oE the nil"ductility reference tempera-re, RTNDT~ is used and this includes the radiation-induced shift, 4RTNDT
'orresponding to the end of the period Eor which cooLdown and heatup curves are generated The ASHF approach Eor calculating the alLowable limit curves Eor various heacuo and cooldown rates specifies that the rotal stress intensicy Eactor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor; KIR, Eor che
>ay bc, used a determine hR~~pv
~pen ~o or Kore. sets oF arHib(~ Surveillance da6-SHEARON HARRIS - UNIT 1 8 3/4 4-11 Amendment No.
. ~
BASES PRESSURE/TEMP ERATURE LIMITS Continued )
metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix 6 to the ASME Code. The K<< curve is given by the equation:-
KIR 25.78 + 1.223 exp t.0.0145(T-RTNO + 160)]
Where: KIR is the reference stress intensity factor as a.function of the metal temperature T and the metal nil-ductility reference temperature RTNOT Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C KrM KIt < KIR (2)
Where: K>M
= the stress'intensity factor caused by membrane (pressure) s-r s-,
'KI = the stress intensity factor caused by the thermal gradients, K<R
= constant provided by the Code as a function. of temperature relative to the RTNOT of the material, C "- 2.0 for level 8 service limits, and lgstk and C = 1.5 far insersice njjdrastetic test aoeretian P'wi&naisslln~
Mc,rcat:eoi vcssc SsL N~ l't any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT>OT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity fac.or, KIT, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
K COOLOOlSN 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. Ouring cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile s resses 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.
SHEARON HARRIS - UNIT 1 S 3/4 4-L?
Hocuaver, Qc Q~~ and coolgoen curves in plant'perot'ng procedures have been ad(usted REACTOR COOLANT SYSTEM Qr +gga jnsfrumcnt errors. 7'nstrunan+ errors ore mrHroilcd by the ~eehn<cai ~PeciRoa+on Procedure P+ l Equipment i is~ Program, Plant BASES PRESSURE/TEMPERATURE LIMITS (Continued) heatup and che time (or coolant temperature) along the hearup ramp. Furcher-more, since the chermal stresses at the outside are tensiLe and increase <<ich increasing heatup rate, a Lower bound curve cannot be defined. Racher, each heatup race of interesc must be analyzed on an individual basis.
Following the generation oE pressure-temperature curves for both che steady-state and Einite heacup rare situations, che final. Limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison oE the steady"state and finice hearup rate data. At any given temperature, che allowabLe pressure is taken co be che Lesser of the three values taken Erom che curves under consideration.
The use of the composite curve is necessary co set conservative heacup Limira-tions because it is possible Eor condicions to exist such that over the course oE the heatup ramp the controlling condition switches from the inside co the outside and the pressure limit must ac all times be based on anaLysis of che most critical criteri in Fiyrcs 3A-2. ard 9 'l-3'ave, no+been e composite curves or t e eatup rate data and the cooLdown rate ac @~adjusted Eor possible errors in the pressure and tempe ature sensing instrument Al.though the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating l.imits are provided to assure compatibility of operation with the Eati ue anal sis erformed in accordance with the ASME Code requirements'/S/Jf>> pr s vQ Q$ +f Ar inscrviec leak tand ph~drostntic,
$ p tests dhcnno LOW TEMPERATURE OVERPRESSURE PROTECTION ~i AN Aereactor Head. Where<'sc, normal fieutup and cooldooon P-7 curvcS o.ppi The OPERABILITY of two PORVs or an RCS vent opening oE at least 2.9 square inches ensures that the RCS will be protected from pressure transients which could exceed the Limits of Appendix C to 10 CFR Part 50 when one or more of che RCS coLd legs are less than or equal to 325'F. Either PORV has adequate relieving capability to protect che RCS Ecom overpressurization when the tran-sient is limited to either: (1) the start oE an idle RCP with the secondary water temperature of the steam generator Less than 50'F above the RCS cold Leg temperatures, or (2) the start of a charging/safety injection pump and its injection'nto a water"solid RCS.
