ML20069F693

Proposed Tech Specs Reducing Thermal Design Flow.Safety Evaluation Encl

ML20069F693
Person / Time
Site:	Point Beach
Issue date:	09/17/1982
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20069F691	List: ML20069F687 ML20069F693
References
TAC-48853, TAC-48938, NUDOCS 8209280204
Download: ML20069F693 (61)

Text

{{#Wiki_filter:_- o 2) Cold Shutdown ~ The reactor is in the cold shutdown condition when the reactor has a shutdown margin of at least 1% ik/k and reactor coolant 0 temperature is <200 F. 3) Refueling Shutdown .The reactor is in the refueling shutdown condition when the reactor C is suberitical by at least 10% ak/k and T is <140 F. A refueling avg shutdown refers to a shutdown to move fuel to and from the reactor core. 4) Shutdown Margin Shutdown margin is the instantaneous i=ount of reactivity by which the reactor core would be suberitical if all withdrawn control reds were tripped into the core but the highest worth withdrawn RCCA remains fully withdrawn. If the reactor is shut down from a power condition, the hot shutdown temperature should be assumed. In other cases, no change in temperature should be assumed. s ~ 7F h. Power Operation _ The reactor is in power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates greater than 2% of FULL power. 1. Refueling Cperation Refueling operation is any operation involving movement of core components L i (those that could affect the reactivity of the core) within ene contain-4x ment when the vessel head is unbolted or removed. b4a e N o. j. Rated Power No 03ino Rated power is here defined as a steady state reactor core output of Es 1518.5 MWT. mo[ k. nermal Power ox QQ GMLA Ther=al power is defined as the total core heat transf erred f r0m the fuel to the coolant. 15.1-4

e o 1. Degree of Redundancr Degrar. of redundanc-r is defined as the difference berveen the number of operable channels and the m4*4 m number of ch'annelm which when tripped vill cause an automatic shutdown. n. Reactor Critical Its reactor is said to be critical vben the neutron chain reaction is self-sustaining and ker., = 1.0. n. Lov Pever Oceration "'he reactor is in the low power operating condition when the reac*cr is cri*ical and the average neutron flux of the power range instrumentation indicates less than or ec,ud to 25 of FULL power. o. Fire Sueeression Water Svstem A 2*:3E SU PRESS *CN WA1'ZR SYS*l*IM shall censist of: a water source: pump (s) ; and distribution piping with associated seenicnalizing cent:01 or isolation valves. Such valves shall ir.clude yard pcs: indica *ing valves and the first valve shad of the water flev alarn device on each sprirJcler, hose sts;..dpipe or spray system riser. p. Full Power , Full power is defined as 1007. of rated power when the RCS flow is > 178000 gpm. '* hen RCS total flow is < 178000 gpm, full pcwer l 1s defined :o be 91* of rated power. i 15.1-5

SATTfY

• C E ~2 73IG SAFTTY SYSIZM S C IG5 1

15.2.0 15.2.; SAFZTY ~4'CT, vACTOR CORE Applicability: Applies :: the li:niting combinations of ther:nal power, reactor coolgt system pressure, and coolant temperature during operation. Cb3octive:" To :narntain the integrity of the fusi cladding. 5pecificacion: 1. The combination of ther=al power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 when RCS Total Flow rate > 178000 gpm, and Figure 15.2.1-2 when RCS Total Flow race < 178000 gpm. The safetf limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line. i t [ l l t l l

. s. :. :. -l.

Sas:fs : c naantain the integra 7 cf the fuel cladding and prevent fiss:.cn prcduct release, it is necessary to prevent overhea: Lng cf the cladding ut. :er all cptrating condit:cns. "'his is accomplisned by operating the het regions of the core within ne nucleato boiling reg =e of heat transf er, wnerein de heat transfer coefficient is very large and the clad surface temperature :.s only a how degrees pahrenheit above the ecolant saturation temperature. Se upper boundary of the nucleate boiling reg =e is termed depar.ure from nucleate boiling (CNB) and at this point there is a sharp reduc icn of the heat transfer cetfficient, which would result in high clad temperatures and the pcssibill:7 of clad failure. CNB is not, however, an observable parameter dur:.ng reacter cpcration. Serefore, the observable parameters; ther=al pcwer, reacter cociant temperature and pressure have been related to ONB drcugn de W-3 ONB correlation. Se W-3 CNB correlatica has been developed to predict the ONB flux and the location of CNB for axially unifcm and ncn-unifers heat flux distributsens. Se local CNB heat flux ratic, defined as the ratic of de heat flux that would cause CNB at a particular core location := the local heat flux, is indicative of the argin to CNB. Se sinimum value of the ONB rat:.c, CNBR, during steady state operation, normal operaticnal transients, and ant:.c:. pated transients is li:nited to 1.30. A CNB ratio of 1.30 corresponds to a 95% i precab:.lity at a 9ft confidence level that CNB will not occur and is encsen as l an appropriate margin ec ONB for all cperating conditions. (1) 3 e curves of Figure 15.2.1-1 and 15.2.1-2 represent the loci of points of thermal power, coolant system pressure and average temperature for which the 2N3 ratio is not less than 1.30. The area of safe operation is below these lines, he saf ety limits curves have been revised to allow f or heat flux peaking effects due to fuel densification and flattened fuel cladding sections. I 1 i __$....gy_ v - - -, _.

---

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Ad6itional p2sking fcctors to acccunt for local petking dua to fuel rod exial gaps and reduction in fuel pellet stack length have been included in the cal-culation of the curves shown in Figures 15.2.1-1 and 15.2.1-2. These curves y arebasedonanThof1.58,cosineaxialfluxshape,andaDNBanalysisas described in Section o.3 of WCAP-8050, " Fuel Densification, Point Beach Nuclear Plant Unit 1 Cycle 2d (including the effects of fuel densification and flattened cladding). Figures 15.2.1-1 and 15.2.1-2 also include an allowance for an increase in the enthalpy rise hot channel factor at reduced power based on the expression: N F = 1.58 {1 + 0.2 (1-P)} where P is a fraction of FULL power g when P < 1.0 Fg = 1.58 when P > 1.0. The effects of roc bow have been included in the determination of a conserv-ative value for F Rod bow effects of up to 14.9% DNBR are offset by g. credits available from the design limit DNBR, pitch reduction, design thermal diffusion coefficient and the fuel densification power spike, which were pre-viously approved.* 'lhe hot channel factors are also'sufficiently large to account for the degree of malpositioning of f'ill-length rods that is allowed before the reactor trip set-points are reduced and rod withdrawal block and load runback may be required. Rod withdrawal block and load runback occur before reactor trip setpoints are reached. The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30. Memorandum from D. F. hose, and D. G. Eisenhut, USNRC, to D. B. Vassallo and K. R. Golle r, " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors," dated February 16, 1977. 15.2.1-3

/ 680 ~ ~ 660 9 2400 PSIA

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640 2000 P3tA 620 1700 PSIA 600 e E 580 I LOCUS OF PolNT3 560 7 F0s unicx sitAu GEE ER A TOR S A FE TY V ALVE3 ARE ACTUATED 540 520 0 20 40 60 80 100 120 POWER (PERCE. T OF FULL) 4 TCTAL RCS FLCW >_ 173000 G.M Figure if.2.l-l Core DNS Sefary Limih Po1nt Eeacn Units i and 2 9

690 \\ 670 - - 2400 PSIA 650 - - 2200 PSIA 630 - 2000 PSIA u o w

n D

i 4 610 - 1775 PSIA ~ 590 - - 1 570 - - I 550 0 20 40 60 80 100 120 POWER (PERCE:IT OF FULL) TOTAL RCS R.0W <178000 GFM FIGURE 15.2.1-2 CORE 0:18 SAFETY LIMIT 5 i 00I'IT Cet~z.. "'I.t.1 u

'5.2.3 'l'C'2!G SAT 75 SYS'"ZM w_.=tas, pr,cn;; vg n;s am s-- ;=; Acclicabilit*/: Applies to nip settings for instruments :senitoring reac== power.and reac=r cocist.t pressure temperature, flew, pressuri=er level, and permissives related to reac== protaction. Chiective: 2 provide fer automatic pretac ive action in the event that the principal process variables approach a sadacy id dt. Seee tfiestien: 1. Protective instrumentation for reactor trip settings shall be as follows: A. Startup protection (1) High flux, source range - within span of source range instrumentation. (2) High flux, intermediate range - 140% of FULL power. l (3) High flux, power range (low set point) - 125 % of FULL power. B. Core lim 4: protection (1) High flux, power range (high setpoint) < 108% of FULL power l (2) High pressurizar pressure - 1 0385 psig. l 15.2.3-1

(3)

• cw p cccurizer proccura - > 1790 psig for Operatten at 20C0 psia pr r.ary systan pressure (4)

