ML20151A332
| ML20151A332 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/30/1988 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20151A300 | List: |
| References | |
| TAC-66699, NUDOCS 8804060453 | |
| Download: ML20151A332 (21) | |
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. Docket No. 50-346 License No. NPF-3 Serial No. 1490 KIAO*0R C00 MC ST! TIM Attachment 3
. I SAT!"!T VA7 7IS - SNO"'DOW
- OC"'IM 00N "*.ON TOR OPI?.A!!ON 3.4.2 Decay Imat Removal System relief valve,DH-4849 shall be OPEPJWI .
vith a lif t. setting of .( 330 PSIG* and isolation valves DE-11 and DE-12 open and zentrol povar to their valva operators removed. AFFLICA3%LITY: WDIS 4 and 3. ACTION: A. With DH-4849 not CPERABLE: .
- 1. Maka the valve CP 'M within si.sht hours; or
- 2. a.
Withis nort one hour, disable the capability of both high prassure injection (IFI) ptmps to inj act water into the reactor cuolant system; and
- b. Within nazz aight hours 1.
Disable the automatic tranaf er of makeup pump suction
'to the borated laval; and vatar storage tank on low makaup taak 2.
laducemakeuptanklevalto473inchasandreduce rasetor coolant systaa pressura and prassuriasta , level within tha-acces. cable region on F1'gures h. (La McDE Q '3r4v2-b 1(in M:m 3). htW BM 2a 3. With D5-11 or DE-12 c.losed, open DE-21 and Da-23 vi.hia one hour. C. With the control power not removed from DE-11 and DE-12, remove the power to tha valve operators at the Motor f%nol Canters within one hour. . mu.u LANCY RZQUT'M a 4.4.2. Decay Isat tanovel System ra11af valva DE-4449 shall be datar-
=.ined CPIRAILEt
- a. par the surv.m- a requd.raments of specification 4.0.3. .
- b. at least onca par 24 hours by verif7tag althart - .
i 1. 1solattaa valvas DE-11 and DE-12 open with control power removed from their valve operators; or
- 2. valves DB-21 and DE-23 opas.
The 1.1.f t setting prassure shall corraspond to ambient ceMitions of the valve at ner.imal operating taaparature and prassura. 0804060453 880330 3/4 4-3 Amendment flo.57 PDR g ADOCK 05000346 g _ _ _ _ _ - _ - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ - - - - - _ - - - -
Dockst No. 50-346
. License.No. NPF-3 Serial No. 1490 Attachment 3 400 350 k UNACCEPTABLE REGION
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$ 300 \
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, 200 L
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100 ACCEPTABLE REGION 1 50 I , JOTE: NOT CORRECTED FOR INSTRL' MENT ERROR I I I I I I I I I
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0 40 80 120 160 200 240 Initial Pressuriter Level (Inches) j , Reactor Coolant System PresFure - Pressurizer Level Limits for inoperable Decay Heat Removal System Relief Valve in MODE 4 Figure 3.4-2a l l i l Davis-Besse L' nit 1 3/4 4-4r, Amendment No. 57
Dockat No. 50-346 l ice:he No. NPF-3 St ici No. 1490 , Att ch:xnt 3 Superseded with nao Qure 3.&2b
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R ctor Coolant Sy s t e:2 Pressure - Pressurizer 1,4 vel 1.imits for inoperabl , acay Heat Removal Systea Relief valve in .WE 5 Figure 3.4.2-b 9 #$ 8 Og 8 J . - I9 s
Docket No. 50-346 Licende No. NPF-3 serial No. 1490 Attachment 3 400 I I I I I I I NOTE: NOT CORRECTED FOR INSTRLHENT ERROR 350 E 300 _E E 5 250 w
'7' ACCEPTABLE REGION a
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I O 40 80 120 160 200 240 Initial Pressurizer level (Inches) Peactor Coolant System Pressure - Pressurizer Level Limits for inoperable Decar Heat Removal System Relief Valve in MODE 5. Figure 3.4-2b l Davis-Besse l' nit 1 3/4 4-4b Anendment No. 57
Docky Yo. 50-346
- J.icento No. NPF-3 i Seriel No. 1490 l Attachment 3 l
Rf ACT04 C0OL. ANT SYSTEM
_3/4.4.9 PRISSURE/ TEMPERATURE LINITS REACTOR COOLANT SYSTEM tlMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant Systes (except the pressurizer) terrerature and pressure shall be limited in accordance with the limit lines shown on figures 3.4-2, 3.4-3 and 3.4 4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: S0*f
- a. A mximum heatup of Fin any one hour period, and
- b. A matem coold:nn of 100*F in any one hour periodgwe'th cold let 4tmptratwt b 270*F GnA o. mMlmum Coclelovm Of SQ*M in OnY APPLICABILITY: At all tires. Cnc hour Period wim colat
!!+ ters.perMutt < 11o
- f .
