ML17151A800
| ML17151A800 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/31/2017 |
| From: | Commonwealth Edison Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML17151A800 (79) | |
Text
{{#Wiki_filter:DRESDEN II DPR-19 Amendment No. 82 1.1 SAFETY LIMIT BASES (Cont'd.)
.:~1 available for any scram analysis, Specification 1.1.C.2 will
-'i ... be relied on to determine if a safety limit has been violated.
During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. l f reactor water 1evel should- drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel* provides adequate margin. This level will be continuously monitored whenever the recirculation pumps are not operating.
*Top of active fuel is defined to be 360 inches above vessel zero (see Bases ~.2).
LIMITING SAFETY SYSTEM SETTING BASES FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the units have been analyzed throughout the spectrum of planned operating conditions up to the rated thermal power condition of 2527 MWt. In addition, 2527 MWt ii the 1icens~d m~xim11m steady-state power level of the units~ This maximum steady-state power 1evel will never knowingly be exceeded. See RefeFeRse X>I NP' 79 7tr Conservatism is incorporated into the transient analyses which define the frl:PR operating limits. Variables which inherently possess little or no uncertainty or whose uncertainty has little or no effect on the outcome of the limiting transient are selected at bounding values. Variables which possess significant uncertainty that may have undesirable effects on thermal margins are addressed statistically. Statistical methods used in the . transient ana yses a scr1 e in The MCPR operating limits are established such that the occurrence of the limiting transient will not result in the violation of the ~CPR Fuel Cladding Integrity Safety Limit in at least 95% of the random statistical combinations of uncertainties. In general, the variables with the greatest statistical significance to the consequences of anticipated operational occurrences are the* reactivity feedback associated with the formation and removal of coolant voids and the timing of the control rod scram. ~ *.j
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DRESDEH II DPR-19. .
.Amendment Ho. 5) , ¢ , '2 , 95 2.1 LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.)
Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps. The bases for indi'li'idual trip settings are discussed in the fol-lowing paragraphs. For analyses of the thermal consequences of the transients, the MCPR' s stated in paragraph a:§ 'fiias the--- 3. 5. L. limiting condition of operation bound those which are conser-vatively assumed to exist prior to initiation of the transients. A. Neutron Flux Trip Settings
- l. APB!! Flux Scram Trip Setting (Run Mode)
The average power range monitoring (APR!!) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because fission chambers provide the basic input signals, the APBH system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of. the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 120 percent scram trip setting during dual loop operation or 116.5 percent during single loop operation, none of the abnormal operational transients analyzed violate the fuel Safety Limit and 1 there is a substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides aven additional ma~gin. An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because. of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. B 1/2.1-11 3918a 84010
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DRESDEN . I I * : DPR-19
- Amendment No. ~, 'j!J, 82 2.2 LIMITING SAFETY SYSTEM SETTING BASES In cOiltpHance with Section III of the ASME Code, the safety valves must be set to open at no higher than 103% of design .pressu!"e, and they must limit the reactor pressure to no more than 110% of design pressure. Both the neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thus exceeding the pressure safety limi~. The pressure scram is available as a backup protection to the direct valve position trip scrams and the high flux scram.
If the high flux scram were to fail, a high pressure scram would occur at 1060 psig. *Analyses are perfonned as described in Fefei:eR&e XN >IF" 79 7'J'Z.fo~ each reload to assure that the pressure safety 1imi t 1s not exceeded. . * .... ( [ ~ flf'JP r-l'R.C.-9Pf'Ovi:d Me~olo'l'f
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DRESDEN II DPR-19 Amendment No. 82, 83, 84, 95, 104 TABLE 3.2.3 rnsiRUMENTATION THAT INITIATES ROD BLOCK Minimum No. of Operable Inst. Channels Per Trip System (1) Instrument Trip Level Setting 1 APRM upscale (flow bias) (7) Dual Loop Operation Less than or equal to ,. (.58 w- plus 50)/FDLRC (See N9te 2) Single Loop Operation Less than or equal to (.58 WD- plu~ 46.~)/FDLRC _(see Note 2) 1 APRM upscale (refuel and Less than or equal to Startup/Hot Standby mode) 12/125 full scale 2 APRM downscale (7) Greater than or equal to 3/125 full scale Rod block monitor upscale (flow bias) (7) Dual Loop Operation Dele.r-c... Single Loop Operatio_n 1 Rod block monitor Greater than or equal to downscale (7) 5/125 full scale 3 IRM downscale (3) Greater than or equal to 5/125 full scale 3 IRM upscale Less than or equal to 108/125 full scale 3 IRM detector not fully N/A .i inserted in the core j j 2 (5) SRM detector not in startup position (4)
)
2 (5) (6)- SRM upscale Less 5than or equal to 10 counts/sec. 1 (per bank) Scram discharge volume (LT/E) 26 inches above water level - high the bottom of the instrument volume
*.)