Over emperature AT <a To (K1-K2 (I'T ' ) (l*T'SI -K (P-P ') f(a!)) 3 let:5 where ATo = indicated AT at FULL power, 'F average temperature, 'T T = 574.2 *? T' = pressurizar pressure, psig ? = 2235 psig P' = t 1.30 for operation ac 2000 psia primary system pressure K 0.0150 K, = K3 0.000791 = 25 sec. T1 = 3 sec. t2 = and f(AI). is an even func icn of the indicated difference between top and bottem detectors of the power-range nuclear ion chambers; with gains to be selec.ed based en ::teasured inst nent response during plant star:np tests, where qt and qb are the percent power in the top and bottcm halves of the core respectively, and qe + qb is total core pcwer in percent of FULL power, such that: (a) for qe - qb wit.W -17, +9 percent, f (a!) = 0. (b) for each percent that the :sagnit.:de of qtm.d aaceM s -9 percent the iT trip set point shall be autrsatically reduced by an equivalent of two jercent of FULL power. 15.2.2

(c) for each percent that the magn-tude of q. - qb exceeds -17 percent the AT trip serpoint shall be autes::atically reduced i by an equivalent of two percent of Fi1LL power. (1.3. (5) } overpowe-AT 1 b. (K4-K5 T2s T-K6 (T-T ' ) -f (A:33 m,. where indicated AT at FULL power, *F AT. = T = average tamperature,

• T T'

574.2 =- K4 1 1.089 of FULL power K5 0.0262 for increasing T = 0.0 for decreasing T = K6 0.00123 for T > T' = 0.0 for T < T' = T3 10 sec. = f (AI) as defined in (4) above, I (6) f1ndervoltage - > 75% of normal voltage l ~ l (7) ww indicated reactor cociant flow per icop-i l I l >90% of normal indicated loop ficw (8) Reactor coolant pump : noter breaker open 1 1 (a) uw frequency set point >57.5 cps (b) ww voltage set point >75% of nor=al voltage l l l l l ( 't a 1.1 ( -- ~

2. Protective instr.=e.ntation settings f== reac cx :-1; inter-iceks shall be as fell =ws: A'. S.e "at power" reacecr trips (1cw pressuriser pressure, h.igh pressurizar level, and icw rsac.cr coolant flew for both loops) shall be unblocked when (1) Power range nuclear flux.> 9% (*11) of ELL power or (2) Mrbine Load > 10% of full lead turbine pressure. 3. S e single loss of ficw ::1p shall be u=hlecked when the power range nuclear flux > Sch of WLL power. C. "'ha power range high flux level icw range ::1p, and 1staz=ediata range hign flux level trQ shall be unbiccked when power is 1 9% (+1%) of FULL power. D. S.a source range high flux reactor trip shall be unbiccked when the intermedi. ate rs=ge flux is 1 10 ~O amperes. i 15.2.3-4

4 power distra;u :.cn, the reac:c: crap l'- , w:.th allowance fer errers, (2) is always below he cere saf s y li=it as shewn en Fig =e 15.2.1-1. If axial peaks are greater than design, as indicated by diff arance be: ween .r:p id d: sep and ro om power range nuclear detec crs, the reac_c is succmatically reduced. (6) (7) overtemperat=e and pressurizer pressure system se poin_s me everpowe, nave been revised to include ef"ect of reduced systan pressure opera:Len (including the eff ects of fuel densification). he rev:. sed se poin.s as given above will not exceed the revised ccre safety limits as snown in Figure 15.2.1-1 and 15.2.1-2. De overpower limit critaria is that core power be prevented from reacning a value at which fuel pellet cantarline melting would oce=. Se reae:c condition by action of the

.s pervented from reaching the overpower limit nuclear everpower ard overpower ST =ips.

D e high and lew pressure reactor =ips limit the pressure range in wnich reactor operation is pectu. :ed. De high pressur:.cer pressure reac cr trip set _ing is icwer than the set pressure for the saf ety valves ( 2485 the reac.or is = pped before the safety valves actuate. psag) such that n e icw pressuriser pressure reae:m: =: p _r ps the reacec: :.n the unl:kely event of a less-of-coolant see: dent. (4)

• NB :. the even cf The icw flow reac.=r =1p protects the cere agains:

in the loops er a sudden loss either a decreasing actual =easured flew of power Oc one or both reactor coolant pu=cs. De se: point spec:.f:.ed analysis. F*) Se icw is consistent with the value used :.n the accident flew as measurec signal is caused by a cond ::en of less enan 9C% loop ficw Se icss Of power signal :.s causec cy by the ;ccp ficw :.ns tr'.:.=ent a t:.cn. 15.2.2-6

the reactor coolant pump breaker opening as actuated by either high current, low supply voltage or low electrical frequency, or by a manual control switch. The significant feature of the breaker trip is the frequency setpoint, 57.5 cps, which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water level reactor trip protects the pressurizer safety valves agains water relief. The specified setpoint allows adequate oper-ating instrument error ( ) and transient overshoot in level before the reactor trips. The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system. (9) Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed setpoint above which these trips are unblocked assures their avail-ability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have 11% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. Sustained operation with only one pump will not be permitted above 10% Full power. If a pump is lost while operating between 10% and 50% of Full power, an orderly and immediate reduction in power level to below 10% of Full power is allowed. The power-to-floe ratio will be maintained equal to or less than unity, wh*ch ensures that the minimum DNB ratio increases at lower flow because the maximum enthalpy rise does not increase above the maximum enthalpy rise which occurs during full power and full flow operation. References (1) FSAR 14.1.1 (4) FSAR 14.3.1 (7) FSAR 3.2.1 (2) FSAR, Page 14-3 (5) FSAR 14.1.2 (8) FSAR 14.1.9 (3) FSAR 14.2.6 (5) FSAR 7.2, 7.3 (9) FSAR 14.1.11 15.2.3-7

l 15.2

• :MITING 00; 27:O:5 ?CR CPE?ATION if.2.1 REAC"'OR COCLANT s'lS"T.M Acclica.cilita?

Appl?es to the operating status of the Reacter Ocolant System. ch ec ave ce specify those limiting conditions f=r operation of the Reac cr Ccclant System which :nust be : net to ensure saf e reac.cr operation. Seecification A. OPERATICNAL COMPCNENTS Specification: 1. Coolant Pumps a. At least one react.or coolant pu:np or the residual heat renoval system shall be in operation when a reducticn is :nade in the bcron concentratien of the reac cr coolant, t b. When the reacter is critical and above 1) Full

power, except for natural circulation tasts, at least ene reactor coolant pump shall be in operat.cn.

c. (1) Reactor power shall not be :naintained ateve los of FULL power unless both reactor coolant pumps are in operaticn. (2) If either reactor ccolant pe=p ceaser Operatinc, immediata power reduct:cn snall r,e in:.tt.ated under adal.nistrat:ve centrol as necessar/ to recuce pcwer to less than 10% of FULL Power-2. Steam Generator Cne steam generator shall be creracle wnenever ne avera9e a. reac_cr coolant temcerature :.3 acove 350*F-15.3.1-1

btenusa of tho low praccuri cr volume and baccusa prsecurtsar boron concentration normally will be higher than that of the rest of the reactor coolant. Part 1 of t.ke specification requires that a sufficient number of reactor coolant pumps be operating to provide core cooling in the event that a loss of flow occurs. The flow provided in each case will keep DNBR well above 1.30 as discussed in FFDSAR Section 14.1.9. Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Heat transfer analyses (1) show that reactor heat equivalent to 10% of FULL power can ce removed with natural circulation only; hence, the specified upper 14-4 t of 1% FULL power without operating pumps provide a substantial safety factor. Each of the pressurizer safety valves is designed to relieve 288,000 lbs. per hr. of saturated steam at setpoint. Below 350*F and 350 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat is removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' l capacity. One valve therefore provides adequate defense against over-pressurization. Part 1 c(2) permits an orderly reduction in power if a reactor coolant pu=p is lost during operation between 10% and 50% of FULL cower. l Above 50% FULL power, an automatic reactor trip will occur if either pumc is lost. The power-to-flow ratio well be maintained equal to or less than 1.0 which ensures that the minium DNB ratio increases at lower flow since the maximum enthalpy rise does not increase above its normal full-flow maximum value. (2) A PORV is defined as OPERABLE if leakage past the valve is less than that l allowed in Specification 15.3.1.D and the PCKV has met its most recent channel test as specified in Table 15.4.1-1. The PORVs operate to relieve, in a controlled =anner, reactor coolant system pressure increases below 1 15.3.1-3