ACTION: With any of the above limits exceeded, restore the ter.perature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to ,.determine
. ..... . ...,. ~.. .,.. ... of thethe Reaceffects of the out tor Coolant System;of limit condition on theiMgr/G determine i tnat the Reactor Coolant System retains acceptable for continued operation
.; or te in at least NOT STA.N;By within the next 6 hours and r; .:: 200 be in CoLb i SHV70cwN en ;. ;;.r. ;;,1;; ;m 200"T .n; l00 ;;;;, . ;;;;;;; .;ij, within
tM9following 30 hours.
SURyt!LLANCE RE0UIREMENTS
- 4. 4. 9.1.1 The Reactor Coolant System tasperature and pmssure shall be i determined to be within the Ilmits at least once per 30 minutes during systen heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.09.1.2 The reactor vessel sterial irr,adiation survet,11ance specirens ' representative of the vessel asterials shall be removed and examined, to deter:1ce changes in material properties, at the int 2rvals :5:n i ? dl: dt/ined ar, BAW 1543A. i . 1. The results of these tusinations shall be used to update Figures 3.&-2, 3.4-3 and 3.4 4. i DAvi$ 8 ESSE UNIT 1 3/4 4 24 Men 6 ment No. 81 l
ocket No. 50-346 ; L ense No. SPF-3 o i Ser al No. la90 Attac ment 3 1 i Superseded uith new 39ure 3.q-;2. 3 =
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- 1. :nce No. NPF-3 Ser 01 No. 1490 Atta e.ent 3 Superseded un# nw Ryure M-3 X
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.- Licen;e No. NPF-3 Scrici No. 1490 Attachrent 3 o
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- REACTOR VE55Et MATI RI AI. IRRADIATION SURVEll.I.ANCE SCitEDtit.E
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2 ) U First w ritest of: 1.5 EFrY; capsule fluence > a 10 I8 n/cm2-h1 est RT NDT of *a enc 8Psulated materi / equale 50F. Second Earlies
]-
si the Ilr f: 3 ErrY capsule flue e m1dumy between that and third capoules. Third Earliest of: 6 fry: capou fluence corresponde *o that N of the E01, fluence f th M esctor veneel 1/4T location. d
$ rourth Scheduie to be sutwel
,W 7 for MRC approval prior to Cycle 6 operatic * . 'D Fifth SCheduie to e sub.ittea ror NR provai prior to cycie s operatsen. '*
_y A-. l s ta th heduie to be subenitted for NRC approval pr to Cycle 6 operation. 3 I c) 4
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1 a 49 N - O) T . E. F M
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Docky No. 50-346
}.icensa No. NPF-3 D% A!b 9$#*$
h M h , )h@tWI(3l
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t n: 4 REACTOR COOLANT SYSTEM
$1MW @ MM!fp Mi$
BASES
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3/4.4.1 REACTOR COOLANT LOOPS t The plant is designed to operate with both reactor c63TaKtlo6ps in ! operation, and maintain DN8R above 1.30 during 411 normal operations and anticipated transients. With one reactor cooliat pump not in operation in i i one loop, THERMAL POWER is restricted by the klear Overpower Based on 3CS Flow and AX1AL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible 'a.ERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR i equal to 22%, whichever is mere restrictive, in MODE 3 when RCS pressure or to.perature is higher than the decay heat removal system's design conoition (i.e. 330 psig and 350'F), a , single reactor coolant loop provides sufficient heat removal capability. 1 The remainder of MODE 3 as well as in MODES 4 and 5 either a sin)1e reactor i co.lant loop or a DHR loop will be sufficient for decay heat removait but ! single failure considerations require that at least two loops be OPERABLE. - Thus, if the reactor coolant loops are not OPLRABLE, this specificition ! requires two OHR loops to be OPERABLE. ' Natural circulation flow or the operation of one OHR pump provides adequate I r flow to ensure mixing, prevent stratification and produce gradual reactivity i changes during boron concentration reductions in the Reacter Coolant System. ! The reactivity change rate associated with boron reduction will, therefors, l be within the capacity of operator recognition and control, t 3/4.4.2 and 3/4.4.3 SAFETY VALVES ! The rressur$zer code safety valves operate to prevent the RCS from being l pressurized above its Safety Limit of 2750 psig. Each safety 11ve is designed [ to relieve 336,000 lbs per hour of saturated steam at the va w's setpoint, i The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an ooerating DHR loop, connected to the RCS, i provides overpressure relief capability and will prevent RCS overpressurization, i During operation, all pressurizer code safety valves must be OPERA' ' to
- prevent the RCS from being pr.essurized above its safet) limit of f psig. !
