ML18227B256
| ML18227B256 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 07/20/1978 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18227B256 (94) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RID DISTRIBUTION FOR INCOMING MATERIAL 0-2 251 REC: STELLO V ORG: UHRIG R E DOCDATE: 07/20/78 NRC FL PWR Zc LIGHT DATE RCVD: 07/28/7 DOCTYPE: LETTER NOTARIZED: NO COPIES RECEIVED
SUBJECT:
LTR 3 ENCL 40 FORWARDING SUPPLEMENT TO APPLICANT"S 07/10/78 PROPOSED AMEND TO LIC NOS DPR-31 8( 41 RE FULL PWR OPERATION WITH UP TO 25/ OF STEAM GENERATOR TUB4ES PLUGGED CONSISTING OF "NON-'LOCA ACCIDENTS AFETY EVALUATION FOR HIGHER LEVELS OF STEAM GENERATOR TUBE PLUG&l<4 PLANT NAME: TURKEY PT 03 REVIEWER INITIAL: X JM TURKEY PT 84 DISTRIBUTER INITIAL:~
DISTRIBUTION OF Tl-IIS MATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTER ISSUANCE GF OPERATING LICENSE.
(DISTRIBUTION CODE A001)
FOR ACTION: BR CHIEF ORBSi BC++W/7 ENCL INTERNAL: +%W/ENCL NRC PDR~~W/ENCL I 5 E4+W/2 ENCL OELD44LTR ONLY HANAUER++W/ENCL CORE PERFORMANCE BR>>W/ENCL AD FOR SYS 5 PROJ44W/ENCL ENGINEERING BR+4W/ENCL REACTOR SAFETY BR4~W/ENCL PLANT SYSTEMS BR>>W/ENCL EEB++W/ENCL EFFLUENT TREAT SYS<<W/ENCL J. MCGOUGH+4W/ENCL EXTERNAL: LPDR S MIAMI, FL~~hl/ENCL TERA~+W/ENCL NSIC~~W/ENCL ACRS CAT BN+W/16 ENCL DISTRIBUTION: LTR 40 ENCL 39 CONTROL NBR'82090005 SIZE: 3P+43P
I wi e P,
u
'I 0
'I Y
~ m
FLORIDA POWER & LIGHT COMPANY July 20, 1978 L-78-242 S
Office of Nuclear Reactor Regulation Is Attention: Nr. Victor Stello, Director 3s Division of Operating Reactors Nuclear Regulatory Commission II U. S. em I Washington, D. C. 20555 ~ LSg
Dear Nr. Stello:
m CSSS I+
Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Supplement to Proposed Amendment to Facility 0 eratin Licenses DPR-31 and DPR-41 Florida Power 8 Light Company (FPL) letter L-78-230 of July 10, 1978, con-tained proposed amendments to the Turkey Point Units 3 and 4 Operating Licenses to allow full power operation wi th up to 25%%d of the steam generator tubes plugged. The proposed amendments resulted from an evaluation of non-LOCA transients. An analysis of the LOCA transients, which could lead to further proposed amendments, is scheduled to be submitted by August 10, 1978.
On July 12, after receipt of the FPL July 10 submittal, the NRC staff initiated a telephone conference call to discuss the FPL safety evaluation submi tted in support of the proposed amendments . During the phone call, FPL representatives and the NRC staff scheduled a meeting for the purpose of continuing the discussion. The meeting was held on July 13, 1978. As a result of the meeting, it was agreed that FPL would submit a supplement to the July 10 letter containing a correspondi ng evaluation developed by the NSSS vendor. The NSSS vendor analysis is attached. The difference between it and the FPL analysis is as follows:
a) Section 3.2 - Evaluation The NSSS vendor did not consider the effects of fuel rod bow in the evaluation and recorIIIended that the generic rod bow penalty on F~H be applied. Where appropriate, FPL modified the evalua-tion such that i t included the effects of a rod bow penalty, consistent wi th a previous submittal (FPL letter L-77-106 of April 4, 1977) and showed that the rod bow penalty could be ab-sorbed without a reduction in F~H. Any other differences are editorial in nature .
b) Section 3.3.1 Uncontrolled Control Rod Assembl Withdrawal at Power The NSSS vendor reanalysis did not include the development of new transient curves. Instead the NSSS vendor examined the setpoints and concluded that there was substantial margin in the current toms
>820500000 PEOPLE... SERVING PEOPLE
Qlk T
1 f ~
Office of:Nuclear Reactor Regulation Page Three Please call if you. have further questions regarding this or. previous s ubmi tta1 s.
Very truly yours, Robert E. Uhrig Vice President REU/HA'S/RDH/cpc Attachment cc: 'Hr. James P. 0.'Reilly, Region II Robert Lowenstein, Esquire
.' l 4> f \
T .Jl
Office of Nuclear Reactor Regulation
,Page Two
,(FSAR)'vertemperature aT setpoint and'hat the Overpower aT setpoint shoul'd be changed. 'FPL: performed a reanalysis of .the transients wi-th the setpoints developed by the NSSS vendor with the rod bow penalty included'nd showed that the DNBR l.imits were not,violated. The Overpower aT and'vertemperature aT setpoints resulting. from the two .reanal'yses .are therefore identical.
c) Section 3.3.2 .L'oss of Reactor Cool:ant Flow. 'Flow Coast-Down ccsdent The 'NSSS vendor reanalysis did not consider, the effects of. fuel, rod bow. FPL modi.fied. the reanalysi's to incl.ude a rod 'bow .penalty.
Al.though, the vendor reanalysis did not consider,rod'ow, the minimum
.DNBR obtained by. the vendor during the transient is 1.48. This
- value shows, that the effect of rod 'bowing can be accommodated without penal'i'zi,ng any operating .parameter,.
d) Section 4.0 - Technical S ecifications The only difference between the Technical Speci'ficati'ons proposed by the NSSS vendor and FPL are the Reactor Core Thermal and Hydraulic Safety, Limits curves, Fi'gure,2..1-lb. In: generating these curves the, vendor did: not incl'ude the effects of rod bow penal.ty., but FPL
'id. The:FPL curves were developed'sing; the COBRA:code exclusively.,
.i.n; conformance wi'th: the methods used'o. generate the companion, curves, submitted in FPL 1'etter L-78-217, on"June 22,
- 1978, for lesser amounts of steam generator tube, plugging. As the rod bow,penalty is already incl,uded in these curves, there is no need for
, any further penalty,to the F<H limi'.t or to any other operating parameter.
e) Section .6.0' References
'FPL,requests,a .timely review of their Safety and Fuel Management
'A'nalysi:s Methods ('eference 6.')'nd of the:DYNODE-P Code (; Reference 7)',
as these methods and codes, are to 'be used'n future analyses to be submitted.:by FPL in support of licensing, actions.
A, question was,raised at the July 13 meeti.ng,. concerni.ng the .NSSS vendor's use of a modified:'FLARE code on the Uni't 4, Cycle 5 design:(refer to Unit 4, Cycle '5 Reload Safety Evaluation, ,'FPL 1'etter L-78-210, dated June 19, 1978).
We have discussed the use of this code with 'NSSS vendor representatives, and they have verifi ed that all uses of the FLARE model for the cycle 5 design were:,'backed'p (confi.rmed) wi.th the use of .thei'r, standard methods.
