ML19206A549

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Certifies Tech Specs Change Request 14 to App a of Ol,Filed W/Nrc & Served on Chief Executives of Londonderry Township, Dauphin County,Pa by Us Mail
ML19206A549
Person / Time
Site: Crane 
Issue date: 07/07/1978
From: Herbein J
Jersey Central Power & Light Co, Metropolitan Edison Co, Pennsylvania Electric Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML19206A540 List:
References
NUDOCS 7904200129
Download: ML19206A549 (18)
v • d • e


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.A. 'larga frca J. H. Taylor, dated June 7,1979 9 -VJ f L sh ts <

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i

TABLE 2.2-1 (Continued) q g REACTOR PROTECTI0ft SYSTEM IllSTRUMEllTATIO?! TRIP SETP0f fits t'1 ~ 5 FutiCTIO!l UtlIT TRIP SETPolflT_ ALLOWABLE VALUES fluclear Overpower ' < 125% of RATED TilERMAL POWER < 125% of RATED TilERMAL POWER 8. based on Pump Monitors (j) with three pumps operating with three pumps operating # )^ o < 56.9% of RATED TilERMAL POWER s 57.18% of RATED TilERMAL POWER ~ Uith one pump operating in each loop Uith one pump operating in each loopf h < 0% of RATED THERMAL POWER with < 0.E8% of RATED TilERMAL POWER with two pump operating in one loop and two pumps operating in one loop and no pump operating in the other loop no pump operating in the other loop # g m < 0% of RATED TilERMAL i>0WER with < 0.28% of RATED TilERMAL POWER with no pumps operating or only one pump no punips operating or only one pump operating operating # 9. Reactor Containment Vessel 1 4 psig 1 4 psig # '? m U ) Trip may be manually bypas. sed when RCS pressure $ 1820 psig t / actuating Shutdown Bypass provMed that: a. The fluclear Overpower Trip Se'tpoint is 1 5% of RATED TiiERMAL POWER O b. The Shutdown Bypass RCS Pressure - liigh Trip Setpoint of 1 1820 psig is imposed, and The Shutdown Bypass is rmoved when RCS Pressure >l900 psig. c. Q

  • Allowable valtb for Channel Functional Test
  • AAllouable value for Channel Calibration
  1. Allowable value for Channel Functional test and Channel Calibration O

e

2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products.to the reactor ccolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surfac.e temperature is slightly above the coolant saturation temperature. Operation above the upper oundary of the nacleate boiling regime would result in excessive cladding temperatures because of the onset of departu.0 fr.ca nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNS is not a directly measurable parameter during operation anc -herefore THERMAL POWER and Reactor Coolant T2=per-ature and Pressure have been related to CNB through the BAW-2 DN3 flux correlation. The DN3 correlatien has been developed to predict the DNB f' s flux and the locatica of DNB for axially uniform and non-uniform heat riux distributions. The local DNS heat flux ratio, CNBR, defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, is indicative of the margin to DNS. The minimum value of the CNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNS will not cccur and is chosen as an appropriate margin to DNB for all operating conditions. The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 377,000 gpm,whichis102*ofthe[ design flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear pcwer peaking fa,ctors.with potential fuel densification effects: N N 'l Fq = 2.67; Fig =1.78; FZ = 1. 50 The design limit power peaking factors are the most restrictive calcu-lated at full power for the raage from all control rods fully withdrawn to ninimum allcwable control rod withdrawai, and form the core DNER design basis. 7s THREE MILE ISLAND - UNIT 2 B 2-1 r r; <? r ed or 4