Notes: (See Next Page) 3/4.2-12
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_,._.,;--** DRESDEN II 'DPR-19 Amendment No. 19, 75, 79, 82, 104 3.3 LIMITING CONDITION FOR OPERATION 4.3 SURVEILLANCE REQUIREMENT (Cont 1 d.) (Cont 1d.)
- 2. The maximum scram 2. At 1s*week intervals, insertion time for at least 50% of the con-90% insertion of any trol rod drives shall be operable control rod tested as in 4.3.C.~ so shall not exceed 7.00 that every 32 weeks all seconds. of the control rods shall have been tested. When-ever 50% or more of the control rod drives have
**:***:** ....... . been tested, an evaluation shall be made to provide reasonable assurance that proper-contro l rod drive performance is being
.:.ril\.~,., 'B ~ maintained.
D. Control Rod Accumulators D. Control Rod Accumulators At all reactor operating Once a shift check the pressures, a rod accumulator status of t~e pressure may be inoperable provided and level alarms for each that *no other control rod accumulator. in the nine-rod square array around this rod has a:
- 1. Inoperable accumulator,
- 2. Directional control valve electrically disarmed while in a non-fully inserted position.
3/4.3-11
Insert B
- 3. Following completion-of each set of scram testing as described above, the results will be compared against the average scram speed distribution used in the transient analysis to verify the applicability of the current MCPR Operating Limit. Refer to Specification 3.5.L.
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*_f~ DRESDEN II -DPR-19
"~ Amendment No. fl, ~. 82
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3.3 J~--~,-*-** LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
- 3. The operability of the scram discharge volume vent and
., drain valves assures the proper venting and draining of the volume. This ensures that water accumulation does not occur which would cause an early termination of control rod movement during a full core scram. These *
-*.: specifications provide for the periodic verification that
, __~
.., the valves are open and for testing of these valves under*
*l
.:*1 i reactor scram conditions during each Refueling Outage.
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.J B. Control Rod Withdrawal
*..l -rke_ ,q "' F . .
l ** N""RC-a.ypraoo.:/] 1. Control rod dropout accidents as discussed in RefereAee c...... l %.-rlxJoLor'1 Ls d . ~-MN NF ~Q l 9, 'lol. 1f-Can lead to si gni fican~ core damage. ~---.-*:-.
- o;::;.::.;_.
j 5Pcz ' 'Tt. '"'- J If coupling integrity is maintained, the possfbility of a . . .. -~_. i *+m. r, 0 '1.. ~. G. A '-I rod dropout accident is eliminated. The overtravel .* i ** 1 position feature provides a positive check as only j uncoupled drives may reach this position. Neutron
! instrumentation response- to rod movement provides a
- J verification that the rod is following its drive. Absence 1 of such response to drive movement would provide cause for 1
suspecting a rod to be uncoupled and stuck. Restricting j ....... recoup 11 ng veri fi cations to power 1eve1 s above 2oi
. ,* provides ~ssurance that a rod drop during a recoupling 1 ve~ification would not result in a rod drop acci~ent.
*1 l
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- 2. The control rod housing support restricts the outward l
i movement of a control rod to less than 3 inches in the*
.~ . *~
- extremely remote event of a housing failure. The amount i
of reactivity which could be added by this small amount of
.J rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system *. The design basis is given in Section 6.6.1 of the SAR, and the design evaluation is given in Section 6.6.3. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing ** Additionally, the support is not required if all control rods are fully inserted and*if an adequate shutdown margin with one control rod withdrawn has been demonstrated since the*reactor would remain sub-critical even in the event of complete ejection of the strongest con-~rol rod.
I . .
- 3. Cont~ol rod withdrawal and insertion sequences are esta~lished to assure that the maximum insequence
- indiv\idual control rod or control rod*~e~weR'e~which are
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7 i
DRESDEN II DPR-19 Amendment No. 82 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
- l~J
- ~ withdrawn could not be worth enough to cause the rod drop
*~ accident design 1i mi t of 280 ca1I gm to , be exceeded if they
<J were to drop out of the core in the manner defined for the
~~ Rod Drop Accident. These sequences are developed prior to
..*!, initial operation of the unit following any refueling
~ outage and the requirement that an operator follow these
.';j sequences is backed up by the opera ti on of the RWM or a
- .,j, second qualified station employee~ These sequences are
] developed to limit reactivity wortlis of control rods and,
- ~~ together with the integral rod velocity limiters and the r action of the control rod drive system, 1imit potential
- 1 . reactivity insertion such that the results of a control
~i . . rod drop accident wi11 not exceed a maximum fuel energy rl.e- ANF Ni<c _ content of 280 cal/gm. The peak fuel enthalpy of 280 ,J .J me.Ti..od.
l
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olo77 l ,-::.re.d '"' cal/gm is below the energy content, 425 cal/gm, at which rapid fuel dispersal and primary system damage have been J Spec.f.~,a-t ".6, IJ. '(. 1 f~~nd to occur based on experimental data as is discussed
! ~Reference XN NF 8Q 19, 't'o1 tame 1. ~ '.:1 The analysis of the control rod drop accident was
- ~*
originally presented in Sections 7. 9.3, 14.2. l.2 and 14.2.1.4 of the Safety Analysis Report. Improvements in analytical capability have allowed a more refined analysis _:J of the control rod drop. accident.