The combined relief capacity of all of these 'ialves is greater tr the ! maximum surge rate resulting from any transient. ; I The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur dyring shu;;down, in the event that this relief valve is not OPERA 8LE, reactor coolaet system pressure, pressurizer level and make up water inventory is limited and the f capability of the high pressure injection system to inject w6ter into the [ reactor coolant system is disabled to ensure operation within reactor coolant j system pressure - temperature limits. , i, Demonstestion of the safety valves' lift settings will occur only during } shutdown and will be perfreed in accordance with the provisions of Section i XI of the ASME Boiler and Pressure Code. I, f i DAVIS.BESSE UNIT 1 B 3/4 4-1 Amendmett No. )r, 38', F ', 92 f f
Docket No. 50-346 ~
= = - =>
IlilS PAGE PROVIDED RE. C.un CCCLUl? SYSEM FORINFORMATION DNLY BASE 5 me pressurizer coce safety valves must. be sei, such that the ceak Reactor Coolant System pressure does not exceed 110% of design syste:a pressure (2500 psig) or, 2750 psig. The control rod grouo withdrawal t l accident will result in the most limiting high pressure in the RCS. The analysis assumes RPS high pressure trip at 2300 psig and the ' code safety valves open at 2500 psig. The tolerance on the RPS instrument accuracy is 30 psi and, it is +3" for the code safety valve settings. The pressurizer electromatic relief valve was assumed not to open for this transient. The resulting system peak pressure was calculated to be 2716 psig. Therefore, the code safety valve setpoint is conserva ' tively set at < 2525 ps.ig which is the maximum pressure of 2500 psig
+15 for tolera7 ice.
The pressurizer electromatic relief valve should be set such that 1t will open before the code safety valves are ooened. However, it should not open on any anticipated trans'ients. Loss of Fee 6,ater (LOFW) was identified as the limiting anticipated transient for RCS pressure. I'1e analysis assunes RPS high pressure trip at 2300 psig; with 30 psi for instrument errors , the resulting peak RCS pressure is calculated te Le 1380 psig. This includes a 50 psig pressure overshoot on a LORf transient. ADDlil0flAL CHANCES PREVl0U$lY PROPOSED BY LETTER S i__I IY YY-. Date&88 8 4 4* O e mh e e lAV!S-5EISE, 'JNIT 1 3 3/4 1-13 Amer.cment n M ,60
Docket No. 50-346 hicense No. NPF-3
=w THIS PAGE PROVIDED FORINFORMATION DNLY REACTOR COOLANT SYSTEM _
BASES The ACTION statement pennitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accermodates possible iodine spiking phenomenon which may occur following changes in THERF.AL POWER. Operation with specific ac-tivity levels exceeding 1.0 uCi/ gram DOSE EQUIVALENT I-131 but within the limits snown on Figure 3.4-1 must be restricted to no more than 10 percent of the units yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube nJpture. Reducing T to 4 S30'F prevents the release of activity should a steam generatori dbe rupture since the saturation pressure of the primary coolant is below the lif t pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be de-tected in sufficient time to take corrective action. Infomation obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses follow-ing pcwer changes may be pennissible if justified by the data obtained. ! 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to 10 CFR 50 and with the themal and loading cycles used for design purposes. l The limitations prevent non-ductile failure during normal operation. I including anticipated tperational occurrences and system hydrostatic tests. The limits also prevent exceeding stress limits during cyclic operation. The loading conditions of interest include: 1 1. Nonnal operations, including heatup and cooldown,
- 2. Inservice leak and hydrostatic tests, and
- 3. Reactor core operation.
