ML20069G395
| ML20069G395 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/17/1982 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20069G393 | List: |
| References | |
| NUDOCS 8209290019 | |
| Download: ML20069G395 (61) | |
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. 2) Cold Shutdown
":he reactor is in the cold shutdewn condition when the reacter -
i . has a shutdown margin of at least 1% Ak/k and reactor coolant temperature is < 2000F.
- 3) Refueling Shutdown
-*he reactor is in the refueling shutdown condition when the reactor 0 A refueling is suberitical by at least 10% Ak/k and
- avg is <140 F.
shutdewn refers to a shutdown to move fuel to and from the reacter core.
- 4) Shutacwn Margin Shutdown margin is the instantaneous a=ount of reactivity by which the reactor core would be subcritical if all witndrawn control reds were tripped into the core but the highest worth withdrawn RCCA remains fully withdrawn.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.
- h. Power operation The reactor is in power operating condition when the reactor is crttical and the average neutron flux of the power range instrumentation indicates l
l greater than 2% of FULL ;:ower. l (
- i. Refueling Cceration_
Refueling operation is any operation involving movement of core components (those that could affect the reactivity of the core) with:.n ene contain-ment when the vessel head is unbolted or removed.
- j. Rated Power of Rated power is here defined as a steady state reactor core output l 1518.5 MWT.
- k. Thermal Power Thermal power is defined as the total core heat transferred from sne 8209290019 820917 fcel to the coolant. DRADOCK05000g
o I' l L. Dez ee e! Redundsnc~r Degree of redundanc r is defined as the difference between -de nu:nber of operable channels and the mN-u= nu:nbar of ch'annela which when tripped vill cause an autenatic shutdevn.
- n. Reactor Critics.1 The reactor is said to be critical when the neutron chain reaction is self-sustaining tnd k er., = 1.0.
- n. Lev Pover Cee stion
'"he reacter is in the icv power operating condition when the reacter is critical and the average neutron flux cf the pover range ins,r mentation indicates less than er ec,ual to 25 of FULL power.
- c. Fire Sureressien *,iater Svstem A CRE SGPRESS CN E'"IR SYS*"ZM shall censist of: a water scurce:
pu=p (s ) ; and distribution piping with associated sectionalizing centr:1 cr isciaticn valves. Such valves shall 1.nclude yard pes: indica ing valves and the first valve ahead of the water flev alarm device en each sprinkler, hose standpipe er spray system riser.
- p. Full Power Full pcuer is detined as 100% of rated power when the RCS flow is
> 178000 gpm. When RCS total flow is < 178000 gpm, full pcwer is defined to be 91% of rated power.
15.1-5
i re e 15.2.3 SAFITY ZC E IDCTDIG SArr Y SYSTZM Srt CiGS 15.2.1 SAFr:T !.,2CT, REACTOR C::RE Applicabiliy/ Applies to the limiti=g combinaticas of ther:nal power, reae:cr cecipe system pressu=s, and coolant temperatura during operation. Cb3octives' To :nau: aln the integrity of the fuel cladding. Specification:
- 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 rate < 178000 gpm. The safety li=it is exceeded if the point defined by the ccabination of reactor coolant system average temperature and power level 0 is at any time above the appropriate pressure line.
,m 2......
mm m
3as s:
*c naintain the integr:.:7 cf the. fuel cladding and pr3 vent fiss:.cn pr due: ,
)
release, it is necessary :c prevent overheating cf de cladding under all , cporating condi :cns. ~his is acchaplisned by operating the het ree cns l of de core within :ne nucleate boiling regime of heat transfer, whe ein ce i hcot ::ansfer coefficient is ver/ large and the clad surface temperature is i enly a Eew degrees Fahrenheit above the cociant saturat:.cn temperature. O.e upper boundary of the nucleate boilinc, regime is termed depar."-* '--m nuclease boiling (DNB) and at this point there is a sharp reduc icn of de hea: transfer ccafficient, which would result in high clad temperatures and the pcssibilit/ cf clad failure. ONB is not, however, an observable parameter during reacter ep; ration. Therefore, the observable parameters; thermal pcwer, reac:cr cociant temperature and pressure have been related to ONB througn the W-3 ONB ccrrelation. O.e W-3 CNB correlation has been developed to predict the :G flux and the location of CNB for a::ially unifem and non-unifcrm heat f '.ux distributions. O.e local CNB heat flux ratic, defined as de ratic of tha heat flux that would cause CNB at a particular core location to the local heat flux, is indicative of the margin to CNB. O.e sinimum value of the ONB ratic, ::NBR, during steady state operatien, nemal operational transients, and ant.c:. pated transiente is limited to 1.30. A CNB ratic cf 1.30 ccrresponds to a 95% pretability at a 95% ccnfidence level that CNB will not occur and :.s :ncsen as an apprcpriate =argin to ONB for all cperating conditions. III The curves of Figure !.5.2.1-1 and 15.2.1-2 represent the loci of points of thermal power, coolant system pressure and average temperature for which the DNB ratio is not less than 1.30. The area of safe operation is below these lines. The safety limits curves have been revised to allow for heat flux peaking effects due to fuel densification and flattened fuel cladding sections.
'= *- , -
Additional peaking factors to account for local peaking due to fuel rod axial 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 are based on an FNg of 1.58, cosine axial flux shape, and a DNB analysis as described in Section 4.3 of WCAP-8050, " Fuel Densification, Point Beach Nuclear Plant Unit 1 Cycle 2" (including the effects of fuel densification and flattened i 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: Fg = 1.58 {1 + 0.2 (1-P)} where P is a fraction of FULL power when P < 1.0 Fg = 1.58 when P > 1.0. The effects of rod bow have been included in the determination of a conserv-ative value for Fg. Rod bow effects of up to 14.9% DNBR are offset by credits available from the design limit DNBR, pitch reduction, design thermal diffusion coefficient and the fuel densification power spike, which were pre-viously approved.* The hot channel factors are also sufficiently large to account for the degree of malpositioning of full-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. Ross and D. G. Eisenhut, USNRC, to D. B. Vassallo and K. R. Goller, " 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 l
i 4 l
'l 1
l l l l .