- 1
.i .-.J Parametric Control Rod Drop Accident analyses have shown that for wide ranges of key ~eactor parameters (which ..-i
-~
I envelope-the operating ranges of these variables), the j fuel en~halpy rise during a postulated control rod drop
~i accident remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient effective delayed neutron fraction and maximum four-bundle local peaki.ng factor are compared with the results of the parametric analyses to detennine the peak.
fuel rod enthalpy rise. This value is then compared against the Technical Specification limit of 280 cal/gm to demonstrate compliance for each operating cycle. If cycle specific values of the above parameters are outside the range assumed in the par~metric analyses, an extension of the analysis or a cycle specific anrlysis may be required. Conservatism present in the analysis, results
) .\ of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop 1e
- .]
'i I Accident analysis are provide@d
~91 WIR8 1 ( SuppJ ements 1 and 2) .~
in RefeFeAee XN N~-ao l9f2-
"?le AN F Nile -=ri'"*M~..cl -
B 3/4.3-17 [ Me."'°d0Lo7y L,~re.J_ '""- S'-pe.c,f,cc..T1.:1"'l 6. (,.A, 't ..
DRESDEN II 'OPR-19 Amendmen_~:.~o.. 58, 75, 82, 104 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) analyse~':~iii(J::Js. *also included in the allowable scram insertion times specified j_ri Spec.ification 3. 3. C. The bounding value described above was used ;n*the trans-ient analysis. The performance of the individual control rod drives is monitored to assure that scram performance is not degraded. Fifty percent of the control rod drives in the reactor are tested every sixteen weeks to I"'-x.,,. C. verify adequate performance_;t -{ The scram.times for all control rods are measured at the time of each
..... refueling CllJ'tage. Experience with the plant has shown that control * -f drive.insertion times vary little through the operating cycle. The history of drive performance accumulated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating time is accumulated. The probability of a drive not exceeding the mean 90% insertion time by 0.75 second is greater than 0.999 for a normal distribution.
D. Control Rod Accumulators ' The basis for this specification was not described in the SAR and, therefore, is presented in its entirety. Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean core .. The worst case in a nine-rod withdrawal sequence resulted in a keff less than 1.0 -- other repeating rod sequences with more rods wi~n-
- drawn resulted ink greater than 1.0. At reactor pressures in excess of 800 psig,etien* those control rods with inoperable accumu-lators will be able to meet required scram insertion times due to-the action of reactor pressure. In addition, they may be normally inserted using the contra 1-rod-dri ve hydraulic system. Procedu.ra l control will assure that control rods with inoperable accumulators will be spaced in a one-in-nine array rather than grouped together.
E. Reactivity Anomalies During each fuel cycle excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration.* As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reac-tivity. Furthermore, using power operating base conditions permits B 3/4~3-20 . -~
\
Insert C Observed plant data-or Technical Specification limits (Specification 3.3.C) were used to determine the average scram performance used in the transient analyses, and the results of each set of control rod scram tests performed per Specification 3.3.C during the current cycle are compared against earlier results to verify that the performance of the control rod insertion system has not changed significantly. If a test performed per Specification 3.3.C should be determined to fall outside of the statistical population defining the scram performance characteristics used in the transient analyses, a re-determination of thermal margin requirements is undertaken as required by Specification 3.5.L. A smaller test sample than that required by Specification 3.3.C is not statistically significant and should not be used in the re-determination of thermal margins. Control rod drives with excessive scram times can be fully inserted into the core and deenergized in the manner of an inoperable rod drive provided the allowable number of inoperable control rod drives is not - exceeded. In this case, the scram speed of the drive shall not be used as a basis in the re-determination of thermal margin requirements.
DRESDEN II DPR-19 Amendment No. 75, 82, 94, 104, 107 3.5 LIMIJ;ING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (cont* d.) ,. (Cont 1 d.) operation is permis-sible only during the succeeding seven(7) days provided that during such time the a...1pr~ HPCI subsystem is operable.