G. OpERAT:CNAL' L 1!ITAT GNS The following ON3 related parameters shall be maintained within the limits shown during FULL power operation: 1. T VO shall be =aintained: A j; 573 F when RCS total flow 3,178000 gpm j;376.9*F when RCS total flow < 178000 gym ~ 2. Reactor coolant system pressure shall be maintained: 1,1955 psig during operation at 2000 psia 3. Reactor Coolant System Total Flow Rate 3 (.95) x 178,000 i Basis: Although the operational limitat/.ons above require reactor coolant syr...sm total flow be maintained above a minimum rate, no direct means of measuring absolute flow during operation exist. However, during initial startup reactor coolant flow was measured and correlated to core ST. Therefore 4 monitoring of aT may be used to verify the above minimum flow requirement is mat. If a change in steady state full power AT greater than 3 F is observed, the actual flow measurements will be taken. l l I i l l 15.3.1-19 l ---_m

The eight main steam safety valves have a total ccumbined rated capability of f 6,664,000 lbs/hr. The total. rated steam flow is 6,620,000 lbs/hr, therefore eight (8) main steam safety valves will be able to relieve the total full steam flow if necessary. In'the -unlikely event of complete loss of electrical power to the station, decay heat removal would continue to be assured for each cnit by the availability of either the steam-driven auxiliary feedwater pump or one of the two motor-driven auxiliary steam generator feedwater pumps, and steam discharge to the atmosphere via the main steam safety valves or atmospheric relief valves. One motor-driven auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from a unit. The minimum amount of water in the condensate storage tanks is the amount needed for 25 minutes of operation / unit, which allows sufficient time for operator action. i An unlimited supply is available from the lake via either leg of the plant service water system for an indefinite time period. ( 15.3.4-2a

e 15.3.5

STam: TATION SYSTEM Opera 10nal Safety Instrumentation Applic ability :

Applies to plant instrumentation systems. Ob3ectives: To provide for automatic initiation of the Engineered safety Features in the even5 that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety. Specification: A. The Engineered Safety Features initiation instrumentation setting limits shall be as stated in Table 15.3.5-1. B. For on-line testing or in the event of a sub-system instrumentation enannel failure, plant operation at FULL power shall be permitted. to continue in accordsnee with Tables 15.3.5-2 through 15.3.5-4. C. In tha event the number of channels of a particular sub-system in service falls below the limits given in the colu=n entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according ta the requirement t shown in Tables 15.3.5-2 through 15.3.5-4, operator Action when i minimum operable channels unavailable. I D. The accident monitoring instrumentation channels in Table 15.3.5-5 ( shall be operable. In the event the nu=ber of channels in a part - cular sub-system falls below the mintmum number of operable channels given in Column 2, operation and subsequent operater ac rion shall be in accordance with Column 3. Bas:s: Instrumentation has been provided to sense accident ccnditlens and to initiate operation of the Engineered Safety Features (l). 15.3.5-1

TADI.E 15.3.5-2 (Cont'd) L 1 2 3 4 5 NO. OF MIN. MINIMUM PERMISSIBI.E OPEHATOR ACTION NO.OF CilANNELS OPERABLE DECHEE OF BYPASS IF COND{TIONS OF No. FUtiLTIUt3AL UllIT CllANNEIS TO CIIANNE!S REDUNDANCY CONDITIONS COIAJHN 3 OR 4 THIP CANNOT,BE HLT 11. Turbine '" rip 3 2 2 1 Maintain <50s of FULL gewer 12. Steam Flow - Feed Water Flow 2/ loop 1/ loop 1/ loop 1/ loop Maintain hot inismatch shutdown 13. Lo to : steam Generator 3/ loop 2/ loop 2/ loop 1/ loop Maintain liot i Liater Level shutdown 14. Undervoltage 4 KV Bus 2/ bus 1/ bus 1/ bus Maintain liot (both buses) shutdown ' 15. Underfrequency 4 KV Bus 2/ bus 1/ bus 1/ bus Haisitain hot (both buses) shutdown i NOTE: When block condition exists, maintain normal operation. FP. FULL power tlut Applicable Onu additional chasinel may be taken out of service for zero power physics testing. Page 2 of 2

TABLE lh 3.5-5 INSTRUMENT OPERATING CONDITIONS FOR INDICATIONS 1 2 3 MINIMUM NO. OF OPERABLE OPERA'IOR ACTION IF CONDITIONS NO. FUNCTIONAL UNIT CHANNELS OIANNEL OF COLUMN 2 CAN'NOT BE MET 1. PORV Position Indicator 1/ Valve 1/ Valve If the operability of the PORV 1 tition indicator cannot be restored within 48 houta, shut the associated PORV Block Valve. 4 2. PORV Block Valve Position 1/ Valve 1/ Valve If the operability of the PORV Block valve Position Indicator Indicator cannot be restored within 48 hours, shut and verify the Block Valve shut by direct ~ observation or declare the Block Valve inoperable. 3. Safety Valve Position Indicator 1/ Valve 1/ Valve If the operability of the Safety Valve Position Indicator cannot be restored within seven days, be in at least Hot Shutdown within the next 12 hours. 4. Reactor Coolant System Subcooling 1 1 If the operability of a subcooling monitor cannot be restored or a backup monitor made functional within 48 hours, be in at least Itot Shutdown within the next 12 hours. S'. Auxiliary Feedwater Flow Rate

• 1 1

If the operability of the auxiliary feedwater flow rate indicator cannot be restored within 48 hours, be in hot shutdown within 12 hours. 6. Control Rod Misalignment as Monitored 1 1 Log individual rod, positions once/hr., af ter a by On-Line Computer load change >10% of full power or after >30 inches of control motion.

• Applies to presently installed combination of auxiliary feedwater pump discharge flow indicators and auxiliary feedwater flow to steam generator indicators.

A.2 Under abnormal conditions including Black Plant startup, one reactor may be made critical providing the following conditions are met: a. One 3'45 K7 transmission line is in service; or the gas turbine is operating. b. The'345/13.8 KV and the 13.3/4.16 KV station auxiliary transformers associated with the unit to be taken critical are in service; or the associated 13.8/4.16 K7 station auxiliary transformer is in service and the gas turbine is operating. c. Reactor power level is linlted to 50 FULL power until 2 or more transmission lines are reatored to service. d. 480 Volt buses 303 and BO4 for the unit to be taken critical are energized. 4160 Volt buses A03, A04, A05, and A06 for the unit to be taken ~ e. critical are energized. f. A fuel supply of 11,000 gallons is availabla; and both diesel gene-stors are operable, Both batteries and DC systems are operable. g. B.1 During power operation of one or both reactors, the requirements of 15.3.7.A.1 may be modified to allow the following arrangements c ' systems i and components: If the 345 KV lines are reduced to only one, any operating reactor (s) a. must be promptly reduced to, and limited to, 50: FULL to-er. If att 345 KV lines are lost, any operating reactor (s) will be reduced to supplying its auxiliary load, until one or more 345 K" transmission lines are again available. 15.3.7-2

b. If both 343/13.8 KV cuxiliary transformtra cro cut of sarrico and only the gas turbine is operating, only one reactor will remain operating and it will be limited to 50% FULL power. The se:Ond reac Cr will be placed in the hot shutdown condition. c. If the 13.8/4.16 KV auxiliary transformers are reduced to,only one, the reactor associated with the out of service transformer ~ zust be placed in the hot shutdown condition. d. Either bus A03 or A04 may be out of service for a period not exceed-ing 7 days provided both diesel generators are operable and the associated diesel generator is operating and providing power to the engineered safeguard bus normally supplied by the out of service bus. e. One diesel generator may be inoperable for a period not exceeding 7 days provided the other diesel generator is tested daily to ensure operability and the engineered safsty features associated with this diesel generator shall be operable. ) f. One battery may be inoperable for a period not exceeding 24 hours 7 provided the other battery and two battery chargers remain operable with one charger carrying the DC load of the inoperable battery's DC supply system. l Basis This two unit plant has four 345 KV transmission line interconnections. A 20 MW gas turbine generator and two 2850 KW diesel generators are installed at the' plant. All of these energy sources will be utilized to provide i depth and reliability of service to the Engineered safeguards equipment i through redundent station auxiliary power supply systems. e 15.3.7-3

If only one 345KV transmission line is in service to the clant switchyaro, a temporary less of this line would result in a reactor trip (s) if the reactor (s) power level were greater than 50% FULL power. Therefore, in order to maintain continuity of service and the possibility of self-sustaining ooerstions, if only one 345KV transmission line is in service to any operation reactor (s), the power level of the affecteo reactor (s) willbelimited}o50%FULLcower. f If both 345/13.8KV station auxiliary transformers are out of service, only one reactor will be operated. The gas turbine will be supplying power to operate the safeguards auxiliaries of the operating reactor and acts as a backup supply for the unit's normal auxiliarids. Therefore, to prevent overloading the gas turbine in the event of a reactor trip, the maximum power level for the operating reactor will be limited to 50% FULL power. These conservative limits are set l to improve transmission system reliability only and are not dictated by safety system recuirements. References FSAR Section 8 15.3.7-5