-~
680 t i 660 2 W PSIA 2000 PSI A 640 - 2000 P3t& l ! 620 - 1 I C 1700 PSIA
- L a, 600 -
j - E i 580 LOCU3 OF PCI RT3 l 560 Foe niCx STIAw GO ER A TOR S AFT TY VALYE3 ARE ACTUATED 540 520 0 20 40 60 80 100 120 POWER (PERCENT OF FULL)
, TOTAL RCS FLCW >_ 173000 GFM Figwe l'.2.l-1 Cm DNS Sdary Umin Poit' Seacn Units 1 and 2
( -- _ ___ _i m ___ -
690 - 6 70 - - 2400 PSIA
$50 - - -.-
2200 PSIA 620 - - -- 2000 PSIA E
,en 4 610 - - 1775 PSIA --
590 - - - 570 - - 550 . . e e a e e n 0 20 40 60 80 100 120 POWER (PERCE.'IT OF FULL) TOTAL RCS FLOW <178000 GPM FIGURE 15.2.1-2 CORE ONS SAFETY LIMIT 5 m -e.cu . .. , r e ,
b 15.2.3
- 2 C IG SATIS S*.'s Di gr-==7c3, pp.cn :- yg ntm 33;; g.; ;3 A=elicabilit*/:
Applie to trip settings fer instr.:monts menitcring react =r ;cwer.and reac== coolant pressure, temperature, flow, pressurizar level, and pe==issives related
.c resce r pr=tection.
chieceivet 2 provide for automatic protec_1ve action in the event that the principal process variables approach a safety if d t. Seee _ficatien: d
- 1. Protective instrunantation for reacter trip settings shall be as follcws:
A. Startup protaction (1) High flux, source range - within span of source range instrumentation. (2) High flux, intermediate range - 140% of FULL power. (3) Eigh flux, power range (low set point) - 125 % of FULL power. B. Core 14~4: protaction (1) High flux, power range (high setpoint)
< 108% of FULL pow'2r (2) High pressurizar pressure - 1 2285 psig.
l 1 l l 15.2.3-1 l
(3) ~.cw pressuriter pressure - > 1790 2517 fer Operat:On at 2000 psia pr.marf system pressura (4) Over emperature a; (kris) - K.
<a 70 (K1-K3 (T-T ' ) #
(p-P ' ) - f (a ) )
- 1+;;S where ATo = indicated IT at WLL power, '?
T = a*.ersga temperature, 'T T' = 574.2 'T p = pressurirer pressure, psig P' = 2235 psig Kt < 1.30 for operation at 2000 psia primary system pressure K, = 0.0150 K3
= 0.000791 T1 = 25 sec.
r2 = 3 sec. and f(AI). is an even function of the indd rated difference between top and bottcm detectors of the power-range nuclear ion chambers; with gains 4 be selected based en =easured instr. ment response during plant startup tests, where ge and qb are the percent pcwer in the ecp and bottcm halves .cf te 4 core respectively, and qe + qb i3 ECC31 CUI" EC"*# i^ F"#C"UC of FULL power, such that: l (a) for qt
- 9b within -17, +9 percent, f (11' = 0.
(b) for each percent that the =agnitude of ge-qb exceeds -9 percent the AT trip set point shall be aut==at cally reduced by an equivalent of two percent of FULL power. 15.2.2-2
(c) for each percent that the magnitude of q - qb exceeds -17 percent the iT trip setpoint shall be automatically reduced i by an equivalent of two percent of FULL power. . s (1.3. (5)] overpower AT 3 t 1 AT. [K4-K5 T3S g- , . T - 4 (T-T ' ) -f (A-)] where l dT. = indicated AT at FULL power, *F T = average temperature, *T T' - 574.2 K4 1 1.089 of FULL power
.K5 = 0.0262 for increasing T
= 0.0 fcr de h T
= 0.00123 for T > T' K6
= 0.0 for T < T'
=
T3 10 sec. f (AI) as defined in (4) above, (6) Undervoltage - > 75% of noz:nal voltage (7) Mw indicated reactor coolant flow per loop-
>9C% of normal indicated loop f'ww (8) Reactor coolant pump :nowr breaker open (a) Low frequenc/ set pcint >57.5 cps
. (b) M w voltage set point >75% of ner=al voltage
11
- 2. Protective instr.=e .tation se .. gs for reac cx cTip inter-Iceks shall be as felicws:
A'. "'he "at power" reacter trips flow pressuri.:er pressure, high pressurizar level, and ic,w rsacecr coolant ficw for > both loops) shall be unbiccked whan: (1) Pcwer range nuclear flux .>. 94 (*11) of ?JLL power or (2) ?.trbine Ioad > 10% of full-1 cad turbine pressure.
- 3. Se single loss of ficw trip sha.11 he unblocked when the power range nuclear flux > 50% of 7L'LL power. -
C. "'ha power ra.nge high flux level low range trip, and intermediate rsnge high flux level trig shall be unbiccked / when power.'is _< 9% (+1%) of FULs power. D ." na source range high flux reac*or.crip sna11 be unbiccked i when the inter =ediata range flux is < 10 ~o amperes. ( f
?
15.2.1-4
/
o z
power distributa=n, the reac. : trip '_d- ., w;. h allewance fer errers, (2) If is always below the c=re safety li: sit as shewn en Figure 15.2.1-1. ax a1 peaks ars greater than design, as ind1=ated by diff erence be: ween
=p and be::::n power range nuclear detectors, the reactor =ip id- :
is au:=matica11y reduced. (6) (7) O.e everpower , overtempera =e and pressuriser pressure system se points have been revised to include effec of reduced system pressure operat:. n The r'avised setpoin s as (including the effects of fuel densificatien) . given above will not exceed the revised core safety limits as snown in rigure 15.2.1-1 and 15.2.1-2. Se overpower 11:21: criteria is that core pcwer be prevented f cm reaching
~he reae:==
a value at which fuel pellet centerline maltang would cecur. is prevented frem reaching the everpewer limit conditien by ac: en of the nuclear everpower and overpower ST =ips. in which n e high and icw pressure reac == = ips li=it the pressure range sact= operation is per:nitted. Se high pressura:e pressure reac:=r trip set ing is lower than the set pressure for the saf ety valves ( 4Ef psag) such tha: te reae c: is =1pped before the sade:7 valves actuate. he icw p assursser prassure reacecr =ip rips the reactor :.n the unlikely event of a less-of-coolant accident.(4) The icw flow reacter cip protects the core agains: CIE :.n the event cf 1 in the loeps er a st:dden icss either a decreasing actual =easured flew O.e set poin s pec:.f .ec of pcwer to one or both react =r coolant pumes. w h the value used in the ace. dent analysis. III he icw as cons:. sten: as :neasurec locp ficw signal is caused by a =endiesen of less enan 90s flew
*he less Of pcwer s:.gnal :.s causec :y by the locp fiev :.ns c.:=ents::.cn.
3e ,- -
- _ - _ - _ _ _ _ _ _ _ _ _ _ - - - - _ _ _ moaoJ.@
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. (') Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere sith 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-flow ratio will be maintained equal to or less than unity, which 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 7 , . -gw- . - myv - - - - -
15.2 *:P. TING COit:TIONS FOR CPr?ATION 13.1.1 REA CCR CCCLANT s'IS m Acelicabilitv l Applies to the operati.ng status of the Reactor ccclant System. 1 Ohaective Co spc:ify those limiti.ng conditions for operac.cn of the Reactor Ccelant System which :nust a met to ensure safe reactor ope. ration. Seecification A. OPERAT:CNAL COMPCNENTS Specification:
- 1. Coolant Pumps
- a. At least one reactor coolant pump or the residual heat removal system shall be in operation when a reducticn is made in the boren concentration of the reac cr ccolant.