- rr-ia..re ~fell e*.1tF1~APLHGR
- reduction factors *(multi- . - . * * *
- pliers) are applied to f\.../-TL~ MlfPLitc..1<_ £,.... ,rs H ~~Pe 3. §*l-j the Auto- '
- . . matiC Press*ure Relief Subsystem of ECCS shall be considered operable\. -
(1) Q,89 fef 8x8 f~el,c__ . SP (2) Q. 76 fo1 9x9 fuel ~[-rf..e. ~APLHG1< L,,,.,,T5 a.....d ~
- 3. From and after the
- a:;rrupru*.-re. "111PtHr;.R. cl. r date that two relief ette. ~OuAd '"'- ,-/..._ fe llC.~'<1.. 'f>;>-c.-rors valves are found or made L,.,,.,~ Ke..ro e.. Core.. Cperc:::.r1; to be inoperable, reac- ~,_.
tor operation is permis-sible only during the succeeding seven days provided that during such time the HPCI subsystem is operable and the multipliers specified in 3.5.D.2 are applied.
- 4. If the requirements of 3.5.D.1 cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to below 150
.' psig within 24 hours .
'i E. Isolation Condenser System E. Surveillance of the Isolation Condenser System I ~
shall be performed as follows: 3/4. 5 DRESDEN II DPR-19
*Amendment- Ncr. --82 ;--a4-;-* 95 ;:* 104~ '"*
3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEII.I.ANCE REQUIREMENT (Cont'd.) ~ (Cont'd.) I. Average Planar LHGR I. Average Planar Linear Heat Generation ~ (APLHGR) During steady state power operation, the Average The APLBGR for each type of Planar Linear Heat Genera- fuel as a fanction of -c___ tion Rate (APLHGR) of all the e:ve~age planar exposttre c.- rods in any fuel assembly_.t, fer G.E. fliel *and avenge
- s a unction o average lnmdle exposuze for A1ft' ~
p nar exposure for G.E. ~shall be determined fue *and average bundle daily during reactor expos e for ANF fuel at operation at greater than any axi or equal to 25l rated not excee thermal power. age planar GR shown i Figure 3.5-1. For ation during Single L op the values of Fig shall be decrea d a multi-plicative fac r of 91. If, concurrent! , one Autom tic Pressure lief Subsyst relief V- ve is out-of-se the v ues of Figure 3.5-1 sha be decreased by a multi-p *cative factor of o;s9 for x8 fuel and 0.76 for 9x9 fuel. If at any time during opera-tion it is determined by normal surveillance that the limit-ing value for APtHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to
~ within the prescribed limits j within two (2) hours, the J reactor.shall be brought to
. .i the Cold Shutdown condition
,, within 36 hours. Surveil-
! lance and corresponding 1 action shall continue until
' reactor operation is within the prescribed limits
- 3/4.5-15
DRESDEN II DPR-19 Amendment No. 76, 92, 95, 104
- 3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)
J. LOCAL STEADY STATE LHGR 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) J. Linear Heat Generation Rate (LHGR) During steady state power The JJIBR"iUiall be checked operation above 25% of rated daily during reactor thermal power, the linear heat generation rate (LHGR) or equal to 25% rated of any rod in any fuel assembly thermal power. at any axial location shall not exceed its maximum steady state LHGR (SLHGR) value shown in _,__ _ Figttre 3.5 lA (consists o'f ~
~~hat is, the fllel Design timiting Batio ~
-E-x*6n £ue+/-3-(FDLRX) shall not be greater than 1.0 where +he Co re o,.elra f/nJ J-1'm1'h ReptJrr+.
LHGR FDLRX = SLHGR
- state e 3.5-lA depicts the s and 9x9 fuel values for 8x8 Dele-te If at any,.time during operation above 25% rated thermal power, it is determined by normal surveillance that FDLRX for any fuel assembly exceeds 1.0, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the FDLRX is not returned to within the prescribed limits within two (2) hours, the reac-tor shall be brought to the Cold Shutdown condition within 36 hours. Surveillance and corresponding action shall continue until reactor opera-tion is within the prescribed limits. 3/4.5-16
- i DRESDEN II DPR-19 Amendment No. 104 3.5 LIMITING CONDITION FOR OPERATION 4.Sn SURVEILLANCE REQUIREMENT (Cont'd.) - (Cont'd.)
K. Local Transient tHGR K. Transient Linear Heat Generation Rate (LHGR) At any time during power The fuel design limiting operation, above 2si rated ratio for centerline melt thermal power the fuel design (FDLRC) shall be checked limiting ratio for centerline daily during reactor opera-melt (FDLRC) shall not be tion at greater than or greater than 1.0, where iepe...
~file!~
equal tp 2si rat~d thermal power
- FDLRC = ..,.CI.H~G,.;-R"'"')("""'l"'"".
(TLHGR) (FRP) 2,,...,)....,.. If during operation, the FDLRC exceeds 1.0 when operat-ing above 2si rated thermal power, either:
- a. The APRM scram and rod block settings shall be reduced to the values given by the equations in Specifications 2.1.A.1 and 2.1.B. This may be accomplished by increas-ing APRM gains as described therein.
- b. *The power distribution shall be changed such that the FDLRC no longer exceeds 1.0.