15.3.10 CONTROL RCD AND POWER DISTRIBUT CN LIMI 3 Acclicability Applies to the operation of the control rods and to core power distribut:.cn limits. Ob3ectEve To insure (1) core subcriticality after a reactor trip, (2) a limit on potential (RCOA) reactivity insertions from a hypothetical rod cluster control assembly an acceptable core power distribution during power operation. ejection, and (3) Soecification A. Bank Insertion Limits 1. When the reactor is critical, except for physics tests and control rod exercises, the shutdown banks shall be fully withdrawn. When the reactor is critical, the control banks shall be inserted no 2. o, 15.3.10-1. further than the limits shown, by the lines on Figure Exceptions to the insertion limit are permitted for physics tests and control rod exercises. The shutdewn margin shall exceed the applicable value as shown in 3. under all steady-state operating conditions from l l Figure 15.3.10-2 of An exception to the stuck RCCA cc=ponent 350*F to FULL power. the shutdown margin requirement is permitted for physics tests. least a ak/k shall for physics tests a shutdown margin of at 4. Except be maintained when the reactor ecolant temperature is less : nan 350*F. When the reactor is in the hot shutdown condit:en or during any 5. i for physics tests, the critical f approach to criticality, except for cerc rod position shall not be icwer than the insertion limit i t i l if the control rods were withdrawn :.n nc:-.a1 power. That is, t the reactor would nct sequence with no other reactivity change, insertien 1:. 10. 1 be critical unt:.1 the control banks were above the i 15.3.10-1 l

3. , ?cwir Oistr.hu :.:. ;. :s 1. a. Execy: hr:.ng ;cw pcwer phys:.cs :ssus, tha hc: channel facters defined :.. ene basis =us: =ee: :ne f=11:w:ng l'-':s: ()<(,..a ) x.< s. ) .c-P>5

n 4

= R. C S '..' ~. a.l p -,CWr3:8 1 F2 (2)<4.64 x K(2) for ? <,3 _ 178000 ~ F (Z) 1 (2.52) : E(Z) for ? >.5 9 RCS Total P Flowrate Q (Z) 1 5.04 x K(Z) for P <.5 < 178000 F5)a.l.Sa x {1 + 0.2 (1-?)} there P is the fraction of FULL pcwer at vnich e.e : Ore is l cperating, K(2) is the functicn in Figure 15.3.10-3 af.d 2 is the core height locacicn of F Q b. Following a refueling shutdown prior to exceeding 90% of FULL power and at effective FULL power ::cnthiy intervals thereafter, power distribution =aps using the moveable incere detec::: system shall be made to confirm that e.e hot channel fac :: l'- :s are satisfied. "'he seasured hot channel fac crs shall be increased in the fc11cwing way: The measurement of Octal peaking facter, F(**8, shall be (1) increased by three percent Oc acecunt f: =anufactur ng eclerances and further increased by five percent to account i for =easuremen: error. (2) The measure =ent of enthalpy rise hot channel facter, F'!, shall be increased by four percent c ac cun for neasure-3ent error. .cwer i=:.: l c. If a =easured hot channel facter exceeds the FULL cf Specificatiicn 15.3.10.3.1.a, the reacter pcuer and pcwer range hagn setpcines snall be reduced until these 1* - s are met. If su sequent flux = app:.ng cannce, w:. thin 2 4 hcurs, dernnstra e :na:

ne full.cwer act :nannel factor ' -- s are met, :ne Overpewer

and ever e=perature 10 trip se:pcines shall be s:.reLarly reduced and reac:=r pcwer l'-- ed suen that specifica icn 15. 2.1:;.3.1. 2 acove :.s mez. 2. a. O.e target flux difference as defined in the basis shall be measured at least quarterly. A target flux difference update s value shall be deter =ined monthly by =easurement, or by 1 near \\ interpolation between the last =sasured value and M at end of cycle life (that is when the boren concentraticn in the ceclant.s zero ppm), or by extrapclatien of the last t ree measured. points. The target flux difference and its associated alarm sespcints need not be updated if the update value for FULL power target flux difference is within 10.5% cf the presently empicyed FULL power arget flux difference value. b. F.xcept for physics testing, execre detector calibratzen (including / recovery), or as modified balcw, the indicated ax:.a1 flux difference shall be maintained wican a range of +6 and -9 percent of the target flux difference. Th.s is defined as :he carge band. c. At a power level greater than 90 percent of FULL power, if :he indicated axial flux difference deviates frem :. s arget band, the flux difference shall be returned to the targe band immediately or reacter pcwer shall be reduced := a level nc greater than 90 percent of FULL power. 1 d. At a power level ne greater than 90 percent of Flitt pcuer, 6 (1) The indicated ax:.a1 flux difference.ay dev:. ate frc=

s -s to -9% targe: cand fer a max:.=um of cne hour ( cu=ula t:.ve s in any 14 hcur per.cd prev:.ded the flux d.Iference does act exceed an envelcpe counded ry -11 percen and

'1,:er:ent 90% FULL cower anc increasing by -1% anc +1% #:r eac.- 2% :f at 15.3.10-3

1 p ower se _$ cw. 0.,.,. If the cumulative '1:e er.ceeds :ne i g m.,.

vuu hcur :.n e.y 24 heur paried, :nen the reac-cr ;cwer shall-de FULL power and the l

reduced i= mediately o no greater than Sog high.eu rcn flux setpoint reduced to no greater e.an 55% j of FULL power. l A power increase ec a level greater can 90s of FULL power (2) is ccnc.ngent upcn the indicated axial flux difference being within its target band. I At a ;cwer level no greater than 50 percent of FULL power, c. The indicated axial flux difference =ay deviate frcs ::s (1) target band. 1 A ;cwerincrease to a level greater than 50% of FULL power (2) is centingent upon the indicated axial flux difference net (cu=ulative) being cutside its carget band for more than two hours Cne half of the _ =e of the preceding 24 hour pericd. out the indicated axial flux dif ference is cut of i_s target of FULL power is to be counted as con:ributing band up to 50% flux diff erence may to the one hour e.:mulat:ye maximum e.e deviate f rcm its target band at a ;cwer level less ::an j or equal to 90't of FULL power. l w:.:n l Alar =s shall nor= ally be used to indicate ncn-cenfor=ance f. 15.3.10.3.2.0 or the flux flux difference requirement Of l the

15. 3.10.3. 2. d (1).

If :he , difference-ci=e requirement of the a:c.a1 flux dif ference alar:.s are temporarily cut-of-ser-n.ce, w:.c. -he 11:1:s assessed every hcur shall be noted r.d ccnfer=ance s and half-neurly.nereafter. ( fcr :ne firs: 24 hours, 15.3.10-4 s

1. Excep: ::: ;r.yst.cs :ss:s, s..cnever : e.n:..ca:ac quar. a.: :0-er

.1: cxcocds 0% tho.11: conditien shall bo ol'---cred within twc hcurs er the felicwing acticns shal.1 he taken:

Reduce ccre p:wer level and the power range hacn flux se poin: a. wo percent of rated values fer ever/ percent of in'.icated quadrant pcwer.ilt. b. If.he ilt is net corrected within 24 hours, but the hot channei fac Ors for rated pcwer are not exceeded, an evaluanaen as := the cause of the discrepancy shall be made and reper.ed := the Nuclear Regulate:/ Consnission. Return to FULL power is permitted, previding l the het channel factors are not exceeded. If the design hot channel facters for FULL power are exceeded or [ c. determined within 24 hours, the Nuclear Regulatorf Oc._ ssicn not shall be notified and the overpower AT and evertemperature l' trip se:- points shall be reduced by the equivalent of M FULL power for every l percent of quadran pcwer tilt. d. "'he execre nuclear instrumentation system serves as the primary quadrant power tilt alarm. If the alarm is not functional for :ve hours, backup meuhods of assuring that the quadrant power :il: is acceptable shall be used. These methods include hand calculations, incore :hormoccupies using either a ces:puter or =anual calculatsens or incere detectors. Aen ene power range channel is incperable and therr.al power is e, greater than 75% of FULL power, the quadrant power til: shall be confirmed as acceptanle by use of the movanle :.nco re detec crs at least once per 12 hours. C. Inceerable Red Cluster Centrol Asser:civ CTC.N 1. An RCCA shall be considered :.ncperanle if cne er more cf.he fclicw:.ng occurs: ..c.. s. _ n _ - e W -