- b. When the reactor is critical and above is Full power, except for natural circulation tests, at least one reacter coolant pump shall be in operact.cn.
- c. (1) Reactor power shall not be saintained ateve 10%
of FULL power unless both reactor coolang pumps are in operation. (2) If either reactor ecolant pu=p ceaser :peratinc, immediata power reduction st.all be in:. .; aced under administrative centrol as necessar/ :o recuce pcwer te less than 10% of FULL power.
- 2. Steam Generator
- a. One steam generator shall be Operacle wnenever :ne aversce reac or coolant tempersture :.s aeove 350*F.
15.3.1-1
4 because of the low pressurizer volume and because pressurizer boren concentration normally will be higher than that of the rest of the reactor coolant. Part 1 of the specification requires that a sufficient nu=ber of reactor coolant pumps be operating to provide core cooling in the event that a 1 css cf 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 l removed with natural circulation only; hence, the specified upper limit 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 re=ove decay heat and thereby control system temperature and pressure. If no residual heat is re=cved 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' capacity. One valve therefore provides adequate defense against over-pressurization. Part 1 c(2) per=its an orderly reduction in power if a reactor coolant pu=p is lost during operation between 10% and 50% of FULL cower. l I Above 50% FULL power, an automatic reactor trip will occur if either pume is tcst. l The power-to-flow ratio will be maintained equal to or less chan 1.0 which ensures that the minicum DNB ratio increases at lower flow since the max 1:um 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 allowed in Specification 15.3.1.D and the PCRV has met its most recent channel test as specified in Table 15.4.1-1. The PORVs operate to relieve, in a controlled manner, reactor coolant syste= pressure increases below
G. OPERATIONAL L:MITAT CNS The f ollowing CN3 related parameters shall be maintained within the limits shown during FULL cower oceration: u
- 1. TAVG shall be =aintained:
-< 578*? when RCS total flow > 178000 gym
< 576.9*F when RCS total flow < 178000 gpm
- 2. ' Reactor coolant system pressure shall be maintained:
) ,1955 psig during operaticn at 2000 psia
- 3. Reactor Coolant System Total Flow Rate >, (.95) x 178,000 Basis:
Although the operational limitations above require reactor coolant system total flow be maintained above a minimum rate, no direct means of measuring absolute flow during operation exist. However, during initial startup raactor coolant flow was measured and correlated to core AT. Therefore monitoring of 2T may be used to verify the above mini =um flow requirement is mat. If a change in steady state full power AT greater than 3*F is observed, the actual flow measure =ents will be taken. 15.3.1-19
The eight main steam safety valves have a total cambined 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-unlikel event of cosplete loss of electrical power to the station, decay heat removal wou'Id continue to be assured for each unit 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 I via the main steam safety valves or ats>apheric 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
I e 15.3.5 :::STR'.:ME:: ATIC:1 SYSTEM 0;eraticnal Safetv Instru=entation Aeplic ability :
~
Applies to plant instrumentation systems. Ob1ectives: To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety circuits necessary to ensure reactor safety. Soecification: 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 channel 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 the event tne 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 to the requirement shown in Tables 15.3.5-2 through 15.3.5-4, Operator Action when minimum operable channels unavailable. D. The accident monitoring instrumentation channels in Table 15.3.5-5 shall be operable. In the event the number of channels in a parti-cular sub-system falls below the minimum number of operable channels given in Column 2, operation and subsequent operator action shall be in accordance with Column 3. Bas:s: Instrumentation has been provided to sense accident cendtticas and to
~
initiate operation of the Engineered Safety Features (l) . 15.3.5-1
tau!E 15. J.5-2 (Cont'd) . 1 2 3 4 5 NO. OF HIN. HINIMUM PERMISSIBIE OPEHATOH ACTION NO.OF CtIANNELS OPERABLE DEGHEE OF BYPASS IF COND{TIONS OF Eo. FUtJCflot1Ab Uti1T CIIANNEIS TO CilANNEIS REDUNDANCY CONDITIONS COIDHN 3 ou 4 THIP CANNOT,BE Hlfr [L I .
- Turbine Trip 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 mismatch shutdown R3. Lo to Stedin Generator 3/ loop 2/ loop 2/ loop 1/ loop Maintain hot Water Level shutdown R4. Undervoltage 4 KV Bus 2/ bus 1/ bus 1/ bus --
Maintain hot (both buses) shutdown
- 15. Underfreiluency 4 KV Bus 2/ bus 1/ bus 1/ bus --
Maintain hot (both buses) shutdown NOTE: When block condition exists, malatain normal operation. FP . - FULL power Ilut Ag>[,l icabl e e One additional channel may be taken out of service for zero power physic.s testing. l l t Page 2 of 2
TABLE lb. 3. 5-5 INSTRUMENT OPERATING CONDITIONS FOR INDICATIONS 1 2 3 MINIMUM NO. OF OPERABLE OPERA'IOR ACTION IF CONDITIONS O. FUNCTIONAL UNIT CHANNELS GANNEL OF COLUMN 2 CAN'NOT BE MET
- 1. PORV Position Indicator 1/ Valve 1/ Valve If the operability of the PORV position indicator cannot be restored within 48 hours, shut the associated PORV Block Valve.
- 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 llock 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 Hot Shutdown within the next 12 hours.
- 3. 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., after a by On-Line Computer load change >10% of full power or after >30 inches of control motion.
0 Applies to presently installed combination of auxiliary feedwater pump discharge flow indicators and auxiliary feedwater flow to steam. generator indicators.
i 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 cransmission 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 KV station aux 111ery transformer is in service and the gas turbine is operating.
- c. Reactor power level is limited to 50: FULL power until 2 or more transmission lines are restored to service.
- d. 480 Volt buses 303 and B04 for the unit to be taken critical are -
energized.
- e. 4160 Volt buses A03, A04, A05, and A06 for the unit to be taken critical are energized.
l f. A fuel supply of 11,000 gallons is available; and both diesel I generators are operable. I
- g. Both batteries and DC systema are operable.
l 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 of systems and components:
- a. If the 345 KV lines are reduced to only one, any operating reactor (s) must be promptly reduced to, and limited to, 50: FULL co er. If alt 345 KV lines are lost, any operating reactor (s) will be reduced to supplying its auxiliary load, until one or more 345 K7 transmission t- lines are again available. %
- b. If both 345/13.8 KV auxiliary transfe=ers are out of service and only the gas turbine is operating, only one reactor vill re=ain operating and it will be limited to.504 FULL cower. The se :nt reactor will be placed in the het 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 transfor:ner
~ must 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 oparable and the associated diesel generator is operating and providing power *a the engineered safeguard bus nomally 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 safety features associated with this diesel generator shall be operable.