The maximum al.lowed PORV setpoint Eor the Low Temperacure Overpressure Protec-tion System (LTOPS) is derived by analysis which model.s the performance of che LTOPS assuming various mass inpuc and heat i,nput transients'peration with a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix C criteria wiLL not be violated with consideration Eor a maximum pressure over-shoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, inscrument uncertainties, and single failure. To sure that mass and heat in u transien s more severe chan chose pGguipm~t OI lAsf~eVH gnCerkainfies pre Con<<oiled bf 'lhe QcVlniekl Sp<cifl~on
- lO4 SHEARON HARRIS - UNIT 1 4S+ pr~~rn plant Proacdurc.
B 3/4 4-14 F'LP ~
Amendment No. ~
REACTOR COOLANT SYSTEM BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) assumed cannot occur, Technical Specifications require Lockout of all but one charging/safety injection pump awhile in MODES 4 (below 325'F), 5, and 6 with the reactor vesseL head instaLLed and disalLow start of an RCP if secondary temperature is more than SO'F above primary temperature.
The maximum aLLowed PORV setpoint for the LTOPS wil.l be updated based on the results of examinations of reactor vesseL material irradiation surveiLLance specimens performed as required by 10 CFR Part SO, Appendix H, .and Qjg d
the reactor 3/4.4.10 STRUCTURAL INTEGRITY yeSse I service, liFe ~
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 Level throughout the Life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicabLe Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commis-sion pursuant to 10 CFR 50.55a(g)(6)(i).
Components of the Reactor Coolant System were designed to provide access co permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1977 Edition and Addenda through Summer 1978.
3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor CooLant System that couLd inhibit natural cir-culation core cooling. The OPERABILITY of Least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function.
The valve redundancy of the. Reactor Coolant System vent paths serves to minimize the probabiLity of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isoLation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.L of NUREC-0737, "Clarification of TMI Action Plant Requirements," November 1980.
SHEARON HARRIS - UNIT 1 B 3/4 4-15 Amendment No. ~
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~ INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System.................................. 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS-APPLICABLE UP TO 11 EFPY................................ 3/4 4-35 FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS-APPLICABLE UP TO 11 EFPY................................ 3/4 4-36 TABLE 4.4-5 DELETED......................................'.... 3/4 4-37 TABLE 4.4-6 MAXIMUM COOLDOWN AND HEATUP RATES FOR MODES 4, 5 AND 6 (WITH REACTOR VESSEL HEAD ON)..................... 3/4 4-38 P ressurizer.................................... ..... 3/4 4-39 Overpressure Protection Systems......................... 3/4 4-40 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE SYSTEM............. ....................... 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY..............;.. 3/4 4-43 3/4.4. 11 REACTOR COOLANT SYSTEM VENTS...................... 3/4 4-44 3 4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS...................................... 3/4 5-1 3/4.5.2, ECCS SUBSYSTEMS - T, GREATER THAN OR EQUAL TO RhO I C'%IV
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T, LESS THAN 350'F............ 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK...................... 3/4 5-9 SHEARON HARRIS - UNIT 1 V111 Amendment No.
2800 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 11 EFPY 2600'ATERIAL PROPERTY BASES:
Controlling Material = Plate A9153-1 2400 Copper Content 0.09K Nickel Content 0.45K Regulatory Guide 1.99, Rev. 2 RTNpT Initial 60'F 2200 RTNpT at 1/4 T = 148'F RTNpT ot 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1600 LIJ 1400 1200 ABOVE 2 20F, SING LE C URVE FO R ALL RATES.
50'F /HR 1000 o D
20'F/HR Z 800 10'F/HR 5'F/ HR 600 400 200 100 140 180 220 . 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN UMITATIONS APPLICABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-35 Amendment No.
2800 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS ISLH APPLICABLE UP TO 11 EFPY 2600 10'F/HR MATERIAL PROPERTY BASES:
Controlling Material Plate A9153-1 00'F/H 2400 Copper Content 0.09%
Nickel Content 0,45%
Regulatory Guide 1.99, Rev. 2 2200 RTNpT Initial 60'F RTNpT at I/4 T = 146'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1600 I LIJ 1400 P) 1200 C5 50'F/ HR 1000 o a
30'F /HR 800 20'F/ HR IO'F/HR 600 400 200 100 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-36 Amendment No.