s. m SAFETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is abcut 30 psi less than core cutlet pressure, providing a more conservative margin to the safety limit. The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and ira-lude the effects of potential fuel densification: 1. The 1.30 DNER limit produced by a nuclesr power peaking factor of F = 2.67 or the ccabination of the radial peak, axia.1 peak a-d position of the axial peak that yields no less than a 1.30 CNi?.. 2. The combinati:.9 of radial and axial peak that causes central fuel melting at the hot spot. The limit is 21.0 kw/f t. Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the pcwer peaking. The specified ficw rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively. The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power ccmbinations shown in BASES Figure 2.1. The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22%, whichever condition is more restrictive. Using a local quality limit of 22% at the point of minimum DNBR as a basis for curve 3 of BASES Figure 2.1 is a conservative criterion even thcugh the quality at the exit is higher than the quality at the point of minimum DNBR. The CNER as calculated by the SAU-2 DNS correlation continually increases frca point of minimum DN3R, so that the exit DNER is always higher. Extrapolation of the correlation beyond its published quality range of 22% is justified on the basis of experimantal data. TH:.EE MILE ISLAND - UNIT 2 B 2-2 or Ud. ', t-

SAFETY LIMITS BASES For each curve of BASES Figure 2.1, a pressure-temperature point abose and to the left of the curve would rcsult in a DNER greater than 1.30 or a local quality at the point of minimum CNBR less than 22% for that particular reactor coolant pump situation. The 1.30 DNER curve for four pump operatica is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of the four pump curve will be above and to the left of the.other curves. 2.1. 3 REACTOR CCCLANT SYSTEM PRESSURE The restricti:n of this Safety Limit protects the integrity of the Reactor Coolant Sys Em frca overpressurization and thereby prevents the release of radienucli:es contained in the reactor coolant frca reaching the containment ata:r:here. /'~ The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boi'.er and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of desien pressure. The Reactor Coolant System piping, valves and fittings, ar; designed to ANSI B 31.7, 2/6B Edition. Reactor Coolant System valves are designed to ANSI 3 16.5-1963, MSSP-61 and MSSP-66. The maximum transient pressure for the Reactor Coolant System valves is pennitted by ASME to be 110%, l 2750 psig of design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3125 psig,125% of design pressure, to demonstrate integrity prior to initial operation. s /"'s THREE MILE ISLAND - UNIT 2 8 2-3 e,9 9 r -- L/ I -.a.

dBb w s s LIMITING SAFETY SYST04 SETTINGS EASES The AXIAL POWER IMEALANCE boundaries are established in order to prevent reactor thermal limits frcm being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The AXIAL POWER IMBALA"CE reduces the power level trip produced by the f ux-to-flow ratio .uch that the boundaries of Figure 2.2-1 are pro".uced. The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.05% for a 1% flow reduction. RCS Pressure - Low, High and Variab;e Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted. During a sl6w reac:i.ity insertien startup accident frca low po'.er or a sicw reactivity inser:icn from high power, the RCS Pressure-High setpoint is reached befcre the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure 2igh, 2355 psig, has been established to maintain the system pressure belcw :he safety limit, 2750 psig, for any design -s s transient. The RCS Pressure-High trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefere set lower than the set pressure for these valves, 2500 psig. The RCS l Pressure-High trip also backs up the Nuclear Overpower trip. The RCS Pressure-Low,1900 psig, and RCS Pressure-Variable Lcw, (13.00 T,,, F-5387) psig, Trip Setpoints have been established to main-

lin :na":/;3 ratio greater than or equal to 1.30 for those dasign accidents that result in a pressure reduction.

It also prevents reactor operation at pressures below the valid range of 0"5 correlation limits, protection against CNB. Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpointof (13.00 T'ou. F-5927) psig. ( Nuclear Overpower Based on Pumo Monitors In conjuction with the power / imbalance / flow trips the Nuclear Over-power Based On Pump Mcnitors trip prevents the minimum core CNBP, from decreasing belwo 1.30 by triping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the po.er level for the number of pumps in operation. O THREE MILE ISLA"3 - U"IT 2 B 2-6 ET/ I 1. e