4i NON-LOCA ACCIDENTS SAFETY EVALUATION FOR HIGHER LEVELS OF STEA11 GENERATOR TUBE PLUGGING JUNE 1978
-.,WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTE'<1S DIVISION P.O. BOX 355 PITTSBURG11, PENNSYLVANIA 15230
0 11 ig) 1'
P 'a
- 1. 0 IllTRODUCTXON The current Turkey Point 3 and 4 safety analyses are valid for steam generator tube plugging levels of up to 19% on Units 3 and ~i.
Above these levels, the current LOCA-KCCS accident analysis as well as some of the current non-LOCA accident analyses are not valid because:
- 1. The RCS flow rate is reduced to below the thermal design value assumed in the current LOCA-ECCS and applicable non-LOCA analyses,
- 2. the adverse impact oi higher steam generator plugging levels on the blowdown and" ref lood pnases of LOCA-ECCS analysis has not been anal-yoked,
<he reduction in RCS volume (from the p'bagged steam generator tubes) can have an impact on some of the current non-LOCA analyses and must now be explicitly considered, and 1'a. the pump coast-down characteristics are more severe chan those assumed in tne current loss of flow analysis.
Xt is the purpose of this report to present an evaluation of the applic-able non-LOCA accidents, considering the above factors, to demonstrate that with appropriate Technical Specifications changes, Turkey Point 3 and 4 can be operated safety from a non-LO"A accident standpoint with up to 25% of the \
steam generator tubes plugged. A LOCA-ECCS accident re-analysis for higher than 19% steam generator tube- plugging and which considers the'bove factors is currently being performed, but is not included. Xt will, how ver, be submitted as a supplement to this report in the near future.
The evaluation provided in this report was conducted as follows:
4~ 0 4i Determine thc RCS flow rate associated with 25% stcam generator tube pluggx,ng.
- 2. Evaluate the impact of this tube plugging and the associated RCS flow rate on those significant parameters which influence the results of the applicable non-LOCA accident analyses.
3.,Reanalyze those non-LOCA. accidents which are either most limiting or most sensitive to the impacts resulting from 25% tube plugging level.
Th remainder of this report is organized as follows:
Conservative flow rates versus level of steam generator tube plugging are developed in Section 2. The applicable non-LOCA accident evalua-(ion and reanalyses are provided in Section 3. The required changes to the Technical Specifications are summarized in Section 4. Conclusions, are given in Section 5. References are provided in Section 6.
I+i l 1 C
I 2.0 FLOH RATE V'RSUS LEVEL OF STEAN GEJ'ERATOR TUl'E PLUGGIJlC Flow measurements have been taken at the Turkey Point Power Station for several levels of steam generator tube plugging. These data were than compared to the flow rates obtained from the analytical model used to calculate the best-estimate flow rate. Deviations between tJIe model prediction and the measuremcnt data points were conservatively accounted for by subtracting a constant bias (equal to the largest deviation between the measurement data and the design prediction) from the model prediction curve of flow'rate versus steam generato" tube plugging level. This measurement bias corrected curve was then further reduced by a factor of 1.02 to account for measurement instrumentation uncer-tainty (see Table 1).
The resulting curve of flow rate versus level of steam generator tube plugging is provided in Figure l. This curve indicates that a tube plugging level of 25% will conservatively result in a flow rate of no more than 5% below the thermal design flow rate of 89,500 gpm per loop.
This value, 85,025 gpm per loop, was then used, along with the tube plugging level of 25%, as the basis for the non-LOCA accident evaluation.
15 j I, i
1
~ 'L
~
3.0 ACCIDEtlT ANALYSIS 3.1 INTRODUCTXOH 0
The impact of higher steam generator. tube plugging levels of up to 258 on the non-LOCA accident analyses presented in Chapter 14 of thc PSAR has
.,"-been-assessed. The basic approach used'as to identify the important
- parameters for each accident, determine which -of these parameters, were affected by the higher steam generator tube plugging levels, and: then.
determined how the impacted paramet'ers affected the accident analysis.
The resulting impacts were determined by either evaluating the accident to qualitatively demonstrate that the accident is not limiting or re-
~
*,'anal'yz'ing the aJffected accident (if the accident was 'found to be limit-
'ing or very sensit'ive to th impact of higher steam generator tube plug-ging levels). The evaluations were consistent with the following assumptions:
Used'n Used in the Currently Thermal, design flow, gprn/loop 85, 025 89, 500 S. G. tube plugging, 25 19 Maximum allowed power, tfwt 2200 2200 T oF 574.2 574.2 avg at 100/ power, 5T at 100% power, F 58.9 55.9 FN l. 55 1.'55 (1.,75 PSAR)
J>>
- .,'FO'"maximum' ' ' '
2.05 2.55 (Eon-LOCA)'
In per'eral, reanalysis and.evaluation technique" w'ere based'n the Exceptions to thxs policy
~
assumptions and methods employed 'in .the
~ ~
PSAR.
J
-:aie noted in the text.
l l
~ I S
I The evaluation o f the non-LO"A accidents is presented in Sec tion 3.2.
The results 'of this evaluation indica tel that the following accidents werc limiting (i.e., the results of the accident analysis are close to safety limits) or most sensitive to the impact of the highe steam gen-erator tube plugging levels:
Uncontrolled RCCA 4'ithdrawal at Power Loss of Reac tor Coolant Flow (Flow Coast-Down "Accident)
Chemical and Volume Control System Malfunction The results of the reanalyses of these accidents are provided in Section 3.3.
3' EYhI.UATXON The following accidents were evaluated and found to have sufficient margin to the accident safety limits.
- l. Uncontrolled RCCA h'ithdrawal From a Subcrx.txcal Condx.txon (1)
~ ~
A control rod assembly withdrawal incident when the reactor is sub-critical 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 reactivity feedback of the negative fuel temperature coeffi-cient. The-power excursion causes a heatup oi .the moderator. How-ever, since the power r:ise is rapid and is followed by an immediate reactor trip., the moderatox:,temperature rise is small. Thus, nucl'ear power-response is .primarily a function of the Doppler tem-perature 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 tran fer coefficient which in turn results
II
<5 I
~ I in a faster fuel temperature increase than reported in the PSAR analysis. (1) The faster tempera ture increase would result in more l
I Doppler feedback thus reducing the nuclear power heat flux excux-sion, as presented in Reference 1, which would partially compensate for the flow reduction. Therefore, ti-.. nuclear transient is only modera te ly sensitive to the impac t o; . "..earn genera tor tube plugging.
>>e PSAR analys'is(1) shows'hat for a 60 x 10 5 Ak/sec reac-tivity insertion rate, the peak heat flu.. achieved is 46.8% of nominal. This is conservative for the higher plugging situation for
.the reasons stated above. The resultant peak fuel average tempera-ture was 951 P. A 5% reduction in flow and the associated reduc-tion in core heat transfer coefficient would degrade heat txansfer from the fuel by a maximum 5% and increase the rise in peak fuel and clad temperatures by a maximum of 5%. Therefore, the fuel and clad temperatures would be less than ~1000 P and ~630 F, respec-tively, for the present evaluation. Th se values are still signifi-cantly below fuel melt (5080 F) and zirconium-}12 0 reaction (1800 P) limits, and the impact of increased steam generator tube plugging, up to 25%, would not result in a violation of safety limits.
A malpositioning of a paxt length rod accident need not be addressed due to Technical Specification restrictions which prohibit power operation with the part length rods in the core.
.3.'- Rod Cluster Control As..emblv (RCCA) Dro .