The major ccenponents of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. The closure l head region, reactor vessel outlet nozzles and the beltline region have ' been identified to be the only regions of the reactor vessel, and con-sequently of the reactor coolant pressure boundary, that determine the ! pressure-temperature limitations concerning non-ductile failure. DAV!S-BESSE. UNIT 1 8 3/4 4-6 1
ockeg No. 50-346 L cense No. NPF-3 Se. al No. 1490 Atta hment 3
/ ejeled g )AgeikbhbO51df e 14.0 E Extrapolation past 10 EFPY baud on repeat of E equilitwium fuel cycle
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't I /! 18 24 32 EXPOSURE ' TIME, EFFECTIVE FULL POWER YEARS Bases Figurs 4-1 Fut Neutron Fluence (E > 1 Mw) as a Function \T of Full Poww Service t.ife DAVIS-BESSE, UNIT 1 8 3/4 4-7
( Rey 4 pue is B 3/'l 4 - 10)
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B m s 4 rs 4 2 Effect of Fluence arid Copper Content on Shift ofd N bT b Reactor Vessel Stoefs Exposed to 6600F Temperatues s l DAVIS-BESSE, UNIT 1 B 3/4 4 8
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Docket.No. 50-346 1,1c'ense No. NPF-3 Serial No. 1490 Attachment 3 REACTOR COOLANT SYSTEM BASES The closure head region is significantly stressed at relativel temperatures (due to mechanical loads resulting from bolt pre-load)y . This low region largely controls the pressure-temperature limitations of the first several service periods. The outlet nozzles of the reactor vessel also affect the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle which can be two to three times the membrane stresses of the s h e *.1. After the first several years of neutron radiation exposure, the RT temperature of the beltline region materials will be high enough'qN that the beltline region of the reactor vessel will start to control the pressure-temperature limitations of the reactor coolant pressure boundary. For the service period for which the limit curves are established, the maximum allcwable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region. The maximum allowable pressure is taken to be the lower pressure of the three calculated pressures. The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the , limiting component for all operating reactor coolant pump combinations. i The limit curves were prepared based upon the most limiting adjusted ! reference temperature of all the beltline region materials at the end of l 18 O the'fiftte effective full pcwer year. *h: fth ";;ti;; f;1' ; r .,: r
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The actual shift in RT of the beltline region material will be established periodically dung operation by removing and evaluating, in i accordance with Appendix H to 10 CFR 50, reactor vessel material irradia-tion surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiat. ion samles and vessel inside the radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent'section of the reactor vessel. The limit curves must be recalculated when the ART detemined frca the surveillance capsule is different from the calculbd ARTNOT for the equivalent capsule radiation exposure. l OAVIS-BESSE, UNIT 1 B 3/4 4-11 t
P Docket No. 50-346 License No. NPF-3 Serial No. 1490 Attachment 3 REACTOR COOLANT SYSTEM BASES The unirradiated transverse impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available. Th; .nirredicted imp::t ; pertie: :nd residual cis.rt: Of the beltline-r:gi:n a terials are listed in Oeses Tebl. 4-1. The adjusted reference temperat. ares are calculated by adding the predicted radiation-induced
$RT and the unirradiated RT ar: ::10:1:t:d ut i the era ^ctive aeutrea Mn.6 ceTh.:
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Figure 3.4-2 presents the pressure-temperature limit curve for normal heatup. This figure also presents the core criticality limits as required by Appendix G to 10 CF8 50. Figure 3.4-3 presents the pressure- t temperature limit curve for normal cooldown. Figure 3.4-4 presents the pressure-temperature limit curves for heatup and cooldown for inservice leak and hydrostatic testing. - 4edh All pressure-temperature limit curves are applicable up to the f4fe effective full power year. The protection against non-ductile failure is assured by maintaining the coolant pressure below the upper limits of Figures 3.4-2, 3.4-3 and 3.4-4.
% pnx.tAures descnkd in Tey \by Gui Ae I .% ,% 1.,
were usea for prealeMn3 % radiah inhced 6RTnst as a. . hncSM Ok Mic Matedal'r Coy DAVIS-BESSE, UNIT 1 B 3/4 4-11 andshket cated and neub 2lucace. c
Docky No. 50-346 Li 6'e n s e. No. NPF-3 Seria2 3. 1490 - Attach. cat 3 1 AEACTOR COOLANT $YSTD4 l I sAsrs The nunber of nactor vessel teradiation surveillance specimens and the frequencies for removing and testing these spect, ens are provided in BAW IF%A Teti; 4.04.: The withdrawal schedule is based on four considerations: (a) uncover possible technical anomalies as early in life as they can be detected (end of first fuel cycle), (b) define the material properties needed to perfors the analysis required by Appendix G to 10 CFR 50, (c) reserve two capsules for evaluation of the effectiveness of thermal annealing in the event ee-inplace annealing becomes necessary, (d) provide matarla1 property data corresponding to the reactor vessel belt l
.line 2:-f:::- conditions at the end of service. ne withdrawal schedule et Teti; i.4-4-is specified to assure compliance vich the requirements of Appendix H to 10 CTR 30. Appendix H ref erences the requirements of ASTM E185 for surveillance prosram criteria. Tet!; i.1-5 1; n isa;4
- ;.u : :he ::pirum:; ;f /5~ E185 - 22.
l l i l MYIS-BESSEi. UNIT 1 B 3/4 4-12 knendinent No. 81}}