3/4.5-17
DRESDEN II DPR-19 Amendment No. 82, 87, 95, 10
*_,~--~,-****
MAPLHGR LIMIT VS. BUNDLE AVERAGE EXPOSURE ANF 8x8 FUEL I
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10.0 0 10.000 20,000 30,000 40,000 BUNDLE VERAGE EXPOSIJR[ /vru) The above graph is bas on the following MAPLHGR summary for ANF 8x8 fuel design. r Bundle.Average GR Exposure (MWD/ ) Limit /ft) . 0 13. 13.0 . 10,000 15,00 13.0 18, 0 12.85 20 000 12.60
,000 11.95 o,ooo 11.20
- 35,000 10.45 Figure 3.5-1 (Sheet 1 of 3) I 3/4.5-18 l
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y ***. *************** DRESDEN II DPR-19 Amendment No. 82, 87, 95, 4 MAPLHGR LIMIT VS. BUNDLE.AVERAGE EXPOSURE ANE 9x9 FUEL 12.0
- 11. s
- 11. 0 10.5
** .. -=9.5 10.0
- Kg* 0 er
~8.5 c..'
~B 0
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- j 7.5 7.0 6.5
.i -:, 6.0 0 10,000 20, )01 30,000 *40, 000 BUNDLE RAGE EXPOSURE UWJ/Miu)
The above graph is base on the following MAPLHGR summa for ANE 9x9 fuel design. I Bundle Average LHGR Exposure (MWD/ Limit kw/ft)
., 0 11. 0
. j
., 5,000 11. 7
.j 10,000 11.40 j
*1 15,00 10.s5
.1 20, 0 9.70
"'I
*: 25 000 8.85 1.
,000 8.00 s,ooo 7.15 40,000 6.30 Figure 3.5-1 l
(Sheet 2 of 3)
*.j 3/4.S-19
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DRESDEN II DPR-19 Amendment No. 95, 104 MAPLHGR VS. AVERAGE PLANAR EXPOSURE GE 8x8 LTAs
---- 11
-=.: 10 DC u
A .. 9
-=
8 7 D 5 10 25 JO 35 40 . 45 (;td/S l) .. The above graph is based summary for the GE LTAs fuel design: Average Planar
.7 ~ Exposure (GWd/St)
- ~
Q.2 l .5 1.0 11. 5.0 11. 9. 10.0 12.l 15.0 12.l
- 2
. *.~ 20.0 11.9
*! 25. 11.3 .. ~ **! 30 0 10.7
- 1'
.o 10.2 ..; 41.6 8.8 Figure 3.5-1 (Sheet 3 of 3) 3/4.5-20
- -1
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<{l _ -*- _* **- . - *- ~ L::(\G
--r>£_ L c.1\.- _ r n DRESDEN II DPR-19 Amendment No. 104 _-
STEADY STATE I.INEAlf.HEAT GENERATION RATE LIMIT (SLHGR) VS *. ANF hi
-l '-~!
c: a u
.,c:
-u
. u
- c a
JD Pia ar Exposure ( , d/MTU)
~
ANF 8x8 Fuel Exposure LHGR 13.4 o.o 16.0 o.o 14.5 13.4 25.4 14.1 5.0 14.5 42.0 9.3 25.2 l0.8 48.0 7.2 Figure 3.5-lA i I I 3/4.5-21 I
DRESDEN II Amendment No. 104 TRANSIENT LINEAR BEAT GENERATION RATE LIMIT (TI.HGR) VS. NOD EXPOSURE FOR ALL ANF FUEL
*-E-
-a D::
c:
/-*-0
...u D
c u u
-=.,
D 11;1 c:
*- Exposure Exposure tHGR **l
. .'."\ o.o 19.2 25.4 16.9 43.2 10.8 48.0 10.0 Figure 3.5-lB
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3/4.5-22
*'] . - *--- *-
.~~r;;~£?:S1::'::~;":;?0-:S?~l['.T:5':'P; :..*."'P*~~o-~?;-,:~::i:'.?~:;f'""}!S;:;;;~:C\::~:.;;/;;.§~i.~;!.::::::"'?i'.*::;;s:.~~~~~>>"-TI:~s>I%'.3~-7S*;':f~iZ003':\'-70-~5',*.*-.,,,,:','.::*'~'"{?i:t;,s
DRESDEN .II-.... - * - -* DPR-19 * --
- Amendment No. 82, 104 3.5 LIMITING CONDITION FOR OPERATION . 4.s*<sURVEILLANCE REQUIREMENT (Cont'd.) ~
~;~~~,~~s~:.~ I d * ) . .
;.~:; .:..~* .-
- t. Minimum Critical Power .,.