D. Misaligned or Dropped RCCA 1. If the rod position indicator channel is functional and the associated RCCA is more than 7.5 inches indicated out of alignment with its bank and cannot be aligned when the bank is between 215 steps and 30 steps, then unless the hot channel factors are sht to be within design limits as specified in Section 15.3.10.B-1 within eight (8) hours, power shall be reduced to less than 75% of FULL power. When the bank position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches. 2. To increase power above 75% Full power with an RCCA more than 7.5 inches indicated out of alignment with its bank when the bank position is between 215 steps and 30 steps, an analysis shall first be made to deternine the hot channel factors and the resulting allowable power level based on Section 15.3.10.B. When the bank position is greater chan or equal to 215 steps, or, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches. 3. If it is determined that the apparent misalignement or dropped RCCA indi-cation was caused by rod position indicator channel failure, sustained power operation may be continued if the following conditions are met: For operation between 10% power and FULL power, the position of the a. RCCA(s) with the failed rod position indicator channel (s) wili be checked indirectly by core instrumentation (excore detectors, and/or thermocouples, and/or moveable incore detectors) every shift and after associated bank motion exceeding 24 steps in one direction. b. For operation below 10% of FULL Power, no special monitoring is required. E. RCCA Drop Times 1. At operating temperature and full flow, the drop time of each RCCA shall be no greater than 1.8 seconds from the loss of stationary gripper coil voltage to dashpot entry. 15.3.10-7

o anomalies which would, otherwise, affect these bases. Axial Fewer Distributien The procedures for axial power distribution centrol are designed to mini =ize the effects of xenon redistribution on the axial pcwer distribution durin,g load follow maneuvers. msically, control of flux difference is required to 11= : the difference between the current value of flux difference ( AI) and a reference value which corresponds to the FULL power equilibrium value of axial offset l hxial offset = AI/ fractional power). The FULL power target flux difference is defined as that indicated flux [ difference of the core in the following condition: equilibrium xenon C.ittle er no ose:11atiorJ and with the full-length rod centrol rod bank more than 190 steps withdrawn (i.e., the nor=al full pcwer positior). Values for all other core power, levels are obtained by multiplying the FULL power value by the factional power. l At'zero power the target flux difference is 0%. S ince the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of +6 and -9 percent AI are permitted f cm the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for =easuring l the target flux difference every month. For this reasen, the specification provides I three =ethods for updating the target flux difference. I S trict centrol of the flux difference hnd red position) is not as necessary [ during reduced power operation. "'his is because xenen distribution centrol at ( l

educed power is not as significant as the centrol at FULL power and allowance l

l l has been made in predicting the heat flux peaking facters for less strict centrol at reduced power. S trict control of the flux difference is not possible during certain physics tests or during required periodic exc=re cal;.brations There fo re, the specifi-which require larger flux differences than per= tted. 15.3.10-12

catiens en power distribution control are not applied dur:.=g phystes tests or excore calibrations. This is acceptable due to the increased core =enttoring performed'as part of the tests and low probability of a significant accident occurring during these operations. In some,tnstances of rapid plant power reduction, automatic rod notion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distributien sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band. However, to stmplify the specification for operation up to 90% of FULL power, a lir.itatton of one hour in any period of 24 hours is placed encperation outside the band. This insures that the resulting xenon distributions are not significantly different from these resulting from operation within the target band. For normal operation and anticipated transients, the core is p:ctected from ~ J overpower and minimum CNBR of 1.30 by an automatic protection system. Ccmpliance with operating procedures is assumed as a pre-condition; hewever, operator error and equipment malfunctions an separately assumed to lead to the cause of the transients considered. Quadrant Tilt The excere detectors are somewhat insensitive to disturbances near the core center such as ::u.saligned inner control rods. It is therefore posstble that a five percent tilt might actually be present in the core when the execre detectors zespend with a two percent indicated quadrant tilt. On the other hand, they are overly responsive to disturoances near the periphery. 15.3.10-13

in a dcliberato manner witacut uncus prosauro en :no cporacing perrennel bscause of the unusual techniques to be used to acc=: odace the reactivity changes associated with the shutdown. Misaliened RCCAS The various control rod barjes (shutdown banks and control banks, A, B, C, and D) are each to be moved as a banks that is, with all reds in the bank,yithin ene step (5/8 inch) of the bank position. Direct informacien on red position indication is provided by two methods: A digital count of actuating pulses which shows the demand position of the banks and a linear position indicator (LVDT) which indi-cates the actual red position. The rod position indicator channel has a de=cn-strated accuracy of 5% of span (+7.2 inches). Therefore, an analysis has been performed to show that a misalignment of 15 inches cannot cause design het channel factors to be exceeded. A single fully misaligned RCCA, that is, an RCCA 12 feet out of align =ent with its bank, does not result in exceeding core limits in steady-state operation at power levels less than or equal to rated pcwer. In other words, a single dropped RCCA is allowable from a core power distribution viewpoint. .If the misalignment condition cannot be readily corrected, the specified reduction in power tc 75% of FULL power will insure that design margins to core Linits will be maintained under both steady-state and anticipated tran-sient conditions. The eight (8) hour permissible limit on roa misaligncent at rated oower is short with respect to the probability of an independent accident. Because the red position indicator system =ay have a 7.5 inch error when a misalign=ent of 15 inches is occurring, the Specification allows only a 7.5 inch indicated =isalignment. However, when the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the censequences of a =1salign=ent are =uch less severe. The differential worth of an individual RCCA is less, and the resultant purturbatien en pcwer distributions is less ~ than when the bank is in its h1,gh differential worth regien. At the top and bottc= of the core, an indicated 15 inch =1salign=ent =sy be representing an actual =1salig-.=ent of 22.5 :.nches. The failure of an LVOT in itself does not reduce the shutdown capability of the 15.3.10-15 r - = v

'r:ds, but it does reduce 2.s operator's capab:lity for dece==ining the pesati:n ~

f c.a: : d by di' rect =eans.

"'t.e operator has available to h :n 2.e excera dotec_cr recordings, incers ther=cc uple readings and periodic incore flux

aces for indirectly deter:nining

=d position and flux tiles shculd the : d with the inoperable *.7/= bec=me :nalposit:. ned. '"he excere and incere' i.nstr=en a- .\\ tion will not necessar:.ly recogni=a a :nisalign:nent of 15 inches because the conce:sni ant increase in power density will nor= ally be less c.an 1% for a if inch

nisalign:nant.

The execre and incere instr.::nentation will, hewever, detect any ::d nisalignment wnich is sufficient to cause a significant increase in het channel fac:crs and/or any s:.gnificant loss in shutdown capabilitf. '"he increased surteil-lance of the core if one or mere rod positicn indicator channels is cut-of-sert:.ca series to guard against any significant 1 css in shutdown :nargin er marg:.n to core thermal limits. The history of :nalpositioned RC:7.A's indicates that.in nearly all such cases, the =alpesitioning occurred during bank :novament. Checking :=d pcsition after tank =ction exceeds 24 steps will verify that the 3 COA with the inoperable 7.ll::T is i

nov:.ng properly with its bank and the 5..ank step counter.

Malpesi:lening of an RC:".A in a stationary bank is very rare, and if it does ec=ur, it is usually gross slippage wnich will be seen by external detec crs. Sheuld it ge undetected, the i ti:ne between the red ;csition checks performed everf shift is sher: with respec i to the pr:bability of Occurrence of another indA;:endent undetected situatien which would fur her reduce the shutdcwn capability of the reds. I Any combination of misaligned rods below 10% FULL power will not exceed the ( design 14" ts. Fcr thiJ reason, it is not necessarf := chech the positi:n of l reds with incperable !.//::T's below los power: plus, the incere :.nstrur.ents:: cn :.s l l not effective for deter =ining red positten unt:.1 t.he pcwer level :.s aceve approx:.=ately 5 4. 15.3.10-16 a..

• - ' ~

15.3.11 MovAE:. IN-CORE ss-cmTAT:::N Aeplicabilitv: Applies to the operability of the movable detector instrumentation system. cbioetive: To specify functional requirements on the use of the in-core instr gentatien systems for the recalibration of the excore axial off-set detection systa=. Seecificatien: A. A minimum of 2 thimbles per quadrant and sufficient movable in-core detectors shall be operable during re-calibration of the excere axial off-set detection system. B. Power shall be limited to 90% of FULL yover if the calibration l requirements for excere axial off-set detection system, identified in Table 15.4.1-1, are not met. Basist / The Movable In-Core Instrumentation.*ystem has four drives, four detectors, and 36 thimbles in the core. The A and B detectors can be routed to eighteen thimbles. The C and D detectors can be routed to twenty-seven thimbles. Consequently, the full system has a great deal more capability than would be needed for the calibration of the ex-core detectors. To calibrate the excore detectors channels, it is only necessary that the Movable In-Core System be used to determine the gross power distribution in the core as indicated by the power balance between the top and bettem halves of the core. ( I e i Unit 1 Amendment 15.3.11-1

9 4 ATTACHMENT A SAFETY EVALUATION FOR REDUCED THERMAL DESIGN FLOW STUDY a POINT BEACH NUCLEAR PLANT UNIT 1

I. INTRODUCTION AND PURPOSE This refety evaluation has been performed to address the non-LOCA safety considerations in allowing Point Beach Unit No. I to operate with significant steam generator tube plugging. Tube plugging in sufficient nunbers may result in three effects: Reacto'r coolant flow is reduced due to increased steam generator flow resi stance. The primary flow and steam generator heat transfer area are red uc ed. Thus to maintain guaranteed steam flow, Tavg must be increased or steam pressure reduced. Primary reactor coolant mass inventory is reduced. The impact of higher steam generator. tube ~ plugging levels of up to 24 percent on the non-LOCA accident analyses presented in Chapter 14 of the FDSAR has been assessed. The basic appro'ach used was to identify the important parameters for each accident, detennine which of these param-eters were affected by the higher steam generator tube plugging levels, and then determine how the impacted parameters affected the accident ana Ty si s. The resulting impacts were determined by either evaluating the accident to qualitatively demonstrate that the accident is not limiting or by reanalyzing the affected accident (if the accident was found to be limiting or very sensitive to the impact of higher steam generator tube plugging levels). The evaluations were consistent with the following assunptions: l Maximtm' core thennal power, MWt 1381.8 Thermal design flow, gpm/ loop 84,500 S.G. tube plugging level, percent 24 T at 100 percent of maximum allowed power, *F 572.86 avg AT at 100 percent of maximtm allowed power, *F 55.5 RCS pressure, psia 2000 1.58 aH ( T "F 547 no lo ad, F0 maximum 2.52 l.