The drop of a Control Rod Assembly xesults in a step decrease in reactivity with produces a similar. reduction in core power, thus reducing the coolant average temperature. The highly negative mod-erator temperature coefficient (-35 pcm/ P) assumed in thc analy-sis results in a power increase (overshoot) above the turbine power
Ii runback value causing a temporary imbalance between core power and secondary power extraction capability. Tais analysis is potenti lly sensitive to steam generator tube plugging due to the reduced flow.
The effect of a 5% reduction in initial RCS flow would be a sma1ler reduction in coolant av rage temperature. Thus the power overshoot would be less than the value shown in section 14.1.4 of thc PSAR.
on the PSAR transient, statepoints were evaluated consistent ',;Based with a 5% reduction in flow. The results of this DNB evaluation showed that the DllBR limit of 1.30 can be accommodated with margin.
"~refore, the impact of increased steam gener'ator tube plugging on r ..::control Rod Assembly Drop Accident analysis would not appre-ciably affect the margin to the safety limits.
- 4. Startu of an Inactive Reactor Coolant Lop~
An inadvertent startup of an idle reactor coolant pump with loop stop valves open results in the injection oi cold water into the core. This accident need not be addres'sed due to Technical Specifi-cations restrictions which prohibit power operation with a loop out of service. 1Iowever, evaluation shows, that the re ults presented in the PSAR would be conservative for any impacts associated with increased levels of steam generator. tube plugging.
- 5. Reduction in Peedwater Enthal Incident
.The. addition of excessive 'feedwater and inadvertent opening of the
'feedwater bypass valve are excessive heat removal incidents which
-resu1't 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.
Section 14.1.7 of the PSAR presents two cases. The first case assumes a zero moderator coefficient, which is used to demonstrate inherent transient attenuation capability during a feedwater reduc-tion. A reduction in flow will have a negligible effect on
i, 0 stability since thc 4
reactivity insertion is identical to the FSAR case duc to the zero moderator temperature coefficient. Dpi's not n considcra tion for this. case since DNDR's do not fall below thc steady state value- This is due to thc relatively large reduction in TGvg . Thc reduction in flow~ however, will result in the ini-tial steady state DNHR being reduced from 1.8 to ~3..72 (this value corresponds to the DNDR which was calculated at time zero in the Loss of Flow reanalysis using THXHC). Thus, adequate margin to steady limits is retained.
h The second c'ase assumes a large negative moderator coefficient-impact of increased steam generator tube plugging (reduction in fl'ow) will result in a slower cooldown and, therefore, a lower reac-tivity insextion rate than'in t'e ISAR analysis. The in egral xeac-tivity insertion due to moderator temperature reduction will be less than t: he FSAR case, thus producing a lower peak nuclear power.
Thexcfore, the reduction in DNHR from the steady state value
(~1 72 for increased st:cam tube plugging levels) would bc no great:er than t:hat shown in the FSAR. The FSAR shows a DNM xeduc-tion of ~0.04. Tints, the 5% flow reduction will result in a mini-mum DNBR of ~1.6~, and considexable margin to safety limits-Evaluation has snown t:hat sufficient margin is available to the safety limits for the Feedwatcx'yst: em Ifalfuncticn Accident for increased levels of stcam generator tube plugging. Xn addition, further protection is assured for both cases via the identified margin in the Overtemperature AT protection sctpoints (sce Section 3.3. 1).
g 6 . Excessive Load Xncrease Xncident
'P S . ~
'.hn excessive load increase event, in which the stcam load exceeds thc core power, results in a decrease in reactor coolant system I
temperature, which is very similar to thc fcedwatcr malfunction
lf 0,
~ ~
1 C
I lp analysis. As in the Fcedwatcr lIalfunction Accident, reduced flow is I the pr indi pal impact on this accident due to increased level,s, of I steam generator tube plugging.
Two Excessive Load Increase cases are presented in FSAR Section
'14.1. 8.
=" =10% Step '1'oa'd inc'r'ease, iith-rrsnual control.
Step load increase with automatic control.
'10%
The worst case results (automatic control) indicate that with no trip actuation, 'steady- state conditions are reached,,with a minimum DHBR of >).30.'Hargin is available in that the F>H used the FSAR analysis 'was 1.75 as compared to the current value of 1 55.
~ This results in a DM3 benefit of ~30%. The impact of in-creased steam generator tube plugging on the Excessive Load Zncrease Accident is an overall reduction in D1~33R of approximately 5%. Thus there is considerable margin avail'able given all of the assumptions usersdoes result in in the original anal'ysis remain valid. Tube plugging small changes. to the initial conditions, however, these changes tend to be in the conservative direction.
Another .area of margin avai:lable to offset the flow reduction penalty is in the Overtemperature hT trip t"hich i for pro-tection in the case with automati'c control. lIargin ha been identi-fi:ed in the Overtempera'ture hT tr'ip prot~..tion setpoints (see Section 3.3..1)*., As stated in the FSAR,;the.. ad'equacy o', this protec-tion was veri'fi:ed. in, the 'Rod 'Mithdrawal't'ower Accident,(see FSAR
-'- '.Sec t i on "14 ..1."2) ." =,.'
- 7. L'oss of Reactor<Coolant Flow '(Locke<1 Rotor Accident) (2)
The FSAR (Section 14.1.9) shows that th- .most severe Loci;ed Rotor Accident 'is an instantaneous seizure of a reactor cool'ant pump rotor
I 0
at 100% power with three loops operating. Following the incident, reac tor coolant system temperature rises until shortly after reactor tr ip.
Thc 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 DTB is conservatively
- assumed to occur at the beginning of the transient. The flow coast-do~m in the affected loop due to the Locked Rotor is so rapid that the time of reactor trip (low flow sctpoint is reached) is essen-tially identical to that presented in Reference 2. Therefore, the
- nuclear power and heat flux responses will be the same as shown.
However, the reduction in flow-would result 'in slightly higher cal-culated system pressures and fuel and clad temperatures.
Tne current'y applicable analysis(2) shows a calculated peak fuel temperature of 2940 F and a peak clad temperature of 1500 F.
Peak temperatures are highly sensitive to the initial hot spot" values assumed in the analysis. The above analysis was based on a hot spot heat transfer calculation which employed heat flux and fuel temperatures based on an Fq of 2.55. LOCA considerations require that an F~ limit of <2.05 (value assumes 'axial stack height and spike penalties) be used., This results in a 20%,reduction in total energy input to the hot spot w'nich will more than compensate for the 5% reduction in flow. Consequently, the expected peak fuel and clad temperatures would remain below the results of the currently appli-cable analysis ~
Xt is estimated that the peak system pressure will increase ~80 "psia above the previous value, however, the maximum calculated value was 2720 psia. This is significantly below the pressure at which vessel stress limits are exceeded (a400 psia xists to this limit), thus, considerable margin exists to absorb any slight pres-sure increase.
~
(It should be noted that the 25% reduction in the
0 h I number of steam generator tube would re ult in approximately a 10%
reduction in primary coolant mass which would decrease tlie Ident capaciL'y, of the RCG by the sam amount. This would noL result in .
higher peak temper" tures or pressures, however, since the peak values are reached in considerable less than one loop transport time constant.)
4
- .------Therefore, operation .at reduced 'flow will not cause -safety limits to be exceeded for a Locke'd Rotor hccident.