- ~:L*:, '.'Hinimum:*Critical ~
~ (MCPR) ~
a.L.L Core. flows J
/, During steady state
,_ ** "' '. -~ (MCPR)
HCPR shall° be determined Ir\. Mo..o11.v...t. or
.f'Low co...,,-rrciL I
' 1 a.vro Operation at'Vrated COL~
-fh'#,- HCPR shall be greater than or equal to l-.39~
daily during a reactor power operation at greater than or equal to 251 rated thermal power and following I- ~oc~: -~ny diange in power level or cor-e flows othe:r than* . * - or distribution that would ted, the MCPR_operating *e:ause operation with a limit-l" it shall be as follows: ing co~trol rod pattern as described in the bases for Specification 3.3.B.S. The value from Figure 3.5-2 she or
-*:..; c. The value from Fig
' 3.S-2 sheet 2~
- i
_:_j __ ----- - 2 1* - - During Single Loop opera-tion, elh"Lhe rated flow MCPR operating limi1?f shall be increased by an additive factor of 0.01. If at any time during steady state power operation, it is determined that the limiting value for MCPR is being exceeded, action 3/4.5-23
DRESDEN II DPR-19 Ame~dment No. 82, 84, 95, MCPR LIMIT FOR REDUCED TOTAL CORE FLOW Cl a: e1.10---------~--------------------------------------------------'----------- Cl
=1.so .....~~-+~~~~~~~1--~~-1-~~~-1--""~~--~~-'
c
=
$1~50t-----------1------~~.......--------1-----------'"---------""'"'"----------'----------1
~1.40t-----------4---------~~----~._--------_,_----rl--~------------..a----------i u
- =1.301--~~--+-~~~-+-~~~f--~~~~~~-1-~~~f--~~-1
.~1.20 c
.. ......~~-+~~~-+-~~--.:.~~-.'--+~~~~~--~~~~-'
Cl
- 1.101--------~----------J----------4-~------~--------1~-------"------_.;;;::..
u
=
- 51. o~ 0
.-. 40 50 80 90 100
- tu RE FLOI (I IATED, II ll9/HR)
*l I
The above curve is base on the following MCPR operating SWllllary for
.; reduced core flow and 11 fuel types:
Total Core Flow C'X. Rated) 100 1.10 90 1.16 80 1.23 i.
- j' 70 1.30 60 1.39 so 1.51 J 40 1.65 l . Figure 3.5-2 (Sheet l of 2) 3/4.5-25
.j 3
DRESDEN II DPR-19 Amendment No~ a2*, 94, 95, HCPR OPERATING LIMIT FOR AUTOMATIC FLOW CONTROL A.
* ~ 1. 6Ot-----+-----~...;::ii,,,-=--~-=-~~---""'""'---'----.i
- 1.50t-----+---+-~~-+---.;;
- ::91ii4-::~--~=---~.f-----1
~
* ** Su l1h 1 B 1. 4D~----.i----t---~-1--..;._--l--~_::i=::::::~..::i::;;~::::J su 1111 2 __*,
~ 1*J 0 Su 11 h 3
- 1-.20 ......~--+------+-----+--~-'-~----1----~--~
c
~
c
- 1. 1~~~--~.o--~s~o-~...._...,._~___.ao___g.._o_ __,1oo II llLB/HR)
The above curve is based on e following HCPR operatin Automatic Flow Control an all fuel types: HCPR Operating 1.35 1.39 1.35 1.39 1.40 1.44 1.44 1.48 1.50 1.54 1.56 1.61 1.66 1. 70 1.81 1.86
*Colwmi headers are HCPR operating limits at rated flow.
Figure 3.5-2 (Sheet 2 of 2) 3/4.5-26 ~~ . - -. - -- -- - -*****- *- -- - -- *- - - - - - - .. -
- )~TEWc~fit\f;l~f~;~1g{~~~i:~f§grf.s8\'.7:~r~;~~;;~~r?~~r~;~:~ffe£%:}'~?~Tn:r0*0;.!Jf{~:0~zy1g;'.K1~?~;~,?~i~0'V~~~-~~I=~~t~l;~:-~~~~-~;:\5
-* ---~-- ---- . *-***-*--.; ___ ___
.DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
ANF has analyzed the effects that Single Loop Operation has on LOCA events ERefireR;Q 4)4! For breaks in the idle loop, the above Dual Loop Operation results are conservative ~Refe1e11ce 1}'?- For breaks in the active loop, the event is more severe primarily ezppropr--ic..-re. - - - due to a more rapid loss of core flow. By .applyingv-ill-multipl ica-tive B.9lrreduction factor to the results of the previous analyses, all applicable criteria are met.
'"'be.Le.re-J. Local Steady State LHGR
.,., This specification assures that the maximum linear heat genera-
~
1 tion rate in any fuel rod is less than the design linear heat generation rate even if fuel pellet densification is postulated. This provides assurance that the fuel end-of-life steady state criteria are met. 1 of Coo ant cc1dent Analyses Report for Dresden Units Quad-C1
- s Units 1, 2 Nuclear Power Stations," NED0-24146 sion 1, Apr1 79 .
.1
;-J I
- j (4)ANF , "LOCA-ECCS Analysis for Dresden Units ion with ANF Fuel," September 1987 .