II. ACCIDENT ANALYSIS The impact of_ reduced power and flow with respect to-operation at 2000 / psia, on the non-LOCA' accident analyses presented in the Point. Beach FSAR has been assessed. 'In general, all of the transients are sensitive to initial power level, steady state primary flow, and changes in' system temperature and pressure. A study was made of each currently applicable accident analysis to identify margins.to safety limits which could be used to offset penalties due to reduced primary flow. Reduction in system power is a cenefit in DNB calculations and more than offsets the flow and T (relative to reduced power) penalties. 3y The most recently applicable analysis used in this report is indicated by the reference number after each title. I Uncontrolled RCCA Withdrawal From a Subcritical Conditio'n 1) A control rod assembly withdrawal incident when the reactor is subcrit-ical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14.1.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the negative reactivity feedback of the Doppler power coefficient. The power excur-sion causes a heatup of the moderator. However, since the power' rise is rapid and is followed by an immediate reactor trip, the moderator temoer-ature rise is small. Thus, nuclear power response is primarily a function of the Doppler power coefficient. The reduction in primary coolant flow is the primary impact which influences this accident. The reduced primary coolant flow results in a decreased core heat transfer coefficient which ir turn results in a faster fuel temperature increase than reported in the most recent analy-sis.III The fast temperature increase would result in more Doppler feedback thus reducing the nuclear power heat flux excursion, as pre-sented in Reference 1, which would partially compensate for the flow reduction. Therefore, the nuclear transient is only moderataly :ensi-tive to the inpact of steam generator tube plugging. The most recent analysis (I) shows that for a 90 x 10-5 Ak/sec reactivity insertion rate, the peak heat flux achieved is 76 percent of nominal. This is conservative for the higher plugging situation for the reasons stated above. The resultant peak fuel average temperature was 772*F. A 5 percent reduction in flow and the associated reduction in core heat transfer coefficient would degrade heat transfer from the fuel by a maximum 5 percent and increase the rise in peak fuel and clad temperature by a maximum of 5 percent. Therefore, the fuel and clad temperat res would be less than ~784 F and ~617*F, respec-tively, for the present evaluation. These values are still significantly below fuel melt (4900*F) and zirconium-H O reaction (1800 F) limits, and the impact of 2 increased steam generator tube plugging, up to 24 percent would not result in a violation of safety limits. Uncontrolled RCCA Withdrawal at Power (2) An uncontrolled control rod assembly withdrawal at power produces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature (Section 14.1.2 of FSAR). Reduced flows result in less margin to DNB. Reduced thermal power results in more margin to DNB. In addition, the reduced primary flow will increase loop transit time which could require new values of lead / lag time constants to be determined for the overtemperature AT set point equation. Thu's to assure adequate core protection the Reactor Core Thermal and Hydraulic Safety Limits have been recalculated consistent with the reouction in RCS flow and thermal power. The resulting overtemperature AT protection limits were essentially unchanged. Based on the current overtemp-erature AT limits, new core limits, reduced RCS flow and reduced rated power, the accident has been reanalyzed to verify the adequacy of protection setpoints and the lead / lag time constants. l Method of Analysis The transient was reanalyzed employing the same digital computer code and assumptions regarding initial conditions and instrumentation and setpont errors used for the FSAP, including: _

1. power levels equal to 102 percent, 62 percent, and 12 percent of 1381.8 MWT; 2. Average temperature 4"F above T corresponding to the initial avg power level; 3. Pressure (1970 psia) 30 psi below nominal; i 4. Reactor trip on high nuclear flux at 118 percent of 1381.8 MWT with trip delay of 0.5 seconds; and 5. The setpoints for the overtemperature aT reactor trip function are those which presently appear in the Technical Specification cur-rently for 2000 psia operation, with allowances for instrumentation errors. trip delay time of 2.0 seconds was used. 6. Nominal flow is 84,500 gpm/ loop. 7. No credit is taken for the high pressuri:er water level and high pressure reactor trips. Results Figures 1 through 3 show the minimum DNBR as 1 function of reactivity insertion rate for 102 percent, 62 percent and 12 percent of full power. Conclusions These results demonstrate that the conclusions presented in the FSAR are still valid. That is, the core and reactor coolant system are not l adversely affected since nuclear flux and overtemperature ai trips pre-vent the minimum DNB ratio from f alling below 1.30 for this incident. Thus the current setpoint equation and reduction in rated power compen-sate for the reduction in theraal design flow. Malpositioning of the Part length Rods (2) A malposiitiong of a part length rod accident need not be addressed since the part length rods have been removed from the core. Rod Cluster Control Assembly (RCCA) Drop ( } The drop of a Control Rod Assembly results in a step decrease in reactivity which produces a similar reduction in core power, thus reducing the coolant average temperature; The highly negative moderator temperature coefficient (-40 pcm/*F) assumed in the analysis results in a power increase (overshoot) above the tur-bine power runback value causing a temporary imbalance between core power and secondary power extraction capability. The effect of 5 percent reduction in initial RCS flow would be a smaller reduction in coolant average temperature and less of a power overshoct. Statepoints were evaluated consistent with a 5 percent reduction in flow and a 9 percent reduction in power. The reduction in power results in additional DNB margin. The resulting DNB evaluation showed that the DNBR limit of 1.30 can be more than accommodated with margin in the cur-rent cycle. The conclusions in the FSAR remain valid. Chemical and Volume Control System Malfunction (2) For a boron dilution. incident during refueling or startup, while the reactor is subcritical, Section 14.1.4 of the FSAR shows that the operator has sufficient time to identify the problem and terminate the-dilution before the reactor becomes critical. Tube plugging has no effect on the analysis at refueling conditions or cold shutdown conditions since only the reactor vessel and RHR system volumes are considered. For a dilution during startup, the effective volume of primary coolant in the steam generator tubes has been reduced by 3 '24% (~323 ft ). Thus the volume of the reactor coolant (excluding the pres-3 3 surizers) is reduced from 5253 ft to 4930 ft. However, the minimum dilution time has baen recalculated for refueling and startup assuming a minimum boron concentration of 1800 ppm, as opposed to 2000 ppm assumed in the FSAR. This will result in a shorter time to dilute to the maximum critical boron concentration of

l ' i 1130 ppm at refuelin; and 1600 ppm at startup. The minimum time required for the reactor to become critical at refueling and startup has been calculated to be 74 minutes and 23.9 minutes respectively. Thus adequate time is available for the operator to recognize and terminate the dilution flow from refueling and startup conditions. For dilution at power, it is necessary that the time to lose shutdown margin be suf ficient to allow identification of the problem and termina-tion of the dilution. As in the dilution during startup case, the RCS volume reduction due to steam generator tube plugging must be con-sidered. The effective reactivity addition rate is a function of the reactor coolant temperature and boron concentration. The reactivity insertion rate calculated is based on a conservatively high value for the expected boron concentration at power (1400 ppm) as well as a con-servatively high charging flow rate capacity (181.5 gpm). The reactor is assumed to have all rods out in either automatic or manual control. With the reactor in manual control and no operator action to terminate the transient, the power and temperature rise will cause the reactor to reach the Overtemperature si trip setpoint resulting in a reactor trip. Af ter reactor trip there is at least 15.1 minutes for operater action prior to return to criticality. The boron dilution transient in this case is essentially the equivalent to an uncontrolled rod withdrawal at power. The maximum reactivity insertion rate for a boron dilution transient is conservatively estimated to be 1.15 pcm/sec and is within the range of insertion rates analyzed for uncontrolled rod withdrawal at power. Prior to reaching the Overtemperature si reactor trip the opera-tor will have received an alarm on Overtemperature si and turbine run-back. With the reactor in automatic control, a boron dilution will result in a power and temperature increase such that the rod controller will attempt to compensate by slow insertion of the control rods. This action by the controller will result in rod insertion limit and axial flux alarms. The minimum time to lose the 1 percent ik/k shutdown margin required at beginning-of-life would be greater than 15.1 ainutes. The time acaid be