MAXIMUM COOLDOWN AND HEATUP RATES FOR MODES 4 5 AND 6 WITH REACTOR VESSEL HEAD ON COOLDOWN RATES TEMPERATURE* COOLDOWN IN ANY 1 HOUR PERIOD*
350-155'F 50'F 155-135'F 20'F 135-115'F 10'F
< 115'F 5'F/10'F**
HEATUP RATES TEMPERATURE* HEATUP IN ANY 1 HOUR PERIOD*
<135'F 10'F 135-150'F 20oF 150-175'F 30'F 175-350'F 50oF
- Temperature used should be based on lowest RCS cold leg value except when no RCP is in operation; then use an operating RHR heat exchanger outlet temperature.
- 10'F/HR cooldown rate may be used if less than three RCPs are operating.
SHEARON HARRIS - UNIT 1 3/4 4-38 Amendment No.
t 500 U
400 0 ) I ) !
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I ) ) t ) ) i 100 200 300 400 MEASURED RCS TEMPERATURE (V)
- HIGH PORV
- VALUES BASED ON ll EFPY REACTOR VESSEL DATA.
INSTRUMENT ERRORS ARE CONTROLLED BY THE TECHNICAL SPECIFICATION E(UIPHENT LIST PROGRAM, PLANT PROCEDURE PLP-106.
FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendment No.
REACTOR COOLANT SYSTE BASES SPECIFIC ACTIVITY Continued distinction between the radionuclides above and below a half-life of 15 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.
Reducing T,, to less than 500'F prevents the release of activity should a steam generator tube rupture occur, 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. A reduction in frequency of isotopic analyses following power changes may be permissible the data obtained.
if justified by 3 4.4.9 PRESSURE TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G, and 10 CFR 50 Appendix G and H.
10 CFR 50, Appendix G also addresses the metal temperature of the closure head flange and vessel flange regions. The minimum metal temperature of the closure flange region should be at least 120'F higher than the limiting RT NDT for these regions when the pressure exceeds 20% (621 psig for Westinghouse plants) of the preservice hydrostatic test pressure. For Shearon Harris Unit 1, the minimum temperature of the closure flange and vessel flange regions is 120'F because the limiting RT NDT is O'F (see Table B 3/4 4-1).
The Shearon Harris Unit-1 cooldown and heatup limitations shown in Figures 3.4-2 and 3.4-3 and Table 4.4-6 are not impacted by the 120'F limit.
The reactor coolant temperature and pressure and system cooldown and heatup rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 and Table 4.4-6 for the service period specified thereon:
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 by interpolation; and SHEARON HARRIS - UNIT 1 8 3/4 4-6 Amendment No.
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REACTOR COOLANT SYST BASES PRESSURE TEMPERATURE LIMITS Continued
- b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure'nly. 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 provided below,
- 3. The secondary side of the steam generator must not be pressurized above 200 psig 70'F if the temperature of the steam generator is below
- 4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 625'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.
The fracture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code. These properties are then evaluated in accordance with the NRC Standard Review Plan.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT~ at the end of ll effective full power years (EFPY) of service life. The service life period is chosen such that the limiting RT~, at the 1/4T location in the core region is greater than the RT>>, of the limiting unirradiated material. The selection of such a limiting RT>> assures that all components in the Reactor Coolant. System will be operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RT~.
the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT>>. Therefore, an adjusted reference temperature, based upon the, fluence, copper content, and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the value of ART~
including margin, computed by Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."
SHEARON HARRIS - UNIT 1 B 3/4 4-7 Amendment No.
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TABLE B 3 4.4-1 REACTOR VESSEL TOUGHNESS CHARPY INITIAL UPPER SHELF ENERGY HEAT Cu Ni Tg)T RTg)T TRANSVERSE COMPONENT GRADE NO ~wt.% ~wt.% ~oF ~oF FT-LB Closure Hd. Dome A533,B,CL1 A9213-1 -10 8 114 Head Flange A508, CL2 5302-V2 135 Vessel Flange 5302-V1 -10 -8 110 Inlet Nozzle 438B-4 -20 -20 169 438B-5 0 0 128 I
438B-6 -20 -20 149 CO Outlet Nozzle 4398-4 -10 -10 151 439B-5 -10 -10 152 4398-6 -10 -10 150 Nozzle Shell A533B,CL1 C0224-1 .12 -20 -1 90 C0123-1 .12 0 42 84 Inter.