) f. One battery may be inoperable for a period not exceeding 24 hours 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.
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 utilired to provide depth and reliability of service to the Engineered Safeguards equipment through redundant station auxiliary power supply systems. 15.3.7-3 l l l
If only one 345KV transmission line is in service to the olant swit:hyarc, 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 orcer to maintain l continuity of service and the possibility of self-sustaining coerstions, if only one 345<V transmission line is :n service to any coeration reactor (s), the i cower level of the affecteo reactor (s) will be limited to 50% FULL , cower. l
\
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 auxiliaries. 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 cower. These conservative limits are set l to improve transmission system reliability only and are not dictated by safety system recuirements. References FSAR Section 8 l i i t 15.3.7-5
15.3.10 CONTROL RCD AND POWER DISTRIBUTION LIM "3 Acolicability Applies to the operation of the control rads and to core power distribution . limits. Objective To insure (1) core subcriticality after a reactor trip, (2) a limit en potential reactivity insertions frem a hypothetical rod cluster control assembly (RCCA) ejection, and (3) an acceptable core power distribution during power operation. Seecification 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.
- 2. When the reactor is critical, the control banks shall be inserted no further than the limits shown,by the lines on Figure 15.3.10-1.
Exceptions r.o the insertion limit are permitted for physics tests and control rod exercises.
- 3. The shutdown margin shall exceed the applicable value as snown in rigure 15.3.10-2 under all steady-state operating conditions from of 350'F to FULL power. An exceptien to the stuck RCCA cemponent I
the shutdown margin requirement is permitted for physics tests,
- 4. Except for physics tests a shutdown margin of at least :% 2k/k sha~1 l
I be maintained when the reactor coolant temperature is less than 350*T.
- 5. When the reactor is in the hot shutdown condit:.on or during any for physics tests, the critical approach to criticality, except red posit:.on shall not be icwer than the insertion limit for cere power. That is, if the control rods were withdrawn n ncr.a1 the reactor would nc:
I sequence with no other reacttvity change, the r.sertica 1:.: .it .
, be critical until the control banks were above N
- 3. 7evn: Siser but:.cn _=.:.:s
- 1. a. T.xcept d=:.ng ;cw pcwer phys:.cs tests, the he channel facecrs defined .n the bas:.s nus nee: :. e felicw:.ng L'- es:
i I I
~~ s e l
- g(e)1(s.4.) x K s, -d w - .or ? > .5 .
i
; RCo . :21 p
for y < 3 :..cwrate
?2 ( ) <4.64 x K (::) -
> 178000 FQ (Z) 1 (2.52) x K(z) for ? > .5 RC .ota.3 P
Flowrate
- Q(,<-) 15.04 x K(Z) for P 1 5 < 178000 L
T{gl. Sax {1+0.2 (1-P)} Where ? is the fractica of FULL pcwer at vnich the core is l cperating, K (*) is the functicn in Figure '.5.3.10-3 and : is the core height loca:Acn of F . Q
- b. Following a refueling shutdown prior Oc exceeding 90% of FULL power and at effective FULL power scnchiy inter rals thereafter, power distributien =aps using the moveable incere detec:c system shall be made to confl.rm that the het channel fac:cr l'- :s are satisfied. The seasured het channel facters shall be t increased in the fc11cwing way (1) The :naasurement of total peaking facecr, F-{**8, shall be increased by .hree percent to account for =anufacturing tolerances and fur-der increased by five percent :c account for measuremen: error.
(2) The seasure:mont of ent.halpy rise act channel fae:cr, T*l . shall be increased by four percent to account for measure-ment error.
- c. If a neasured het channel factor exceeds the FULL ;cwer ist.: l cf Specifica:icn 15.3.10.3.1.a, the reacecr ;cwer and ;cwer range hign se points snail be :sduced unt:.1 these ; - s are met. If sunsequent flux napp:.ng cannce, w:.th:.n ::4 hcurs, dercnstra:e ns:
- ne full power bc :hannel factor . -- 0 are met, :ne :vericwer
and everta=perature 10 trip se:pc = :s shall be s --larl/ reduced and react = ;cwer li=1:ed suen that specificati:n 15.2.10.3.1.a acove is met.
- 2. a. The target flux difference as defined in the basis shall be measured at least quarterly. A .arget flux difference update
- value shall be determined monthly by measurertent, er by 1= ear interpolation between the last =sasured value and 0% at end of cycle life (that is when the bore- concentratien in the ::clant is zero ppm) , or by extrapciation of ne last three measured poin-s.
The target flux difference and its associated alarm se points need not be updated if the update value for FULL power .arge: flux difference is within +0.5% of the presently empicyed FULL power target flux difference value,
- b. Except f=r physics testing, execre detec Or calibration (including
/
recovery), or as undified belew, the indicated axial flux diff erence shall be maintained within a range of +6 and -9 percent of the targe flux difference. Th s is defined as the carget band. i ! c. At a pcwer level greater than 90 percent of FULL power, if the t indicated axial flux difference deviates frem its targe: band, i I the flux difference shall be returned to the target band t tmmediately or reac cr pcwer shall be reduced := a level ne i t I greater than 90 percent of FULL power. .-
- d. At a power level ne greater than 90.percen: Of FUb1 pcwer, I L (1) The indicated axial flux difference nay deviate from ::s -s l
l f Oc -9% target band for a r.ax =um of one hcur (cu=cla :.ves ( in any 24 hour persed previded :ne flux d;fference does net exceed an en.reicpe =cunded my -11 percen: and '1 percent at 90% FULL cower and increasing by -1% an +1% for e3cn 2% :: i m
I i i 1
?~1L power below 90* . If the cumulative :i=e exceeds :ne I
/
hcur in any 24 hcur pericd, snen -J.e reac- : pcwer shall- be 17 .c no greater than 50g FULL power and the l reduced i=sedia high neutren flux se: point reduced to no greater =an 55%' j of FULL power. 1 (2) A pcwer increase to a level greater than 90% of FULL power is cen:ingen: upon the indicated axial flux difference being within its :arge: band. I
- c. At a power level no greater than 50 pei: cent of FULL power, (1)
The indicated axial flux difference =ay deviate f cm :.:s target band, f (2) : powerincrease to a level greater than 50% of FL1L power is cen:ingen upon the indicated axial flux difference nc: being cutside its carget band for more than two hcurs (cu=ulat:.ve) out of the preceding 24 hour period. One half of the ::.=e the indicated axial flux difference is cut of its targe: , band up to 50% of' FULL power is to be counted as contributing I to the ene hour cu=ulative maximum the flux difference ma*/ dev:. ate from its targe: band at a pcuer '_evel less : nan l or equal c 90t of FULL powe:.. f. Alar =s shall nor=al'.y be used to indicate ncn-cenformance w:. n l the flux difference requirement of LS.3.10.3.2.c or the f ux If :he 4.fference-time requirement of 15. 3.10.3.2. d(1) . the ax.a1 flux difference alarms are te.7,crar:.ly cut-of-serv:.ce, shall be noteds and cenfc:-.ance w:.th the la=ats assessed every hcur for :ne firs: 24 hours, and half-heurly thereafter. 15.3.10-4
- 1. Excep- :c: .: nys:.cs :
ests , e..cnover :ne :...c.:.ca:ac. quar.:7 .: .r-er
-alt exceeds 2% the til: cendi.icn shall be elire.nated w:.th:. . : c 4
hcurs er the felicwing ac nens shal.'. be taken:
- a. Reduce ccre power level and the pcwer range hagn flux se:po:.n:
- two percent of rated values fer every percent of indicated I quadrant power tilt. .