significantly ' longer at end-of-life due to the low i-nitial baron concen- ~ tration and 2.77 percent ik/k shutdown margin. I2) Rup:ure of a Steam Pipe The steamline break transient is analyzed for hot zero power, end' of life conditions (Section 14.2.5 of the FSAR) for the following cases: Hypothetical Break (steam pipe rupture) Inside Containment with and without power Outside Containment with and without power Credible Break (Safety valve opening) A steamline break results in a rapid depressurization of the steam gen-erators which causes a large reactivity insertion to the core.via primary cooldown. The acceptance criteria for this accident is that no DNB must occur following a return to power. This limit, however, is highly conservative since steam line break is classified as a Condition IV event. As such, the occurrence of DMB in small regions of the core (-5 percent) would not violate NRC acceptance criteria. The impact of increased levels of steam generator tube plugging would affect the accident principally due to the reduced flow, reduced RCS inventory, and reduced heat transfer coefficient. These impacts would result in changed cooldown and feedback reactivity characteristics such i that the return to power as shown in the previous analysis would be slightly conservative with respect to the lower initial flow condi-tions. In addition, the time of Safety Injection actuation would be unaffected by flow conditions for the Hypothetical Breaks. This coupled with the slightly ' slower return to power would result in a reduction in peak average power for the cases with and without power and indicate i results conservative with respect to the current analysis. 4 i e

However, as this is a limiting accident with respect to available DNB margin at reduced pressure, the limiting cases were reanalyzed and limiting statepoints evaluated. ./ Method of Analysis Analysis methods and assumptions used in the reanalysis were consistent with those employed in the most recent safety analysis. These assump-tions included: 1) Minimum shutdown margin equal to 2.77 percent. 2) The most negative moderator temperature coefficient for the redded core at end of life. 3) The rod having the most reactivity stuck in its fully withdrawn po si tion. 4) One train of safety injection fails to function as designed. Results The minimum value of the DNBR for the hypothetical breaks was greater than the 1.30 limit. Results for the credible break confirmed that the core remained subcritical throughout the transient. Table ' presents the cora parameters for the 4 hypothetical break cases used in DH3 evaluations. Figures 4 through 7 present the transient results for those cases summarized in Table I. Figure 8 presents the transient results for the credible break. Conclusions The steamline rupture accident has been shown to meet the ONB design basis for the hypothetical breaks and remains subcritical for the credi-ble breaks for the 24 percent tube plugging. _ ~ Startuo of an Inactive Reactor Coolant Looc(2) ~ l Startup of an idle reactor coolant pump results in the injection of relatively cold water into the core. This accident need not be addressed due to Technical Specifications restrictions which prohibit power operation with a loop out of service. However, a brief dis'cussion of the impact of the new operating conditions is included. The analysis in FSAR Section 14.15 shows that the reactor does not trip and reaches a peak power of 21 percent full power. The lower loop flow would result in a slightly lower reactivity insertion rate, resulting in a lower peak heat flux. Therefore results presented in the FSAR would be conserva-ti ve. I2I Reduction in Feedwater Enthalpy Incident The addition of excessive feedwater and inadvertent opening of the feed-water bypass valve are excessive heat removal incidents which result in a power increase due to moderator feedback. Increased levels of steam generator tube plugging would impact this analysis principally due to the reduced flow. i Section 14.1.6 of the FSAR presents two cases. The first case assumes a zero moderator coefficient, which is used to demonstrate inherent tran-sient attenuation capability during a feedwater reduction. A reduction in flow will have a negligible effect on stability since the reactivity insertion is identical to the FSAR case due to the zero moderator tem-perature coef ficient. DNB is not a consideration for this case since DNBR's do not fall below the steady state value. This is due to the relatively large reduction in T The reduction in flow is more avg. j than compensated by the reduction in nominal power, resulting in an increase in the initial steady state DNBR. In addition, as discussed in the FSAR, the reactor will trip on low pressure trip. i-The second case assumes a large negative moderator coef ficient. The i reduction in thermal design flow aill result in a slower cooldown, and therefore the reactivity insertion rate will be less than in the FSAR _9 r--., ,,r,,,

0 analysis. The integral reactivity insertion due to moderator temperature l reduction will be less than the FSAR case, thus producing a lower peak nuclear power. The reduction in nominal power results in a net increase in steady state DNBR. Protection for this accident is provided by the overpower-overtemperature AT protection. The adequacy of this protection was verified in the rod withdrawal at power reanalysis. Excessive Load Increase Incident (2) An excessive load increase incident is defined as a rapid increase in steam generator flow that causes a power mismatch between the reactor core power and the steam generator load demand. Four cases were analyzed in the FSAR, Section 14.1. 7. A 10 percent step load increase was analyzed for manual and automatic control, at beginning of life (BOL) and end of life (EOL). As in the Feedwater Malfunction Accident, reduced flow is the principal impact on this accident due to increased levels of steam generator tube plugging. The worst case results (automatic control-E0L) indicate that with no trip actuation, steady state conditions are reached with a minimum DNBR of > 1.30. The reduction in thermal design flow will result in a slower ~ cooldown, and therefore a lower reactivity insertion rate. The integral reactivity insertion due to moderator temoerature will be less than the FSAR case, thus producing a lower peak nuclear power. The reduction in nominal power results in a net increase in steady state DNBR. Protection for this accident is provided by the overpower-overtemperature >T protection. The adequacy of this protection was verif t-d in the rod with-drawal at power reanalysis. Loss of Reactor Coolant Flow / Locked Rotor (2) As demonstrated in the FSAR, Section 14.1.8, the most severe loss of flow transient is caused by the simultaneous loss of electrical power to both reactor coolant pumps. The reduced thermal power results in a net increase in initial steady state DNB ratio. The increased steam genera-tor tube bundle resistance has an extremely small impact on the flow coastdown during the critical first few seconds of the transient. Therefore, the reactor trip on low reactor coolant flow will be gen-erated at approximately the same time in the transient. With the' higher initial DNS ratio and same time to trip, a loss of flow event.frca these condit' ions will result in more margin to the DNB. limit. This was veri-fied by evaluation of the statepoints consistent with a 5 percent reduc-tion in flow and 9 percent reduction in power. The resulting DN3 evaluation showed that the DNBR limit of 1.30 can be more than acccmmo-dated with the margin in the current cycle. The conclusions in the FSAR remain valid. The FSAR shows that the most severe Locked Rotor Accident is an instan-taneous seizure of a reactor coolant pump rotor at 100 percent power. Following the incident, reactor coolant system temperature rises until shortly after reactor trip. The impact on the Locked Rotor Accident of increased steam generator tube plugging will be primarily due to the reduced flow. These impacts will not affect the time.to DNB since DNB is conservatively assumed to occur at the beginning of the transient. The flow coastdown in the affected loop due to the Locked Rotor is so rapid that the time of reactor trip (low flow setpoint is reached) is essentially identical to most recent analyses. The nuclear power and heat flux responses will be somewhat lower due to reduced thermal power. The reduction in power t l also results in reduced initial hot spot values. This would of fset the j slight increase in fuel and clad temperatures effect of reduced flow. Consequently, the expected peak fuel and clad temperatures would remain l the same as results of the currently applicable analysis. j It is estimated that the peak pressure will not increase above -the pre-vious value due to reduced power, however the maximum calculated value ~ i was 2778 psia based on 2250 psia operation plus 3D psia uncertainty. l This is significantly below the pressure at wnich vessel stress limits l l ' l

are exceeded. In addition, this is conservative as noted by the con-clusions of the FSAR. The 24 percent reduction in steam generator tubes would result in approximately a 8 percent reduction in primary mass which decreases the heat capacity of the RCS by the same amount. This would not result in higher peak temperatures or pressures since the peak values are reached in considerably less than one loop transport time constant. Thus operation at reduced flow wi.1 not cause safety limits to be exceejed for a locked rotor accident. Loss of External Electrical Load (2) The result of a loss of load is a core power level which momentarily exceeds the secondary system power extraction causing an increase in core water temperature. The impact of increased levels of steam generator tube plugging would be again principally due to the reduced flow and the decreased RCS mass i nvento ry. Two cases, analyzed for both beginning and end of life con-ditions, are presented in Section 14.1.9 af the FSAR: a. Reactor in autcmatic rod control with operation of the pressurizer spray and the pressurizer power operated relief valves; and b. Reactor in manual rod control with no credit for pressurizer spray or power operated relief valves. The FSAR analysis results in a peak pressuri:er pressure of 2514 psia following reactor trip. A reduction in loop flow and RCS mass inventory will result in a more rapid pressure rise than is currently shown. The effect will be minor, however, since the reactor is tripped on high pressuri zer pre:sure. Thus, the time to trip will be decreased wnich will result in a lower total energy input to the coolant...