8 3/4.5-36 l
Insert D The calculational procedure used to establish the maximum average planar LHGR values uses ANF calculational models which are consistent with the requirements of Appendix K IOCFRSO. The approved calculational models are listed in Specification 6.6.A.4. l 1 -1
- j
-1
*** *-~---*- ~. -* * *---~ - **.;.,.._ _____ .._.._*.:. ** _** - - - ~*---*- * .....:....--;.**..:.O .... - .
DRESDEN II DPR..,19
* *.:___**Amendment No. 82,84,95,104,101 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont 1 d.)
K. Local Transient LHGR This specification provides assurance that the fuel will neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and *termi-nating at 120% of rated thermal power. L. Minimum Critical Power Ratio (MCPR) The steady-state values for MCPR specified in the Specification were determined using Uie*1'11ERP4E~thermal limits methodology escr1 e in XN-NF*8Q 19, Velt1me 3~ The safety limit implicit in the Operating limits is established so that during sustained operation at the MCPR safety limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition. The Limiting Transient delta CPR implicit in the operating limits was calculated such that the occurrence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combina-tions of uncertainties. Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive in terms of thermal margin requirements. The generator load rejection/turbine trip without bypass is.typically the limiting event. The thermal margin effects of the event are evaluated with t.Al#+MERME~Methodology and appropriate MCPR limits con-s1stent with the~ritical power correlation are determined. Several factors influence which transient results in the largest reduction in critical power ratio, such as the cycle-specific fuel loading, exposure and fuel type. The current cycle's reload licensing analyses identifies the limiting transient for that cycle.
~-. .
- r:-~ E For core flows less than rated, the MCPR Operating Limit estab- *
- lished in the specification is adjusted to provide protection of the MCPR Safety Limit in the event of an uncontrolled recircula-tion flow increase to the physical limit of pump flow. This protection is rovided for manual and automatic flow control by c cosing the MCPR operating 1m1t as e va ue rom Fig~Pe 3.§ 2~
Qeet 1 01 +.he J"ated C8f!e fl aw va 1 ua, *Ali el::leveF is gFeatercz. For Automatic Flow Control, in addition to protecting the MCPR Safety Limit during the flow run-up event, protection is pro-vided against violating the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow. This protection is provided by the reduced flow MCPR limits shown in~ Fi g~Fe 3. S 2 SAeet 2 1:lere tha cur e corres130AEli fig to the L
\1 1 11 *
- l . B 3/4. 5- ~7 -ke..
- Co,.e C}ero..117) .
Ji l 1
>f4*rs Ke'.crr * ..J-
\*li);f,'.~i~CT~~i\~~i~mriiAt~~~~~i~~~~(~~i~~~~.;
Insert E As described in Spec-ification 4.3.C.3 and the associated Bases, observed plant data or Technical Specification limits were used to determine the average scram performance used in the transient analyses for determining the MCPR Operating Limit. If the current cycle scram time performance falls outside of the distribution assumed in the analyses, an adjustment of the MCPR limit may be required to maintain margin to the MCPR Safety Limit during transients. Compliance with the assumed distribution and adjustment of the MCPR Operating Limit will be performed as directed by the nuclear fuel vendor in accordance with station procedures . i
- 3. 5. LIMITING CONDITION, FOR OPERATION BASES (Cont'd.)
ettFFe&t. rated flew MG~ limit is ~seEl (Hnear.iil:terpolatiou 0 eetweell the MCPR limit lines depicted is permissible). Tlfere-= fore, ieE l.ttttemetie Flew Geat.Fel, the KCPR 0peiating Limit is~ eBesea as the oalttc fzom FigHre 3.5 2 Sheet 1, Sheet 2 01 the e reteEl flew Valse J WBi~bEW8Ji' i& gi:eate&t !?: Analyses have demonstrated that transient events in Single Loop Operation are bounded by those at rated conditions; however, due to the increase in the MCPR fuel cladding integrity safety limit in Single Loop*Oper~tion, an:*equivalent adder must be.uniformly
~---- apphed to~~MCPR t.CO*to .maintain.the same margins to the MCPR fuel cladding integrity safety limit. *****
M. Flood Protection Condensate pump room flood protection will assure the availabil-ity of the containment cooling seI'Vice water system (CCSW) during a postulated incident of flooding in the turbine build-ing. The redundant level switches in the condenser pit will preclude any postulated flooding of the turbine building to an elevation above river water level. The level switches provide alarm and circulating water pump trip in the event a water.level is detected in the condenser pit. -.j
.j
- .1
.*J
*-~r
.\
- t
. *~ B 3/4.5-38
*.:i J**1
- ~
~
... --: .. -- .. ***- . . . . . -*. . ~ ----....:. . . --** .... ~:.:-:.-*-***-* *-::--- .- --~-**:t-- .. -*--:-:--*- - --~ --** .--- __, - . *---7""?--:-- ~-.- ... ...~.*~----: - * *--**--~ ... -- .... ---:--;:- ....,
DRESDEN II DPR-19 Amendment No. 82,84,95,104,107 4.5 SURVEILLANCE REQUIREMENT BASES (Cont'd.) evaluation of the average planar LHGR below this power level is not necessary. The daily requirement for calculating average planar LHGR above 25 per cent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes. J. Local Steady State LHGR Fbt..R.x~for all fuel shall be checked daily during reactor operation at greater than or equal to 25 per cent power to determine if fuel burnup or control rod movement has caused changes* in power distribution. A limiting LHGR value is pre-cluded by a considerable margin when employing acpermissible control rod pattern below 25% rated ~hermal power. K. local Transient LHGR The fuel design limiting ratio for centerline melt (FDLRC) shall be checked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution. The FDLRC limit is designed to protect against centerline melt-ing of the fuel during anticipated operational occurrences. L. Minimum Critical Power Ratio (MCPR) At core thermal power lev.els less than or equal to 25 percent,
- the reactor will be npP.rntino nt minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicates that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.