- b. If .he tilt is not corrected within 24 hours, but he het channel fac:crs for rated pcwer are not exceeded, an evaluacen as := the cause of the discrepancy shall be :nade and repersed := the Nuclear Regulatory Comunissicn. Return to FULL power is permitted, prev:. ding l the het channel factors are not exceeded.
- c. If the design hot channel facters for FULL power are exceeded or l 1
not determined within 24 hours, the Nuclear Regula cry Cc=sr.:.ssicn 1 i sh211 he notified and the overpower iT and everteeperature
- T tr:.; se:-
i points shall be reduced by the equ:. valent of 2) FULL power for every l 1 l percent of quadrant power tilt. i I d. The execre nuclear instrumentation system serves as the pri:na.ry quadrant power tilt alar:n. If the alazu is not fune icnal for tvc I 2 hours, backup methods of assuring that the quadrant power til: :.s 4 acceptable shall be used. These :nethods include hand calcula:1cns, incere theruccouples using either a cen:puter er nanual calcula:: ens f or incere detectors,
- e.
- hen ene pcwer range channel is incperable and therr,a1 power is v
1 ( greater than 75% of PJLL power, be quadrant power :il: l ' shall be confirmed as acceptable by use of the ::cvarle :.ncore t ! detec crs at least once per 12 hours. i f C. Ince_erable Red Cluster Centrol Assez.civ. T E C.N.
- 1. An RCCA shall be considered incperarle if cne er -cre cf he foll::e:.ng l
l I cccurs : l l
'.5.2.10-5
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 oank and cannot be aligned when the bank is between 215 steps and 30 steps, then unless the hot channel factors are shown to be within design limits as specified in Section 15.3.10.8-1 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 determine the hot channel factors and the resulting allowable power level based on Section 15.3.10.B. 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.
- 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:
- a. For operation between 10% power and FULL power, the position of the RCCA(s) with the failed rod position indicator channel (s) will be checked indirectly by core instrumentation (excore detectors, and/or thermocouples, and/or moveable incore detectors) every shif t and af ter associated bank motion exceeding 24 steps in one direction.
- b. For operation belcw 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
anomalies which would, otherdise, affect these bases. . Axial Pcwer Distribution The procedures for axial power distribution control are designed to minimize the effects of xenon redistribution on the axial pcwer distribution durin,g 1 cad follow maneuvers. hsica$1y,centroloffluxdifferenceisrequiredto11=1: 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 j hxial offset = AI/ fractional power) . The FULL power target flux difference is defined as that indicated flux l difference of the core in the following conditions equilibrium xenon 0.ittle er no oscillation) and with the in11-length red centrol rod bank more than 190 steps d..e. , the nor=al full power position) . Values for all other core power withdrawn 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 . s value was noted, no allowances for excore detector error are necessar/ and _ indicated deviation of +6 and -9 percent AI are permitted frem the indicated reference value. During periods where extensive load following is required, ! it may be impractical to establish the required core conditions for =easuring i the target flux difference every month. For this reason, the specification provides three =sthods for updating the target flux dif:.erence. S trict centrol of the flux difference und rod position) ir. %?. as necessar/ during reduced power operatien. This is because xenen Astributien centrol at reduced power is not as significant as the centrol at FULL power and allowance l has been nade in predicting the heat flux peaking facters for less strict control at reduced power. S trict control of the flux difference is not possible f l l during certain physics tests or during required pericdic execre calibrations I Therefore, the specifi-which require larger flux differences than permitted. 15..'.10-12
cations en power distribution centrol are net applied during physics tests or excere calibrations. This is &cceptable due to the increased core monitoring performed'as part of the tests and low pechability of a significant accident occurring during these operations. In some, instances of rapid plant power reduction, automatic rod rotion will cause the flux difference to deviate from the target band when the reduced pcwer level is reached. This does not necessarily affect the xenen d:.stributien suf ficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band. However, to simplify the specification for operation up to 90s of FULL power, a lir.itation of one hour in any period of 24 hours is placed cnoperation 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 protected frem .J overpower and minimum DNBR of 1.30 by an automatic protection system. Ccmpliance with operating procedures is assumed as a pre-conditien s hcwever, operator errer and equipment malfunctions are separately assumed to lead to the cause of the transients considered. Quadrant Tilt The execre detectors are somewhat insensitive to disturbances near the core center such as misaligned inner control rods. I: is therefore possible that a five percent tilt might actually be present in the core when the execre detectors respond with a two percent indicated cuadrant ._ tilt. On the other hand, they are overly responsive to disturbances near the - per.phery. 15.3.10-13
.. . ., .. l
. in a daliberato manner witnout uncus prossure en uno epsra ng persennel because
, of the unusual techniques to be used to accommodate the reactivity changes -
associated with the shutdown. 4
.v.isaliened RCOAS
- The various control red banks (shutdown banks and control banks, A, B, C, .and D) are each to be moved as a banks that is, with all rods in the bank yithin ene step (5/8 inch) of the bank position. Direct information on rod 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 rod position. The rod position indicator channel has a demen-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 hot channel factors to be exceeded. A single fully misaligned RCOA, that is, an RCCA 12 feet out of alignment with its bank, does not result in exceeding core limits in steady-state operation at power levels less than or equal to rated power. In other words, a single dropped RCOA is allowable from a core power distribution
_ viewpoint. If the misalignment condition cannot be readily corrected, the specified reduction in power to 75% of FULL power will insure that design margins to core limits will be maintained under both steady-state and anticipated tran-sient conditions. The eight (8) hour permissible limit on rod misalignment at rated oower is short with respect to the probability of an independent accident. Because the rod positica indicator system may have a 7.5 inch error when a misalignment of 15 inches is occurring, the Specification allows only a 7.5 inch indicated misalignment. However, when the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the consequences of a misalignment are much less severe. The differential worth of an individual RCCA is less, and the resultant purturbation en power dist:D::utions is less than when the bank is in its high differential worth region. At the top and bottem of the core, an indicated 15 inch misalignment =ay be representing an actual misalign=ent of 22.5 inches. The failure of an .NDT in itself dcis not reduce the shutdown capability of the 15.3.10-15
- ds, but it dcas reduce .he operator's capacility f:r dece===ing 2.e ; css: en Of e.at : d b'/ di' rect =eans. *he opera =r has available to hr= the execre dotect:r rec =rd=gs, incers ther=cccuple readings and periodic incere flux
. races f== indirectly deter =ining ::d position and flux tilts should the =d w:.th the inoperacle UI::T beceme salposi: ened. "'he exc=re and incers' inst:.:=enta-tien w 11 not necessar:.ly recognize a sisalignment of 15 inches because the
~
- encersnatant increase in pcwer density will nor= ally be less 2.an it for a 15 inch
=1salign= ant. "he execre and incere inst:.:=entation will, however, detect any :=d -4< align =ent which is sufficient :o cause a significant increase in hde channel fac crs and/c any significant loss in shutdcwn capability. The increased surveil-lance of the core if one or more red positien indicator channels is cut-of-ser-r:.ca serves to guard against any significan: loss in shutdown marg n er margin to core c.ermal li=its.