II Shell II A9153-1 .09 .45 -10 60 83 B4197-2 .10 .50 -10 91 71 Lower Shell C9924-1 .08 .45 -10 54 98 II C9924-2 .08 .44 -20 57 88 Bottom Hd. Torus A9249-2 -40 14 94 Dome A9213-2 -40 -8 125 Weld (Inter & Lower Shell O
Vertical Weld Seams) .06 .91 -20 -20 )94 Weld (Inter. to Lower Shell Girth Seam) .04 .95 -20 -20 88
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REACTOR COOLANT SYSTE BASES PRESSURE TEMPERATURE LIMITS Continued The cooldown and heatup limits of Figures 3.4-2 and 3.4-3 are based upon an
- adjusted RT~, (initial RT~, plus predicted adjustments for this shift in RT plus margin).
In accordance with Regulatory Guide 1.99, Revision 2, the results from the material surveillance program, evaluated according to ASTM E185, may be used to determine hRT when two or more sets of credible surveillance data are available. Capsules will be removed and evaluated in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H. The results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The cooldown and heatup curves must be recalculated when the h,RT~, determined from the surveillance capsule exceeds the calculated d RT~, for the equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various cooldown and heatup rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
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 semielliptical 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 nonductile 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, RT~ is used and this includes the radiation-induced shift, ERT~ corresponding to the end of the period for which cooldown and heatup 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 for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K,, for the SHEARON HARRIS - UNIT 1 B 3/4 4-11 Amendment No.
e REACTOR COOLANT SYS BASES PRESSURE TEMPERATURE LIMITS Continued metal temperature at that time. Kg is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K,R curve is given by the equation:
K,= 26.78 + 1.223 exp I'0.0145(T-RT~, + 160)] (1)
Where: K, is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RT. Th'us, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C KiM+ Ki, a KiR (2)
Where: K, = the stress intensity factor caused by membrane (pressure) stress, K = the stress intensity factor caused by the thermal gradients, Kg - constant provided by the Code as a functi on of temperature relative to the RT>>, of the material, C 2.0 for level A and B service limits, and C - 1.5 for inservice leak and hydrostatic (ISLH) test operations with no fuel in the reactor vessel.
At any time during the heatup or cooldown transient, K, is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT~ and the reference fracture toughness curve. The thermal stresses 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. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
COOLDOWN For the calculation of"the'llowable 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't 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.
SHEARON HARRIS - UNIT 1 B 3/4 4-12 Amendment No.
REACTOR COOLANT SYST BASES PRESSURE TEMPERATURE LIMITS Continued heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of interest must be analyzed on an individual basis.
Following the gener ation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary 'to set conservative heatup limita-tions 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.
The composite curves for the heatup rate data and the cooldown rate data in Figures 3.4-2 and 3.4-3 have not been adjusted for possible errors in the pressure and temperature sensing instruments. However, the heatup and cooldown curves in plant operating procedures have been adjusted for these instrument errors. The instrument errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.
"ISLH" pressure-temperature (P-T) curves may be used for inservice leak and hydrostatic tests when no fuel is in the reactor vessel. Otherwise, normal and cooldown P-T curves apply. 'eatup Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of oper ation with the fatigue analysis performed in accordance with the ASME Code requirements.
LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.9 square inches ensures that'he 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 325'F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (I) the start of an idle RCP with the secondary water temperature of the steam generator less than 50'F above the RCS cold leg temperatures, or (2) the start of a charging/safety injection pump and its injection into a water-solid RCS.
The maximum allowed PORV setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance SHEARON HARRIS - UNIT I B 3/4 4-14 Amendment No.
REACTOR COOLANT SYS BASES LOW TEMPERATURE OVERPRESSURE PROTECTION Continued of the LTOPS assuming various mass input and heat input transients.
Operation with a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. LTOP instrument uncertainties are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106. To ensure that mass and heat input transients more severe than, those assumed cannot occur, Technical Specifications require lockout of all but one charging/safety injection pump while in MODES 4 (below 325'F), 5 and 6 with the reactor vessel head installed and disallow start of an RCP secondary temperature is more than 50'F above primary temperature.
if The maximum allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and the reactor vessel service life.
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 level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 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 Boiler and Pressure Vessel Code, 1977 Edition and Addenda through Summer 1978.
3 4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core. cooling. The OPERABILITY of least one Reactor Coolant System vent path from:,,the r'eactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737,"Clarification of TMI Action Plant Requirements," November 1980.
SHEARON HARRIS - UNIT 1 B 3/4 4-15 Amendment No.