- 8. l.oss of Fxternal Electrical Load~
l.'he result of a loss of load is a core power level wnich 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 Lhe decreased RCS mass inventory. Two cases, analyzed fo" both beginning and end of life conditions, are presented in Section 14.1.10 of the FSAR:
- a. Reactor in automatic rod control with operation of the pressur-izer 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.
t
.Tlie FSAR analys'is<>~ results in a peak pressurizer pressure of '
.2517;psia following reactor trip and a minimum D'i~BR of 1.61., A reduction 'in 1'oop flow and RCS ma s inventory will re ult in a. more
~ - - --'rapid, pressure rise than is currently shown. The effect will be
>~W
'inor, however, since the reactor is tripped on high pressurizer pressure. Thus, the time to trip will be decrea ed which will result in a lower toL'al energy input to the coolant. Therefore,
IP, 0' although the initial margin to DNA will be reduced, the min mum transient DN13R will be only slightly affected and tne margin to the safety limit" will be maintained. In addition, the identified margin in the Overtcmperature AT setpoints (see Section 3 3-1) will assure adequate margin to DNG.
9~ Loss of Normal Feedw ter<<J This transient is analyzed to determine that the peak RCS pressure does not exceed allowable limits and that the coxe remains covered wi th water. These critex ia are assured by applying the more strin-requirement that the pressurizer must not be filled with water.
Tncreased steam generator tube plugging levels would impact the accident principally clue to reduced flow. The effect of these impacts 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 analysis results presented in Section 14.1.1) of the FSAR show that considerable margin is available. This analysis shows that the peale pressurizer, volume reached is 1100 ft3 on an approximate 300 ft change in volume. This result was due to a 26oP change in coolant average temperature..-Using thc highly conservative assump-tion, that the average temperature d lta would increase by 50% due to flow reductions, this would result in.a maxim'ncrease of less
'han'150 ft3 in liquid volume.,This is still below the 1300 ft capacity of the, pressurizer thus no reanalysis is necessary.
Therefore, the results of. this reevaluation indicate that the impact of increased levels of steam generator tube plugging will allow sufficient margin to be maintained to the safety limit associated with the Loss of Normal leedwater Accident.
l0. Loss of AC Power (I)
This transient is analyzed to show that upon loss of all AC po" r to the unit auxiliaries, the auxiliary feedwater system is sufficient to remove stored and residual heat without water release th ough pressuri.zed relief valves.
As in a Loss of Normal Feedwater Accident, increased steam generator-tube plugging would impact the Loss of AC Power transient primarily due to reduced flow.
Upon the loss of power to the reactor coolant pumps, coolant flow neces- sary for core cooling and t¹ removal of residual heat is maintained by natural circulation in the reactor coolant loops. In the FSAR analysis, the natural circulation flow was calculated using an analytical method based on the condition of equilibrium flow and maximum loop resistance. The reduction in natural circulation flow rat'es due to the 5% reduction in thermal design flow is negligible.
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 gener..tion is directly proportional to initial power level preceding.,the trip. The accident assumed the power to be at 102% of the maximum turbine rating 2300 Hwt. Thus, total energy input to the system would be w4.3% less than H
- ,assumed in the FSAR This re"evaluation indicates that the impact of increased levels of steam generator tube ptug. zng will not adversely-impact thc Loss of
,AC Power Accident. Sufficient margin to the safety limits are maxntalned.
ll. Ru ture of a Steam Pi e(
The steamline brealc transient is ana)yzed for hot zero power, end of life conditions (Section 14.2.5 of the FSAR) for the following case.",:
0 lt
-Hypothetical Brcak (steam pipe rupture)
, Inside Containnent with and without power Outside Containment with and without power Credible Break (Dump valve opening)
A steamline break results in a rapid depressuriration of .the steam generators;wnich.causs a large reactivity insertion to the core via
~ ~
primary cooldown. The cceptance criteria for this accident is that no DHB must occur following a return to power.'his limit, however, is highly conservative since steam line break is classified as a Condition III event. As such, the occurrence of DNB in small regi'ons of the core (~5%) would not violate 1lRC acceptance cri.
'eria.
.The impact of increased level's 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 that the return to power as shown in, the pre-vious analysis( ) would be slightly conservative with respect to the lower initial flow conditions. In addition, the time of Safety Injection actuation would be unaffected by flow condition" for the Hypothetical Breaks. Tnis 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 t&e current- analy is.
V 1
- IV
-..-,.Thus.'.the impact'of increased levels of steam generator- tube plugging
wi;ll not resul't in a viol'ation- of westinghouse or IiPC safety limits.
II' I
l2. 1tu ture of a Control Rod Drive Hechanism <<lou.,in RCCA E ection(2)
The rupture of a control rod drive mechanism housing which allowed a control rod assembly to be rapidly ejected from the core would result in a core thermal 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 P.
activated by high nuclear, power signals.
The rod ejection transient is analyzed at full'ower and hot standby 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 peale temperatures. The current analy-sis ( ) results and inputs are summarized in Table IX. As is shown, all cases have significant margin to fuel failure limits ~
The effect of reducing flow by 5% is to primarily increase the peale clad temperatures by ~50 1'. The current analysis shows that for.
all cases a value of at least 400 F can be accommodated before peak clad limits are reached (2/00 F). The fuel temperatures will also increase, however, they will increase much less than the clad increase due to the rapid nature of the transient.
In addition, there is a significant degree of conservatism in the inpuLs. The ejected rod worths'nd post ejection peaking factors a'e'5% abov the calculated Turkey Point reload values. Also for the fuell power cases, tlirt initial hot spot fuel temperatures were
'-:. calculated assuming an 1'+ of 2 55. Due. to 1,0CA considerations, q l the F limit ll will be b
<2 05 This rcsul ts in more than a 175 reduction in .initial fuel temperaL'ure which translates into a
>75 F reduction in peak transient fuel, temperature which vill compensate for th reduction in thermal design flow.
0 f I C
Therefore, the impact of increased levels of steam generator tube
,plugging on the Rod Bjection Accident will not signi.ficantly reduce the margin to the safety limit due to the conservative inputs and large margin to the limits.
3.3 REANALYSIS
" The-.fol'lowing"accident:s-ilare-'reanal'yzed"bccau e-.they vere either'limit=
- << - . E ing or were sensitive to the impact of. increased'team generator tube.
plugging.
'l." Uncontrol'led Control Rod Assembl ithdrawal at Power Z
- * . E E
'An uncontrolled control rod assembly withdraval at power produces a mismatch in steam flow and core power,'esulting in an increase in reac tor. coolant temperature. Increased steam genera tor tube plug-ging will impact t'ne analysis principally due to the 'influence of the reduced flow, the elevated outlet temperature, and the increased
.loop transit time. The first two impacts will result in less ini-tial margin to DI.B. The third impact requires new values of lead/lag time constants to be determined for the Over temperature hT setpoint equation.
The Uncontrolled Rod Uithdrawal at Power Accident analysis utilized the overtemperature hT trip for protection during "slow" reac-
'tivity-insertion'vents,and the high neut:on flux tri'p for "fas't"
- *reactiv'ity"insertion rates. FoEr"the fast or 'high insertion rates,
'the;.DIKRs',,reported in the 'PSAR'oul'd 'be reduced by ~5/ due to the
-. -fl'ow'-reducti'on.>. -'The':DHBR',s.'reported,.in 'the PSAR were ana'yzec.