,g poo l ph rmc oua tPE 0 c 0 at n 4 3 8, eni 1 R a 9 9 2 l g 0 0 8 eon 6 2 1 noi OCt L 1 f% a o0 r 1 e p 5 n O a fh ot sr S s e T et I ca M rs xe I op er L t mg g cun ) n api j 0 if e t ( 4 0 o Rt a 3 e nr 4, se 6 eae 9 5 0 as el p 0 0 8 g ea roO 6 2 2 n re h o i cr TC 1 t nc a in r i m, pmp 2 0 ae N rt 2 I s s G s p R 3 R rp m ER A omg u WE E M t un 0 P OW L cpi 0 PO 4 t P B B a t 3 0, A N et a 0 n L T D R nr 9 6 7 a AL ae 0 0 7 l MA rl p 6 2 3 o RM uoO o ER oo C HE FC 1 1 TH r T o a t a c r a er e ho R t ) i e 2 2 et ( u g m h gn i p t ni s g i i m p w rur e p d eR e t o pE g r a o e W e u R l l RO L s bEP s w e aW tF e o h COL o r l t i PA lf g P F l M T o pLR t t t t pA E r n, n n aMH e ae a a e RT t l r l l l tE e ou o o b oHD m ot o o a nTE a C a C C c T r r i tDA a re r r i ER P op o o p

m. A f T

t m t t p ce c c A LR c aT a a ) e e e 2 R R R ( d= p Ig1 :.- ig p c T r ,'CN 3.

TABLE 3.3-1 (Continued) TABLE NOTATION

  • With the control rod drive trip breakers in the closed position and the control rod drive system capable of rod withdrawal.
    • When Shutdown Bypass is actuated.
  1. The provisions of Specification 3.0.4 are not applicable.

-10

    1. High voltage to detector may be de-energized above 10 amps on bcth Intermediate Range channels.

(a) Trip may be manually bypassed when RCS pressure < 1820 psig by actuating Shutdown Bypass provided that: (1) The Nuclear Overpower Trip Setpoint is < 5% of RATED THERMAL POWER. (2) The 5 -d:wn Bypass RCS Pressure--High Trip Setpoint of < 1320 psig is imposed. ~ (3) The Shutdown Bypass is removed when RCS pressure.> 19CO psig. ,s (b) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3. ACTICN STATEMENTS ACTICM 1 With the nurber of channels CPERAGLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and/or open the control rod drive trip breakers. With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels STARTUP and/or PCUER OPERATION may proceed provided all of the following conditions are satisfied; a. The inoperable channel is placed in the tripped condition within one hour. ~ b. The Minimum Channels OPERABLE requireaent is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1, r THREE MILE ISLAND - UNIT 2 3/4 3-3 -...,,9 <3' [} g 'L - ~ -

9 TABLE 3.3-3 (Continued) TABLE NOTATICN Trip function may be bypassed in this MODE with RCS pressure below 1920 psig. Bypass shall be autcmatically removed when RCS pressure exceeds 1950 psig. 3 channels per Autcmatic Actuation Logic, Each R. B. Pressure High Channel trips cne Sa'ety Injection Channel and one R. B. Cooling & Isolation Channel. 3 channels per Autcmatic Actuation Logic, R. B. Spray Valves are actuated by R. B. Cooling and Isolation.

        • Trip functicn may be bypassed in this mode with steam generator pressure < 500 psig.

Rypass shall be removed when steam generatcr pressure > 5:: psig. The provisicns of Specification 3.0.4 are not applicable. ( / N. THREE MILE ISLA:'O - U::IT 2 3/4 3-13 u4_._,.,,,A

g _l 2400 RCS Pressure - High Trip P = 2355 Psig, T=619 F 2300 ~ m RCS Outlet I Teeperature-E High Trip E 2200 t ACCEPTABLE P = 2160 Psig C?EP.ATION O O 2 12 2!CO 3,* \\ Safety - g q. } Jg Limit J@ a. 3' 2* 8 JV y 2CCC h@ Ug // UNACCEPTABLE ,lJ // OPERATICM - c. RC3 Pressure - Lo.4 Trip

g P = 1900 Psig1

/ J l i I 540 560 5S0 600 620 640 Reacter Cutlet Temperature, F THI - U!!!T 2 REACTOR CORE S AFETY LlHli Figure 2.1-1 - : o r e-U4 ". wJ