The history of malpositioned RCCA's indicates that in nearly all such cases, the
=alpesati ning occurred during bank move =ent. Checking :=d pcsition after bank
=ction exceeds 24 steps will verify that the RC:".A with the inoperable 2/:T is w
mov:.ng prcperly with its bank and the bank step counter. ..alpositzening of an RCCA in a stati na:/ hank is very rare, and if it does occur, it is usually gross Should it go undetected, the slippage wnich will be seen by external detec.= s. came between the red positi n enecks performed every shift is shcr. with respect to the pr bability of =ce.1 rence of another independent undetected situatien which would fur her reduce e.e shutdcwn capability of the reds. Any :=mcination of misaligned rods below 10% FULL power will not exceed the Fer this reason, it is not necessary es -hecx the positica =f design l'm ts. reds with incpsrable r/ T's below los power: plus, O.e incere inst:t=entacten is not ef f ective for deter =ining :=d posa cn until the pcwer level :.s aseve approx =stely 54. 15.3.10-16
. a--..
15.3.11 MovAB:. CI-CORE OtS..s s ;;A;;;N Applicabilitv Applies to the operability of the mevable detector instrumentation systam. cbiective: To specify functional requirements on the use of the in-core instrumentation systems for the recalibration of the excore axial off-set detecticn system. Soecification: A. A = h4= = of 2 thimbles per quadrant and sufficient movable in-core detectors shall be operable during re-calibration of the excore axial off-set detection systam. B. Power shall be limited to 90% of FULL power if the calibration l requirements for escore axial off-set detection system, identified in Table 15.4.1-1, are not met. Basis: The Movable In-Core Instrumentation System has four drives, four detectors, cad 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 detec*ars. ' To calibrate the excere 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 *wp and bottem halves of the core. , m Unit 1 Amendment 15.3.11-1
0
- 6 I
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ATTACHMENT A SAFETY EVALUATION FOR REDUCED THERMAL DESIGN FLOW STUDY POINT BEACH NUCLEAR PLANT UNIT 1
?
I. INTRODUCTION AND PURPOSE This safety evaluation has been perfomed to address the non-LOCA safety l considerations in allowing Point Beach Unit No.1 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 reduced. Thus to maintain guaranteed stean flow, Tayg must be increased or stean pressure reduced. Primary reactor coolant mass inventory is reduced. - The impact of higher stean 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 t..e important parameters for each accident, detemine which of these param-eters wre affected by the higher stean generator tube plugging levels, and then detemine how the impacted paraneters affected the accident ana ly si s. The resulting impacts wre detemined 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 wre consistent with the following assunptions:
Maximun core thermal powr, MWt 1381.8 Themal design flow, gpm/ loop 84,500 S.G. tube plugging level, percent 24 ; ayg at 100 percent of maximum allowd powr, *F T 572.86 ,
~
AT at 100 percent of maximun allowd powr, *F 55.5 l l RCS pressure, psia 2000 AP 1.58 I no lo ad , F 547 ) F0 maximum 2.52
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 pe '1 ties due to reduced primary flow. Reduction in system power is a benefit in DNB calculations and more than offsets the flow and T,yg (relative to reduced power) penalties. The most recently applicable analysis used in this report is indicated by the reference number after each title. Uncontrolled RCCA Withdrawal From a Subcritical Conditio'nIII 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 temper-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 in 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 flo'.i g reduction. Therefore, the nuclear transient is only moderataly :ensi-tive to the inpact of steam generator tube pluggirg. es
The most recent analysis (1) 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 terrperature 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 temperatures wo_uld be less than ~784*F and ~617 F, respec-tively, for the present evaluation. These values are still significantly below fuel 'ielt (4S00*F) and zirconium-H2 O reaction (1800*F) limits, and the impact of increased steam generator tube plugging, up to T.4 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 powed, resulting in an increase in reactor coolant ternperature (Section 14.1.2 of FSAR). Reduced flows result in less margin to DNB. Reduced therrr.al 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 constar.ts to be determined for the overtemperature AT set point equation. Thus to assure adequate core protection the Reactor Core Thermal and Hydraulic Safety Limits have been recalculated consistent with the reduction 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 liinits, 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. Method of Analysis The transient was reanalyzed employing the sacc digital computer code and assumptions'regarding initial conditions and instrumentation and setpont errors used for the FSAR, including: :
- 1. Power levels equal to 102 percent, '62 percent, and 12 percent of .