'E EE assuming an Fgq'.of- 1.75; .however, the curient FDII limit
", l'.55." .This, results in greater than 20% addition DiIBR" margin. 'hus the flow deduct'ion penalty is morc than offset by P BII benefit for the fast reactivity insertions ~
1 I
J
C
. For the. slow reactivity insert:ion rates, thc overtempcraturc AT trip function is used. This trip is based on the Rcactox COx'c 1
~
Thermal and )Iydraulic Limits. To assure adequate coxe protection~
the Reactor Core Thermal and IIJdraulic Safety I.imits have been recalculated consistent with the reduction in RCS flow. )3asrd on these ncw protection lines, thc Overtempcraturc and Overpower AT setpoint equations constants have been recalculated consistent with
=
the new Core Limits and,the methods outlined in Reference 5.
I Table XXX give" a comparison of the oxiginal. (CESAR) and the xecalcu-lated constants. For the overtemperature equation, the FSAR values arc significantly more limiting than the xccalc'ulated values. Thzs is directly due to thc PAII reduction from 1-75 t:0 1-55 in recal-culating the Reactor. Core Thermal anhd Itydraulrc Safety Laments. 'lo offset the effects of an RCS flow reduction, the I'SAR ovcrtempera-tuxc AT trip constants will bc maintained in the Tec)~nxcal'1 Specz-ficat:ions (see Sect:ion. ~i'0)-
For t:he most limiting case shown in t'e I'SAR {ISAR I'igure 1~i-l.2-6),
the MIIR i rctluccd from a st:eady -tate value of 1.62 down to '6.30; Assuming a xcartor coolant flow of 85025 gpm/loop and a l-55I lAH l~~lt, ~
e Ute steady state MB)< is >1.7. By using t:he ovcrtcmperatuxe setpoints derive<I in the 1'SAR,tern. approximately thc same reduction in MIIR would occur f01 t: he limiting reactivity insertion rat:e. Thus a mipimum DLILR of
~1.5 is anticipated. Thexcfore, considerable uargin ia available t:o
~
..the MI3 limit to offset minox system changes due to a flow reduction.
.,As anticipated, Table XXX show tltat the overpower AT setpoint leave
- = - become. more limiting wi'th the- flow reduction. (scc Reiexence 5 for a
. 'iscussion
\
of this functi'on); - This will require a Technical Specifi
-.,- cation change as shown in Section 4'.0. However, no additional. analysis
$s xequircd since the'verpower AT t:xip function is not explicitly usctl in any cf t:he 1'SAR transients analyzed. Rathex':his trip function is con idexcd a divcxsc mean of trip for thc power Range Nuclear Xns tx'u men t: a tion Sy" l
Il The above evaluation shows that the conclusions presented in the FSAR are stUl valid. That is, the core and reactor coolant: system are not adversely affected since the nuclear flux and overtemperature AT trips prevent the minimum DiiB ratio from falling below 1.30 for thii incident.
- 2. Loss of Reactor Coolant Flow (Flow Coast-Down Accident)
~ (2)
As demonstrated in the FSAR, Section 14.1.9 the most severe loss of flow tran ient is caused by the simultaneous loss of electrical power to all three reactor coolant pumps. This transient was 1.eanalyzed to determine the effect of steam generator tube plugging on the minimum DNBR reached during the incident. Tube plugging will result in a significant decrease in margin to safety limits due to the following effects:
-High loop resistances result in a more rapid flow coast-down.
k
-Lower initial flows result in less margin )o the 1.30 DHBR 1l!Bits.~ .- '-
Thus reanalysis is required.
0 I II
Analysis meth'ods -and assumptions used in the reanalysis were con-sistent with those employed in the FSAR. The e assumptions l includc: ~
I Lnitial operating conditions most adverse with respect 'to the margin to DNBR, i.e., maximum s teady state power level (102% o f nominal), minimum pressure (2220 psia), and maximum core average temperature (578. 2 F);
- b. Maximum.Doppler power and temperature coefficient and most posi-tive moderator temperature coefficients;.
c~ Time from loss of power to all pumps to the initiati'on of con-trol rod -assembly motion (reactor trip) of 1.6 seconds; and
- d. 4% hk trip reactivity from full power using frozen feedback.~
- e. Loop resistances equivalent to 25% tube plugging.,
The fl'ow coastdown was calculated by the PIIOENIX code, and the resultant system transient was simulated using the LOFTRAII code. The calculation was performed for a zero constant moderator coefficient value.
The minimum value of DNBR presented in the FSAR for this incident was 1.31. The DNB evaluation performed in the reanalysis was based on the evaluation model described in 4'CAP-S075(4). Figures 2 9
-.-'- ~,- -, through 6 show the flow coastdown, nuclear power, heat flux, and
-.".;=;minimum DhB*ratio vs. time..Figure 6 clearly, .shows that the effect
-...,,-.of rod',*bowing can be. accommodated'itli margin in this analysi's.
I Steam generator tube plugging appreciably affects the re ults of the
'-='omplete loss of flow tran'sient; however, the minimum DHBR .remains he reactivity rate versus control rod insertion following a reactor scram used in this analysis is slightly 1c ss (i.e., nore conservative than that u ed in the previously applicable "nalys<.s. (Plea.e note that ihe conclusion of the previously applicable anal> ises 'are, not affected by the use of thc new reactivity rate insertion curve).
~i . ~ ,
above 1.30 for this incident. The three pump case was reanalyzed since it is the most limiting one presented in the FSAR. Loss of a single pump with a ll loops in service or with a single loop out of service w re less limiting.
- 3. 'hemical and Volume Control S stem Halfunction Section 14.1.5 of the Turkey Point Units 3 6 4 FSAR, shows that for a boron dilution event the operator has sufficient time to identify the problem and terminate the dilution before the reactor returns critical or loses shutdo~m capability. The standard acceptance criteria and FSAR calculated values for, operator action are summar-ized below:
F SAR Acceptance Hode (minu te s ) Cri teria Re fueling > 120 30 Star tun > 240 Power
- a. Hanual Control > 15 15
- b. hu toma t ic Control 21 Steam generator tube plugging has no affect on the analysis at refueli'ng conditions since only the re'actor vessel volume is assumed act'ive. The coolant loop volume are conservatively assumed stagnant.
For dilution during startup" and at power, there is an effect due to the reduction in primary c'oolant volume. The effective volume of primary coolant in the steam oenerator tubes is assumed to be reduced by 25% (a 510 ft ). Thus the total volume assumed in ft the analysis has been reduced from 7S00 . 3 to 7290 .3 . This ft translates into approximately a 7% reduc" ion'n thc originally cal-culated dilution-time from startup conditions (240 minutes). The result is still significantly greater than the required operator action time, therefore no safety concern exist.
4i I
~
~ ~
For dilutions during power operation a high)y con ervative reac-tivity insertion rate of 1.1 x 10-5 6k/scc was -as" umed in the FSAR consistent with an initial boron concentration of 1200 ppm.
FSAR Figure 14.1.5-1 (Reactivity Xnscrtion Rate vs Boron Concentra-tion) has been recalculated consistent with thc lower primary value and is shown in F igure 7
. ~ The results show that the reactivity
'insertion -rate. a'ssumed in the,FSAR is still'- conservative. Therefore.
-'no additional analysis. is required.,lt,should be. noted,, however, h
that tl>e FSAR analysis 'is stil.l highly conservative with respect to the current cycles since the analysis assumed that only 1% shutdown margin is available. The Turkey Point Units have been designed such that > 2.5% shutdown margin is always available for- BOL. condi.
tions. The resul't is that operator action times would'e > 70 minuteswith more a realistic, value.