THERMAL POUER,5 120 CNER LIMIT (D ( 33.5,112) "

(24.5.112)

-- 110 KF/FT LIMIT ACCEPTABLE (-49.2.100) 4 yp -- 110 (40.3.100) 0?IRAT10N (1) I-50 (-33.5,E4.E) n(24.5,84.5) -- 53 A:CE?iAELE (-43.2,73.1) (40.8,73.1) 3 g 4 pg;,7 -- 70 C?ERATION ( 33.S.57.4) (((} _'_ g3 c O(24.5.57.4) ACCEPTABLE -- 53 (-49.2.45.9) 2,3,& 4 PU"P (40.8,45.9) OPERAil0N 40 -- 30 -- 20 16-I f f f f f f f f I t f 60 -50 -40 -20 -29 -10 0 10 20 30 40 50 60 Arial P:ser Iccalance,, CURVE REACTOR C00LANT FLO# (GP") 1 377,000 250,400 182,500 s REACICS CORE SATETY,LIMii3 Figure 2.! 2 r r$- <? + )' i} ~

1 l-5 0F RATED THER:iAL POWER ( 23.7,1C5), (15.8.105) l -- 103 - ACCEPTABLE I I (2S.8,93.5) (-31.1,93.5) i i 4 PCMP __ gg iJPERAT10N l f E0 (78.1) l 1 I ACCEPTABLE 70 l (-37.1.SS.6) 1 (29.3,65.5) 3 & 4 PUttp i I CPERATl0N 60 g I, (50.9) ~~ 53 1 l (.37,1,39,4)[ ACCEPTA3tE __ 40 gg j l 2. 3. & 4 l l PUMP 30 l lCPERATICN I I 20 y g d C = is u 10 f l f E!

I

= e t i t i ! t i t 50 -43 -30 -20 -10 0 13 20 30 40 50 Axtal Pcaer incalsn e, TRIP SEIPCINT FOR NUCLE AR OVET:Por9 BASED GN RCS FL3# AN3 AXI AL PC'llER !!!, AL ASCE 3 FiL3re 2.2-! ,e-e+vy c- ( Y

.s ( 5 0F RATED THERMAL POTER ~ ~ ' (105.125) i I lACCEPTAELE " 100 l l 4 FU:AP l (93.54) . ( g 3,5y I :?ERATION 90 l l ,- 83 (78.225) l ACCEPTABLE l 70 1 2 L 4 PUMP (66.64) (EB'64) l OPERATION l -- 60 l l (51.025) , !CCE? TABLE - ,0 l 1l23E4 .. 4g l (33.44) \\d3 4',', 1 PU:AP l l OPERATION 0 l a g I l I 20 ^ co g o g . l. a ci -- 10 l S a 1 i, i, n n ( ~i e m m, im I, t i. i -50 -40 -30 -20 -10 0 10 20 20 40 50 Axial Power I.031ance. ALLOWAEL E V ALUE FCR NUCL E A-{ OVER?0iiEP. SASED ON RCS FLD'1 AND AXi AL PO4ER !!! CAL ANCE Fi ure 2.?-2 e - ofa n- ( .O

t e .1 2400 I 2200 l l2 ? E u 3 2000 e a b. o E 1800 0 7 6 1G00 520 600 620 640 660 Reactor Outlet Terperature, F C U R ',' E NH FC; ER PD":5 0;ERATING IT'r?E ': L t "jil, 1 377,000(IC05) !!25 FCURFUMPS(CSSR) 2 290,400 (74.45) 8'4.65 THREEPUMPS(DSSTs) 3 182,800 ('48.55) 57.'45 ONEFUH?PERLCC?(QUAL!TY) THi - UNIT 2 PRESSURE / TEMPER ATUR E LlHITS AT HA:4 l HUM ALLOWASLE POWER F OR M lH ilidH DNSR F i.g u r e 2.I r - a, + n Y~ a; .)}}

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