The minimum transient DNBR evaluated'at reduced pressures of 2000 psia minus 30 psia uncertainty (DNBR is more limiting at reduced pressure) will result in a net increase due to reduced power. However, this tran-sient is bounded by the Uncontrolled Rod Withdrawal at power transient. Less of Nomal Feedwater/ Station Blackout (2) This transient is analyzed to detemine that the peak RCS pressure does not exceed allowable limits and that the core remains covered with water. These criteria are assured by applying the more stringent requirement that the pressurizer must not be filled with water. The effect of reducing initial core flow would be a larger and more rr.pid heatup of the primary system. The resulting coolant density ch.unge would increase the volume of water in the pressurizer. The a< alyses in FSAR Section 14.1.10 and 14.1.11 show that the peak pressurizer volume 3 3 reached is 780 ft on an approximate 250 ft change in volume. This result was due to a - 26*F change in coolant average temperature. Using the highly conservative assumption that the average temperature would increase 50 percent due to flow reductions, this would result in a maxi-3 taum increase of less than 125 ft in liquid volume. This is still 3 below the 1000 ft capacity of the pressurizer, thus no reanalysis is In addition, due to the relatively long duration of the necessary. transient following trip, the results are highly sensitive to residual (decay) heat generation. Residual heat generation is directly propor-l tional to initial power level preceding the trip. At reduced power, the total energy input to the system be likewise reduced. III Rupture of a Control Rod Drive Mechanism Housing, RCCA Ejection The rupture of a control rod mechanism housing which allowed a control l rod assembly to be rapidly ejected from the core would result in a core themal power excursion. This power excursion would be limited by the Doppler reactivity effect as a result of the increased fuel temperature and would be teminated by a reactor trip activated by high nuclear power signal s.

The rod ejection transient is analyzed at full pcwer and hot zero power for both beginning and end of life conditions (Section 14.2.6 of the FSAR). Reduced core flow is the primary impact resulting from increased levels of steam generator tube plugging. This impact would result in a reduction in heat transfer to the coolant which would increase clad and fuel peak temperatures. Reanalysis was performed using the conservative ejected rod worths and post ejection peaking factors assumed in the latest analyses wnich are above the calculated Point Beach reload values. Reanalysis was per-formed to show that the increase f.n initial F from 2.47 in the pre-g vious analysis to 2.52 is still acceptable. idethod of Analysis Analysis methods and assumptions used in the reanalysis were consistent with those employed in the most recent analysis and FSAR 14.2.6. The calculation of the transient is performed in two stages, first an aver-age core calculation and then a hot region calculation. The average core is analyled to determine the average power generation with time including the various total core feedback effects, i.e. Ooppler reactiv-ity and moderator density reactivity. Enthalpy and temperature tran-sients in the hot spot are determined by addinj c multiple of the aver-age core energy generation to the hotter rods and performing a transient heat-transfer calculation. The asymptotic power distribution calculated without feedback is pessimistically assumed to persist throughout the transient. The DNB time is not calculated. DNB is conservatively assumed to occur near the start of the transient. Results The analysis results and inputs are summarized in Table II. The condi-tions at the hot spot fuel rod do not exceed the limiting fuel cri-teria #) The conclusions of the FSAR, therefore are still valid. ~ I O 4

III Conclusions To assess the effect of non-LOCA accident analyses on operation of Point Beach Unit 1, with significant levels of steam generator tube plugging, a safety evaluation was performed. The transients and/or statepoints were analyzed for rod ejection, steam-lire break, boron dilution, dropped rod and loss of flow. In addition, an evaluation was performed to identify the effect of the reduced operating conditions (power, flow and pressure) on the remaining tran-sients and to quantify margins available to affect penalties. Those accidents that are sensitive to nigher operating pressures were addressed al so. Based on this evaluation, operation at these reduced conditions and a maximum 24 percent effective steam generator tube plug-ging level will not result in violation of safety limits for the tran-sients evaluated at either 2000 psia or 2250 psia operation. REFERENCES 1. Davidson, S. L., Editor, " Reload Safety Evaluation Point Beach Nuclear Plant Unit 1, Cycle 9A," December 1980. 2. Fin &1 Safety Analysis Report Point Beach Nuclear Plant, Unit's Number 1 and 2. 3. Davidson, S. L. Editor, " Reload Safety Evaluation Point Beach Nuclear Plant Unit 2, Cycle 9A," May 1982. 4. Risher, D. H. Jr., "An evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods", WCAP-7588, Revi sion 1-A, January,1975. TABLE I CORE PARAMETERS USE0 IN STEAM LINE BREAK DN8 ANALYSl5 Outside Break With Power outside Break Without Power inside Break With Power inside Break Without Power Time (Sec.) 72.4 75.2 78.4 81.6 104.8 10 8.0 111.2 114.0 58.4 61.2 64.8 67.6 94.4 97.6 102.4 107.2 Core inlet Temperature (*F) Loop A 403 401 399 398 361 357 355 352 356 355 353 351 295 292 288 285 Loop B 454 451 448 446 489 489 488 487 436 434 430 427 509 508 50 7 507 RCS Flow 100 100 100 100 10.1 9.8 9.6 9.4 100 100 100 100 10.9 10.6 10.2 9.8 (percent of nominal) Heat Flua 21.8 22.9 24.2 18.5 10.4 10.5. 10.8 10.5 39.3 41.0 42.9 34.6 17.0 17.2 17.3 16.9 (percent of 1381.8 Mwt) RCS pressure 595 694 691 689 1039 1045 1053 1061 656 656 655 653 1008 1020 1027 1030 (pstal e." d 't a

TABLE II

SUMMARY

OF RCD EJECTION ANALYSIS PARA.NETERS AND RESULTS BOL BOL EOL EOL Power Level, percent 0 10 2 0 102 Ejected rod worth, percent so 0.91 0.34 0.95 0.42 Delayed neutron fraction, percent 0.0049 0.0049 0.0046 0.0046 2.52 2.52 Fg before rod ejection 11.2 5.03 13.7 5.10 Fg after rod ejection Humber of operating pumps 1 2 1 2 Maximum fuel pellet center 3504 4543 3719 4458 temperature, 'F Maximum fuel pellet average 3095 3492 3289 3392 temperature, 'F Maximum clad average temperature, *F 24a0 2129 2550 2071 Fuel Pellet Melting, percent 0.0 0.0 0.0 0.0 Maximum fuel enthalpy (btu /lb) 231.4 266.7 248.7 257.8 6

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-2 5000 % '

tv -3.0000 ~ = c g a a g s 2 a 4 O o = [ b 2 C o e !!wE (SEC) FIGuitE 6 e a

STEAMLI;1E BREAK IllSIDE THE C0tlTAIrlME.'IT (AT EXIT OF STEAM GEilERATOR) LOSS OF OUTSIDE POWER AT T = 0 600.00 1 550.C0 g s. T: 500.00 t ? "a i $ $ 4!0 00 t, w w 3.., 400.CO ? I v. 350.00 4 300 00 2000.O 2 1800.0 - \\ 3 l S 1600.O <- ? \\ l. wa s 5i 1*00.0 e f 1200.0 + T 1000.00 N = 300.CO 3.0000 9 1 2 e 2 5000 = W wa$5 2.0000 1.5000 1 -= 3 xo a w .T. *r 1.C000 EO 50000 T, o e w 0.0 40000, i Af /m N I i l} y-50000 j [ g w s 4 t -1 C000 v N,. I il s ! -t 5000 & [ ? \\ . i.' -2 0000 +i + v. e y -2 5000 g -3.0000 g 3 = a = = o -g = o = = = o = o o e o a 2 c o ~ o 19E t!Et' FIGURE 7

STEAM BREAK EOUI'/ ALE:lT TO OtlE STEAM GEllEMTOR SAFETY '/AL'/E 400 00 575 00 - !!O 00 ll - =~ !25 00 + + ~O < 9 500 CO - II 475.00 t o*- 3 450.CO ; 425 CO 1 400 CO ' OJ 6 3 -t 0000 - 1 d,-2.0000 / s -3. 0000 + i s 1 -* 0000 + s O I \\ {-50000 \\[ t 4.0000 - 2000.0 NN I 1750 0 t m t 1500.0 t x m \\ 4 i a i o s, !250 0 ? 1 I 1000 00 7 \\ u W

0 0 '

2 <-w 0 = = 5- .2 =, o I o 9 2 8 3 5 r, a e a 1 IIP! I'd { } l FIGUPE 1 i e l l l}}