Thus for all cases it has been shown tnat steam generator tube plug-ging up to 25% will not affect the safety con'clusions of the FSAR for boron dilution.
ldll C
0 1t 0
4.0 TFCH~AJ.CAL PECIFICATXOilS This section contains the technical content of proposed changes to the Technical Specifications. These changes are consistent with the plant operation necessary for the design and afety evaluation conclusions stated previously to remain valid. lnis report i" contingent upon approval of'hese changes.
Delete the following three pages of the technical specifications:
TS 2.3-2a
- 2. XS 3.l-7a
- 3'. TS 2.3-3 Add the appropriately mar1:ed pages which follow.
~,
lieactor Coolant Tem ieratur'e Overtempera-
"turc pT < ATo I1 1 0 ~ 0107 ('l 574) + 0-000453 (p-2235) f(~q)j
'P'ndicated'T Average.temperature, at rated
.F power, F
- Pressurizer'.pressure, psig '
function of the indicated difference be tween top and bo t tom de tec tors o f the power-range nuclear ion chambers; with gains to be .,elected based on measured instrument.
response during startup tests such that:
A
- << - qb) within +l0,percent and -14 Percen<<>>ere qt and qb are the percent power in the top and bottom halves oi;the core respectively, and q +
total core power in percent of rated power, For each. percent that the magnitude of (q qb .
exceeds +10 percent, the D lta-T trip setpoint shall be automatically reduced by 3.5 percent of its value at interim power.
For each percent that the magnitude of ('qt d 'r
,,q>).;exceeds -14,percent, the Delta-T
.trip setpoint shall .be automatica'1'ly reduced by 2 percent of its value't interim power.
(Three Loop Operation) = 1.095<
(Two Loop Operation) = 0.88 1 ~ 095 for 'steam generator tube plugging < 25 percent
0 0 (i')
D Over- d'J' po~;cr AT < hT 1.13 "- '1 dt: K2 I- " (l 0 I hT 'L>>dicatc<l hT at rat:c:d. po<rer,, F T = Avc:rage L'c mpera t>> re, I' 1>>diect <<d average tempera i:urc at: >>or>i>>al co>><lition., an<1 rat:ed -potrer, F 0 for <lecrcasi>>g average temperature, D
)P v
0.'2 se<:../P for",fncreasing average temperature
=-
li2 0.00068 for. T c.qual to or. more than T';
0 fox T less t:han T'1T Rate of change of temperat:ure P/sec dt f(hn) .= As .defined above Pressurizer Low 1 ressurizcr pre sure equal to or greater than 1835 psig.
lligh Pre: surizer pressure cc,ual to or less tlt'an 2385 psig. 0 lligh Pressurizer ~'ater level equal to or. less than 92% of full scale.
3 Reactor Coola>>t F3.'ot; Lo~r react:or coolant: flo:r equal to or greater. than 90% Qf'ormal indicated floi<.
Low reactor coola>>t pump motor frequency equal to or great:er tha>>. >6D 1 Hz I
\ D
-Under voltage on eacLor .Cool!>>L pump Elotor bus equal
't:o .Or,great'c.r.tha>> -60% of >>ormal volta< e I
D
~ D D
SLeam Geiici ~to'rs r
LOM-lour t:eam generator t..ater level equal to or great:er tha>> 5% of narro: range. instrument scale
~jIDsDI, I D 1..13 for. st:ca>i gc>>c'rcttor tube D ~
fact:01. xs plug;;'rng < 15 percent
~
this fact:or is 1.10 for r
steam gener>>Lor tutic pltif t,xn> l&,icxc~>>L and <
This factor is 1.08 fox steam generator plugging < 25 percent I'ube TS 2. 3-3
0 ll
PEAC'IOR COR~IIERllAL Af!D IIYDRAULIC SAFETY I.
" TS, THREE LOOP OPERATIOII
$ ,:,'ll
-j -','00 :2 psia
' ':- i.::l
'j.;
, I.::. lK 'II I 1 650 , ~ I;,h I :~ .-'I I -,I, . h 640' l'I'5 I ' 2250 la ~ h ~
t .'..I"
~ I ~ ~
- .I *"
L'~::: .,
I": I ~ +
~ I j -:-:,.-. h:.::.:I 630 J
j ':.:.I': l ~ . ~ ~
~ ~
h
'-;:--.l .'
r ',.: . I ht " ' ~ "
I '
620 I 'I "., I
~
I
.'"" ~
I I
"2000 psia l I
l -..: \ ~
610 I I",,'
I
-- "I'.I
~
I~
I LL" i -H
'I
~
609 I I I
. I.= .'=". I. '.'-":
I h
~
- I .::7 :
f~i 1
I 590'
,--i
- !j.i:. 1?75 sia
~ ~ . ~
.I'-.'
~ ~
~
.580 I~
- ,'I<OTE; I-These curves are appli'cable with I
<Ream genetawoj. tube pluggllig --
'5-percenC.
570=' ' I h I
~ '
I
[
I 560 20 40 60 80 100 120 140 RATED POIIER (PERCENT)
TS 2.l-lb
i 5.0 COÃCLUSXOMS analytical and empirical study- of the Turkey 1'oint iUnits 3
'ased on an and 4 flow characteristics, a 25% steam generator plug'- ing level was conservatively determined to result in a flow reduction 5/ below thermal design flow*
The 'impact of this higher st:earn generator tube plugging level on the now-I.OCA accident analyses. applicable to Turkey Point Unit" 3 and 4 has been assessed'he reevaluation has indicated that the impacts asso-ciated with increased steam generator tube plugging can be accommodate with margin to the safety limits for toe following accident analyses:
Uncont:rolled RCCS Withdrawal from a Subcritical Condition Halpositioning of a Part Length Rod Rod Cluster Control As-embly (RCCA) Drop
- tart-up of an Inactive Reactor Coolant Loop Iteduclion in Feedwater Enthalpy Incident Excessive Load Increase Incident Loss of Reactor Coolant Flow (Locked. Rotor. Accident)
Loss of External Elect:r.ical Load Loss of. Informal Feedwater Loss of A.C. Power Rupture of a Steam Pipe Rupture of. a Control Rod Hechanism Itousing Several accidents were determined to be either limiti'ng or most sensi-tive to the impact of higher .steam generator tube plu ging levels. .The following accidents were reanalyzed or required partial reanalysis:
Uncontrolled Control Rod Withdrawal at Power Loss of Reactor Coolant Flow (Flow Coast-down Accident)
Chemical and Volume Control System Halfunction
0 4 6; Dnp PARAIIETERS The fo 1'lowing DilB re la ted parameters limits I sha I
11 be maintained during power operation:
a- Reactor Coolant System Tavg < 578'.2 )'.
Prcssur'izer Pressure > 2220 psia""
- c. Reactor. Coolant Plow > 268,500 gpm+
With any of thc above parameter@ exceeding its limit, restore t'e param ter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5% of
'rated thermal poser using. normal shutdown,procedur'es.
Compliance with a. and b. is d monstrated by verify-ing teat each of the parameters is within -its limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Compliance with c. is demonstrated by verify'ing that the parameter is within- its limits after each refuel-ing cycle.