1381.8 MWT; i-
- 2. Average temperature 4*F above T corresponding to the initial 3g
. power level;
- 3. Pressure (1970 psia) 30 psi below nominal;
- 4. Reactor trip on high nuclear flux at 118 percent of 1381.8 MWT with t
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. A 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 a function of reactivity insertion rate for 102 percent, 62 percent and 12 percent of full power. 4 Conclusions These results demonstrate that the conclusions presented in the FSAR are still valid. That is, the core and reactor coolant system are not adversely affected since nuclear flux and overtemperature aT trips pre-vent the minimum DNB ratio from falling below 1.30 for this incident. Thus the current setpoint equation and reduction in rated power compen-sate for the reduction in thermal design flow. : a +
Ma1 positioning 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 (3) 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 Ond 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 overshoot. 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-surizers) is reduced from 5253 3ft to 4930 ft3 . However, the minimum dilution ;
time has been 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
~
1130 ppm at refuelini and 1600 ppm at startup. The minimum time required for the reactor to beccme 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 sufficient 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 y 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 ai trip setpoint resulting in a reactor. trip. Af ter reactor trip there is at least 15.1 minutes for operator 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 trans','.1t is conservatively estimated to be 1.15 pcm/sec and is within i the range of insertion rates analyzed for uncontrolled rod withdrawal at power. Prior to reaching the Overtemperature aT reactor trip the opera-tor will have received an alarm on Overtemperature aT and turbine run-i back. l i With the reactor in automatic control, a boron dilution will result in a power and temperature increase such that the rod controller will attempt i to compensate by slow insertion of the control rods. This action by the controller will result in rod insertion limit and axial flux alarms, f The minimum time to lose the 1 percent ak/k shutdown margin rec auired at beginning-of-life would be greater than 15.1 minutes. The time would be 1 s
a significantly longer at end-of-life due to the low initial ' boron concen-tration and 2.77 percent ak/k shutdown margin. . Rup;ure of a Steam Pipe (2) 2 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: i Hypothetical Break (steam pipe rupture) Inside Contaimrent with and without power Outside Contair. ment with and without power
- Credible Break (Safety valve opening)
A. steamline break results in a rapid dapressurization 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 D!lB in small regions of the core (-5 percent) would not violate NRC acceptance criteria. l The impact of ir. creased levels of steam generator tube plugging would affect the accident principally due to the reduced flow, reduced RCS inventory, and reduced heat transfer ccefficient. These impacts would result in changed cooldown and feedback reactivity characteristics such that the return to power as shown in the previous analysis would be slightly conservative with respect to the lower initial flow condi-i 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 results conservative with respect to the current analysis. i .
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 rodded core at end of life.
- 3) The rod having the most reactivity stuck in its fully withdrawn position.
- 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 I presents the core parameters for the 4 hypothetical break cases used in DNB 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 DN3 design basis for the hypothetical breaks and remains subcritical for the credi-ble breaks for the 24 percent tube plugging.
1 Startuo of an Inactive Reactor Coolant Looo I2I 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-
~
f 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 14.1.5 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 si'ghtly lower reactivity insertion rate, resulting'in a lower peak heat flux. Therefore results presented in the FSAR would be conserva-ti ve. Reduction in Feedwater Enthalpy Incident (2)
' The addition of excessive feedwater and inadvertent opening of the feed-water bypass valve are excessive heat removal incidents which result in 4
a power increase due to moderator feedback. Increased levels of steam
~
generator tube plugging would impact this analysis principally due to the reduced flow. 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 coefficient. 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 avg. Tne reduction in flow is more 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.
?
The second case assumes a large negative moderator coef ficient. The
' reduction in thermal design flow will result in a slower ccoldown, and therefore the reactivity insertic, rate will be less than in the FSAR
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 'i in steady state DNBR. Protection for this accidenyis 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-EOL) 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 teriperature 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 verifed 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 i l l 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 DNB ratio and same time to trip, a loss of flow event from these conditions 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 redoction in power. The resulting DNB 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 i.utan-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 also results in reduced initial hot spot values. This would offset the slight increase in fuel and clad temperatures effect of reduced flow. Consequently, the expected peak fuel and clad temperatures would remain the same as results of the currently applicable analysis. It is estimated that the peak pressure will not increase above the pre-vious value due 'to reduced power, however the maximum calculated value was 2778 psia based on 2250 psia operation plus 30 psia uncertainty. This is significantly below the pressure at wnich vessel stress limits
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 primaar 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 will not cause safety limits to be exceeded for a locked roter accident. Loss of External Electrical Load (2P 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 nventory. Two cases, analyzed for both beginning and end of life con-
- ditions, are presented in Section 14.1.9 of the FSAR:
- a. Reactor in automatic 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 pressurizer 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 T effect will be minor, however, since the reactor is tripped on high pressurizer pressure. Thus, the time to trip will be decreased which ,
'~.
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.
. l Loss of Nomal Feedwater/ Station Blackout (2)
This transient is analyzed to deterinine 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 rapid heatup of the primary system. The resulting coolant density change would increase the volume of water in the pressurizer. The analyses in FSAR l Section 14.1.10 and 14.1.11 show that the peak pressurizer volume 3 reached is 780 ft 3on 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 pert:ent due to flow reductions, this would result in a maxi-mum increase of less than 125 ft3 in liquid volume. This is still below the 1000 ft3 capacity of the pressurizer, thus no reanalysis is . necessary. In addition, due to the relatively long duration of the transient following trip, the results are highly sensitive to residual (decay) heat generation. Residual heat generation is directly propor-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 . 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 terminated by a reactor trip activated by high nuclear : power signals.
O The rod ejection transient is analyzed at full power and hot zero power (
^
for bcth 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. Inis impact would result in a reduction in heat transfer to the coolant which would increase clad and fuel peak temperatures. 1
- Reanalysis was performed using the conservative ejected rod worths and
! post-ejection peaking factors assumed in the latest analyses which are above the calculated Point Beach reload values. Reanalysis was per-formed to show that the increase in initial Fg from 2.47 in the pre-- vious analysis to 2.52 is still acceptable. . Method 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 i calculation of the transient is performed in two stages, first an. aver-l age core calculation and then a hot region calculation. The average core is analyzed to determine the average power generation .with time
- including the various total core feedback effects, -i.e. Doppler reactiv-ity and moderator density reactivity. Enthalpy-and temperature tran-sients in the hot . spot are determined by adding a 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 DH3 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 'I. The condi- ; tions at the hot sput fuel rod do not exceed the limiting fuel cri-3 te ria I4) . The conclusions of the FSAR, therefore are still valid. i w m.. -.-- ,-- . - - , - - - - , .
.v---+.
l 4 l 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-line 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 higher operating pressures were addressed also. 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. E
~
REFERENCES-
-1. Davidson, S. .
L., Editor, " Reload Safety Evaluation Point Bea'h Nuclear Plant Unit .1, Cycle 9A," December 1980.