M 0
Limit n'ot app'1'icable during. e'ither a TIIERlSL..POHER ramp .increase in
--excess. of-(5%), RA'TED TII R'iIAL" POWER per m'inute or a THE14IAL POi<ER step increase in. excess of =(10%) -RATED TIIFR':IAL POl;ER.
5 "e.
+ Reactor Coolant Plow'<,268,500gpm.- for stcam, generator tube,plug ing
< 15%.
Reactor Coolant Plow > 263,1'30 gpm for steam generator tube plugging 15% and < 19%.
Reactor.'Coolant Plow > 255,075 gpm for steam generator tube plugging 19% and < 25%.
~
TS 3.1-7a
iS 0
Based on these reanalyses> it wa., determine 1 th..t the results meet the appropriate safety limits with the conservative Technical Specifications changes to the Reactor Core Thermal and Hydraulic Safety Linits-I Therefore, it is concluded that the impact of increased steam generator pluggin levels, up to 25%, can be accommodated without adversely
'-.'.'affecting'the safe op'ation of Turkey Point Units 3 and ~i.
~ ~
~ I ~
- 6. O'G;I EREIICHS I
- 1. "Final Safety Analysis Report, Turkey Point Plant, Units 3,and 4,"
Docket Nos. 50-250, 50-251.
- 2. "Reload 'Safety Evaluation> Turkey Point Plant, Unit 4, Cycle 4,"
January 197?-
a
- 3. "Reload Safety Evaluation> Turkey Point Plant, Unit 4, Cycle 5 j" June 1978.
- 4. "Fuel Densification, Turkey Point Plant Unit Ho. 3," MCAP-8074 (Proprietary) and HCAP-8075 (Non-'Proprietary), February 1973.
O'. "Design Basis for the Theraial Overpower AT and Thermal Over-temperature AT Trip Functions," HCAI'-8745 (Proprietary) and 4CAP-8746 (Non Proprietary), Harch 1 9/7.
- 6. "Calculation of Flow Coastdotm After Loss of Reactor CoolanL Pump (PIIOEIIIK Code)," HCAP-7551, August 1970.
- 7. LOFTRAH Code Description, 4'CAP-7878, IIarch 1972.
C
(Qi ly ly
~ ~
~ ~
~
c
~ 4 )
TABLE 1 REACTOR COOLAHT FLO';l M:ASUI'EllEHT UHCERTAIklTY Parameter Uncertainty RC Loo Flow Unccrtaint Feedvater Flow +1.25% +1.25%
Peedwater Temp. +1'.OoF +0.2%
=
Steam Pressure . +30 psi +0.1%
Hot
~
,0 5oF +0.9%
Cold 0 5OF +0.9%
RCS Pressure +50 psi
+0.2%'1.
8% Z2 V
h
.<Qi I ~
TABLE IX l
l I
EOL EOL Power Level, 102 0 102 0 Hjec ted rod work, /kk .71 .30 .84 Delayed neutron fraction, .50 .50 44 .44 Feedback reactivity weigliting 1. 20 1.30 1.2 2.6 Trip rod shutdown, %6k ejection 2.55 2 ~ 55 q before rod Pq after rod ejection 5.~i8 7.0 5. 52 14.3 Number of operating pumps 3 3 2 initial fuel temperature, oP 2660 '547 2475, 547 maximum fuel pellet average temperature, oF 2040 3080 Haximum clad average temperature, 3490'580 oui 2370 2125 2270 Fuel Pellet Helting, 0 0 0 -0
i>
e
~ ~
C
~ ~
TABLE IXI hT Protection Setpoints Overtcmperature hT Technical Speciiication Section 2 SetPo'int <ATo LK1 A (T-574) + B(P-2235) f (Aq)j FSAR Analysxs Current Limit Recalculated Values Values (19% S.G. Plu~nin )
.Kl 1. 095 1;. 08 1.18
. 0107 . 0107 . 00959
.000453 .000453 . 000420 Overpower AT Technical Specifica'tion Section 2 Setpoin < ATo t'Ko-Kl K2 (T T ) l(g,q)j t.
1.10 1.08 Kl 0.2 0.2 0.2
i'
)~
FIGURE 1
~. )~ )
TURi EY POI i' UIHTS 3 AflD 4 RCS FLOlf VS STEAN GEi"fERATOR TUBE PLUGGIifG 1.00000 I ~
' .I
<< ~ ~ )
~ ~ . I f ~
i
- << ~ f
~ ~
~ f I
~: ~ ~ ~
I
~ f ~ ~ I ~ ~
~
I<< ~ ) ~
~ ~ ~ ~ ) ~ ~
~ I ~~ ~- ~ ~ ~ ~
~ ~ I
~ ~
i; .'.'.
90000
I' '~
. ~ I '
<< 89500 I '
~ ' ~ I iAi I
I
',.g '4J
~
i .3.
~ f i 85025 I~
CI <<
I~ ' - ~ f .,~l
~ ~ ~
~ ~
-i. .:(
~
~
~
~ ~
~ +
'0,
~ I~
I ~ ~ I~ ~
~
1.-
~
~ ~
80000 0 10 '... =,20 . ', 40 STEAN GENERATOR TUBE. PLUGGIftG (%)
A Best Estimate Calculated RCS Floe>
B -- Curve A Hinus 4200 GPii: Lowest Expected Measurement C -- Cul ve B TlfIles 0 98~
0 -- Curve B Times 0,91
Qi 0 I,~
~
'1
~
~ ~
~
~
~ FIGURE 2 LOSS OF FLOH CORE FLOH VS TItlE 1.2 i
~ I ~ ~ ~ ~-
C
~ ~ ~ I= ~
~
I I ~ I 1.0 I ~
I ~ ~
I i
~ = +I 1
~-
0 8
~ ~
t
~
~
l
~ ~ ~ ~
II ~ =- I -i ~ ~
I .'
I ~
I 3
-',-...N, 0.6 =-I '
I
~ ~ l << ~
~ ~
t " ~ !
C>
-4 I- l I
- ~
014 I
=
I I I
= ~ -I
~ ~ ~ . ~ -I ~ ~ - -. - I
~ ~
CX
~
~ I=
.~ -... ~ . .= I C) 0.2
~ 1 ~
I ~
~ ~
0.0
' 10 14 18 0 2 TItlE (SECOtlDS)
I I' 4 , ~
~ ~ ~
FIGURE 3
~ '
p LOSS OF FLOW NUCLEAR PAltER VS TI!lE l.2 I 1 II 1
':i '. ~: "'
~ i
- -I""
~ I l i
~ !
r 1.0 '---- - J I
I I I - = ~
L
- ';*. I t i 0.8
~ I * ~ ~
~ :-= I: r I~ ~
~ . ~ : " ~
I '- '."i .'.
j ~
I
! I I
' I - ~ I O.e
~ 4 ~
M i=
=-
~
I CL
- 0. it = I - ~
~ ~ 1 ~
0.2
~ ~
~ '
~-
~ ~
4 0.0 0 2 .
4 6 8 lo TItlE (SECONDS)
0 t ~'
FIGURE 4
~ ie 7 LOSS OF FLOH HOT CHAl'll,!EL flEAT FLUX VS TII1E 1.2
~ h ~
h V
\
r
.*-- '. - L.:.l-: ~ ~ ~
I t
I 1.0
~ ~
- .I"... I
~ ~ ~ ~ ~
I I
~, w I I I
~
~ ~ ~ h ~
~
t ~
"~ ~
0 6 I
1 ~
c
\ ~ i.. ~
h ~
LL::::. ':
~ *
- ~
I I
~ ~
~ ..
"~ ~ ~ + ~
~ '-"I
~ i t
I ~ I ~ ~
= ~ I ~ -,
f 0.2 l ~ ~
~ rh . ~ hl ~ ~ ~ ~
h ~ ~ ~ ~ t
~ I
~
I
~ = ~ ~ I * ~, ~
--:.-'>>I h
l I ~
!....1.
-~ h ~
\~ .....,.... !
~ h 0 -
2.0 40 60 80 100 TINE (SECONDS)
I ~ ~
4i f
~ .~ I ~
~ LV \ ~
~ LOSS OF FLOM DNBR VS. TINE (TriIt~oLE cELL)
'L=
2.4 P ~ I t ~
2.2
~ ~ ~
'2.0 1.8
'L ~
'.6
~ rt I ~
f
~ f
. I1.4 t'
f' I 1~ ~' ~
~
ttI
~,I F~ ~ ~ ~ ~ ~
I ' ~
t 4
i ~
~ ~ t i ~ ~
1.'2 ~ I h
~ ~
~ L i
~ h h ~ ~
~ ~ I ~ ~ ~ t ~ t A ~ ~
~ t
~ i I
1.0 ~
0 (Time) (Seconds)
1 0
~ ~> ~
, I
,kb ta I ~
t .
- I ~
I 2.0 ~ 1 ~ ~ ~
' ~ I -1 ~ I
~ ~
Variation in Heactivity Insertion Hate
~ 1 I~
'itb initial Boron Concentr"tion for bb'r' I
~
~ ~
'I 1 ~
I1 ~ ~
~ ~
I ~
~ I 1 ~ ~
- 1 I'
4
~ ~
- a ~1'oron Dibution Hate of 230 GPR tl ~
~ I 1 l>> ~ ~
I ht ~
h 1 gs
~ ~ >
~ I~I ~ ~ ~ 1 ~ II ~ abs
~ ~
~ I ~ ~ I ~ ~
t,
~ I ~ ~ II ~ I ~ ~ ~ > ~ I "1 I 1 t ~
th ~
1
~ 1
~ I
~ ~ Sls
~ ~
~
I ~
II
' Wl '1 ~ ~
~ .I I
~ I
~ >ss ~ I ~
4
> ~
1
~
I I ~
I~ 1 ~ ~
I I ~ I 11 Q ~ ~
,1 I~
0 l ~ 1 ~ ~
~ ~
I
~
( 1 I!
b ~
~
k 1 ~
C'I ~ t 1 ~ ~ ~ I I ~ )
asb
~ .
' 1 1 I '
11
~
1 11 "t 4
Il ~
~ ~ ~ ~
~ ~,
I 1 ~
I~ ~ I ~ ~ - 1' ~ I~
1 ~ I 1 I~ ~
I I~ t1
~>>
~ 1 ~ ~ ~ ~
II I "!
Q I I I ~
I~ I~ I I s ~ ~
~ ~ I ~ . i ~
1 ~ I~ ~ ~ ~ ii I~ ,. ~ Ia I
I ~
~ I I~ ~ ~
I
~ ~ I
~ ~ ~
I ~
I \l 1 1
I~
0 ~ ~ I1 I ~
la 1>a I
~ R I
~
~ ~
M ~ II I ~ ts 'I ' ~
It ~ ~ \ I ~ ~
Ikk 1 ~ '1 ~ ~ I 1
- I; I r., ~ - I - I1
~ ~ ~
~ ~ 4 ~
ml IlaI 1
~ ~ ~
~ ~ ~
tt 1>
~ ~
~
~
I> 1I Is
~ ~
- ~
's
~ ~ 1 ~ ~ ~ >
~" 1 ~ 1> ( I 1 ~ I I ~ I ~ 4 I 1 ~
I, ~ I1 ~ 4 I I~ ~ I 4+ ~ ~ ~ ~
~ ~ I '1 ~ ~
I ~
~ A 4
~ ~ ~ ~ ~
stL I yJ ~ ~ I ~ ~ ~ h U 1 ~ 1
'Ih ~ :
I1 ~
~ ~ ~ ~ ~
~ t I Cii ~ I ~ ~ ~ lbbb 4 ~ ~
Q ~ ~ ~
4','1 I ik' ~ I~ I I' I I 1 ~ I
",I ~
~ I~
I~
~ ~ , ~ I~ ~ ' ~ ~
>>1 1 ~
>>111
~ ~ 1 ~ ~ I ~
~ I ~ I 14>>
'I ~ I I ~ ~ ~ I ~ It ~ ~ 4 ~ I 1 ~ '
>(>> ~ =
b; I*
' 1 1
1
'1
~ ~ I ~ 1 ~ ~
II
~ ~ ~ '1 ~ I ~ ~ ~ ~ 1 ~
I ~ 1 ~
~ II ~
~ I~
~
1>
I ~ ~
~ I ~ ~ is ~ ~ ~
't 14 Il 1
1',
I ~ ~ I 1 ~ 1 II I I~ s;I
~ 1 ~ ~ ~ ~
I Ik ~ ~ ~
I ~
0 1 ~ ~
0 'I 1000 '000 2500'>.'.
Boron Concentration, PPN I 0
0 ~ ~I
, ~, fa
Atta.hment No. 1
~ 0 7 The standard design methods utilized in this 955 thermal design flow .
analysis of Cycle 5 of Turkey Point Unit 4 ensure that the generic DNB margins (Reference 1) exist in the Cycle 5 design and are avail-able to offset DNBR reductions due to fuel rod bow. Utilizing these generic DNB margins to offset the current NRC rod bow penalties for low parasitic 15xl5 fuel would result in the F H reductions vs. burn-up curve of Reference 2 being applicable for Cycle 5.
These generic DNB margins are, however, more than enough to offset the rod bow penalties calculated in Reference 3 (partial rod bow test results) for burnups up to 33,000 f1~rfD/t<TU. For fuel regions with high burnups, the depletion of fissile nuclides and buildup of fission prod-ucts greatly reduces power production capacity. These combined burnup effects reduce F sufficiently to cover residual rod bow penalties be-yond a region average burnup of 33,000 HllD/HTU.
Ql
~ ~
1'1
REFERENCES HRC's "Interim Safety Evaluation Report on Effects of Fuel Rod .
Bowing on Thermal Margin Calculations f'r Light llater Reactors",
- .-. Revision 1) Feb. 16, 1977.
3.' letter (C. Eicheldinger) to NRC (D. R. Ross, Jr.), NS-CE-1580, dated Oct. 24,"1977.
ol 'l
.Ql i- ie <i
~ 0
STATE OF FLORIDA )
) ss
)
Robert E. Uhrig, being first duly sworn, deposes and says:
That he is a Vice President of Florida Power 6 Light Company, the Licensee herein; That he has executed the foregoing document; that, the state-ments made in this said document, are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.
Robert E. Uhrig Subscribed and sworn to before me this 0~ay of l9 ~8 NOTARY PUBLIC, and for the County of Dade, State of Florida ÃOYrV 'JSUC /TATS Ot: AMlOA ct !ARC!f MY COMNSSIOM EXRihSS MARCH 27, 1P2 My COmmiSSion eXpireS BDMD D TRRG IIAYMARD BGMDIMG AGEMcY
'gl A.,
' ~
~ (~
,).$ )ll)H)p),
(y
\ I' r~