- 2. Final Safety Analysis Report Point Beach Nuclear Plant, UnitIs 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, Revision 1-A, January,1975.
s + l 1 j.- I'
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,a..- - - - - -
4
' T AB LE I CORE PARAMETERS USED IN STEM LINE BREAK DN8 ANALYSIS Outside Break With Power Outside Break Without Power Inside Break With Power Inside Break Without Power Time (Sec.) 72.4 75.2 78.4 81.6 104.8 108.0 111.2 114.0 58.4 61.2 64. 8 67.6 94.4 97.6 102.4 107.2 Core inlet 4
Temperature ('F) Loop A 403 401 399 398 361 357 355 352 356 355 353 351 295 292 288 285 Loop 8 454 451 448 446 489 489 488 487 436 434 4 30 427 50 9 50 8 50 7 50 7 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 j nominal) Heat Flus 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 Hwt) RCS pressure 695 694 691 689 1039 1045 1053 1061 656 656 655 653 1008 1020 IG27 1030 (pstal b o
+
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TABLE 11 SLNMARY OF ROD' EJECTION ANALYSIS PAR.4ETERS AND RESULTS BOL. BOL EOL E0L Power Level, percent 0 10 2 0 -102
. Ejected rod worth, percent ao 0.91 0.34 0.95 0.42 Delayed neutron fraction, percent- 0.0049 0.0049 0.0046 0.0046
.F g before rod ejection 2.52 -
2.52 - 11.2 5.03 13.7 5.10
, Fg after rod ejection Number of operating pumps 1 2 1 2 r
. Maximum fuel pellet center - 3504 4543 3719 4458 ' temperature, *F Maximum fuel pellet averag'e 3095 3492 3289 3392 i. temperature, *r Maximum clad average temperature, 'F 2440 2129 2550 2071~ f Fuel Pellet Melting, percent 0.0 0.0 0.0 0.0 i Maximum fuel enthalpy (btu /lb) 231.4 266.7 248.7. 257.8 i I
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- g. -
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<s *SO. 00 - -
Ta c 400 00 - - w - 350.00 4 - i 300.00 2000.O N G 1750 0 s e . s -
~ 1500.0 -
s s E ! \
$ 1250 0 +I w
= i
" !C00.00 t
- 2 e.
= 750.00 4 1 600 00 '
3.0000 e : 2 -
- e 2 !000 *
= w I w .
s. 35 2.0000
- a. 2 i
#3 r o 1.5000 \N ,
a -* a w
- : 1.0000 1
< l.
T* a
- 50000 ' '
w 0_. 0 40000 ; 7w
- 0.0~ .
=
* \\ / a
- 50000 9 '
s
". l -
N 1 S -1 C000 kl / <
! \ ',
i 4
+
2' -1.5000 -
/ \
3 i N' 4 w
-2.0000 -
/
y -2 5000 - s
-3. M00 m a 2 = o g o
o e
= a
~
Q
~
o = .~. m p E %
, c a ~ - - &-
- -t 5E:
.r .; a*Pr L2 Lib ".
STEAMLI:1E I?1 SIDE TriE C0tiTAIllMElli (AT EXIT OF STEAM GE?tERATOR) . OUTSIDE POWER AVAILABLE sc: :c
!!0.00 %,s -
w 500.00 + \ -
- - 450 00 &
57 -
+
? E 400.00 + 7 5i o 350.00 e
- r
+
* ~
200.00 + - 250.00 &' + 200 00 __ _ ___ 2000.0 2 1750.0 ' 7 s %
- 1500.0 7 3
1250.0 7
\ l w
f 1000.CO - I C 750.00 -
+
l 500.00 g 3.0000 ! c .+ z
" T:s 2.5000 ; I 5 w$
$= 2.C000 -
m , I I a
# 3 1.5000 + $
x o M *
* ! 0C00 -
I j
'00C0 - ~
53
- O. 0 _.
4CC00 l /%
= 3.0 ;
~
- 50000 A\ ;
l I i a I w l \ . 5 -1 0000 Ll i 'N
-1. 5CC0 Nf N +
0 p
- i N~
-s -2 CCCO +il :
s s
< l .
g -2 SCC 0 +f .
-3.0CC0 - - e = =
=
o y g = = = -
- ~ - , =
p 2 e ~ P N
- 1 $
a ,, 2
- ME '5E;;
;. ruLoo..cw s c
a STEAMLIllE 3REAK OUTSIDE THE C0tlTAlitMENT (DOWNSTREAM OF THE FLOW MEASURIllG .'10ZZLE) LOSS OF OUTSIDE POWER AT T = 0 sco 0; 575.CO + w
$50.00 % '
s - [-
= 3 525.CO + '\. ~
$ 5 500.00 + \
E$ *15.00 1 450 00 it - 425.CO + I y
*00 00 200G v -.
5 r 1900.0 t \s , 1600.Oh '
=z T, N 1400 0 +
E 1
- a. i ,
- 200.0 + ,
+ +
1000 00 +
?00.CO ' 4 3.0000 I ~
E -
*o 2.5000 --
I w + E, E 2.0000 \ '
- a. =
#3 1.5000 ;
5 7 E' !.0000 +
-2 +
4 i
.- E* 50000 - I w
- 0. 0 t 40000 ,
- 0. 0 I
' fw
'I- jl ,N O
=
w
- 5C000 t, / \ ~
t -1.C000 t / s i 0 -1.5000 & / *
/
C-2.0000q, 7 .
/
/
- y -2.5000 +i !
- ib
-3.0000 -
8 o 8 o 5
=
5 5 5 o = p = 9 O o = o q p n = =N A
==
FIGUE 6
STEAMLIllE 3REAK IflSIDE THE C0ilTAINMEtiT - (AT EXIT OF STEAM GEliERATOR) LOSS OF OUTSIDE POWER AT T = 0 500.aa I - 550.00 I 3 + 2 g 500.00 y . h E 450.00 t. !
$ k_ 400.00 r w f 350.00 + 5 300.00 2000.0 2 1800.0 -
E
- 1600 0 +
- '\ 5 a l
I I 3 1400.0 - 4
- i f 1200.0 -
4 y1000.00 - x h 800.00 ~ 3.0000 i o z a
*E 2.5000 - F
" w ;
7 5 2.0000 - +
- a. 2 Yr 1.5000 -
IS l E* 1.0000 - y
*5 ,
E= o + 50000 + - 4 w
- 0. 0
~
40000 - af / 0.0 . s .
/'
2,
- 50000 d' N-
- l a
~ ,
N
- i t-10000p! 'N
\
T E
-t 5000 h ! l
-2 0000 el t
!\l' +
g -2 5000 'j'
-3.0000 = a ;
,, e c o = = a a 0 =_ =_
o C 3 C 5 S U _ rise iSE.: FIGUPE 7
STEAM 3REAK EQUIVALEtti TO 0?!E - STEAM GEi!EPATOR SAFETY VALVE 500 00 . 575 00 - -
- !!O 00 - I
#- t s
= 525 30 - .
IC
< is 500 CO - -
WI 4 75. 00 + . 450 00 425 00 1 . I 400 00 0.0 '1
- -1.0000 \. ~
w \ i -2.0000 - \
- t
, -3.0000 *
/ 4' O l .
3 -4 J000 + * {-5.0000+l *
-4.0000 2000.0 -s
- I '
5 e 1750. 0 +
- A i
[ 1500. 0 + '
- 3 i~
i .
, 1250.0 ? s 1 l N !
,100000J t
~w 30.00
-v e o = g g o O O ~= ~
o o 2 9 E E
. e a a 7 UPE <!EO
.=
FIGURE 3}}