The daily requirement for calculating MCPR above 25 percent rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power -~ or control rod changes. ..J In addition, the reduced flow correction applied to the LCO
., provi~es margin for flow increase from low flows .
- 1
-~
M. Flood Protection The watertight bulkhead door and the penetration seals for pipes and cables penetrating the vault walls have been designed B 3/4.5-41
DRESDEN II DPR-19 Amendment No. 82, 85, 95, 104 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT (Cont 1 d.) (Cont'd.)
- e. The suction valve in the idle loop shall be closed and electrically isolated except when the idle lcop is being prepared for return to service; and
- f. If the tripped pump is out of service for more than 24 hours, imple-ment the following additional restrictions:
- i. The flow biased RBM Rod Block LSSS shall be reduced by 4.0%
(Specification 3.2.C.1); ii. The flow biased APRM Rod Block LSSS sha 11 be reduced by 3.5% (Specification 2.1.B); iii. The fl ow biased APRM scram LSSS shall be reduced by 3. 5% (Specification 2.1.A.1); iv. The MCPR Safety Limit shall be increased by 0.01 (Specification
- 1. 1. A); .,_, -Pl
- - - - - - - Jrettetl fJW
- v. The-i°MCPR Operating Limit shall be increased by 0.01 (Specification 3.5.~f; 3/4.6-15
DRESDEN-II DPR-19 ,_,_ ~* : ~* Amendment No. 82, 85, 95, 104 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE REQUIREMENT (Cont 1 d.) (Cont 1 d.) vi. The MAPLHGR Operating Limit
~ shall be reduced a:r?"°J"'"-re ~multipl ica-tive factor~ .f'ro'°" -;(e.. Co,.~ C>per-o...*n"'J t,,.,,,. qs Ke?orr
~(Specifica- .
tion 3.5.I). If, concurrently, one Automatic Pressure Relief Subsystem relief valve is out-of-service, the MAPLHGR Operat-
-rke_ ing Limit shall be t:Cpp~Nc...'re. ~multi-plicative factor"----+
ef 9. 89 fey:* 8x&e-1 to,.. -n...e.. cor-e.. OpeI-c...-r'; £-,,.,, *-n f?epo 17
- fuel aRd Q.76 fa~
9x9 f1.1e1-:C-
- 4. Core thermal power shall not exceed 25% of rated without forced recircu-1ation. If core thermal power is greater than 25%
of rated without forced recirculation, action shall be initiated within 15
**: . minutes to restore
. 1 operation to within the prescribed limits and core thermal power shall be returned to within the prescribed limit within two (2) hours. I. Snubbers (Shock* I. Snubbers (Shock) Suppressors) Suppressors) The following surveillance requirements apply to safety related snubbers .
- 3/4.6-16
;131\'J;{Qff:':'?J;'::f';""'r;,'(0'.;'.i'.~1~~0~~~~~I*"f;*li~!T'};C'"~::"B)~miji;;~§'k,'t~i¥?1"i?i0:f~~~~~~~
DRESDEN II DPR-19 Amendment No. 82, 85, 95, 104 3.6 LIMITING CONDITION FOR OPERATION BASES {Cont 1 d.} In addition, during the start-up of Dresden Unit 2, it was found that a flow mismatch between the two sets of jet pumps caused by a difference in recirculation loops could set up a vibration until a mismatch in speed of 27% occurred. The 10% and 15% speed mismatch restrictions provide additional margin before a pump vibration problem will occ~r. Reduced flow MCPR Operating Limits for Automatic Flow Control are not applicable for Single Loop Operation. Therefore, sustained reactor operation under such conditions is not permitted. Regions I and .II of Figure 3.6.2 represent the areas of the power/flow map with the least margin to stable operation. Although calculated decay ratios at the intersection of the natural circulation flow line and the APRM Rod Block line indicate that substantial margin exists to where unstable operation could be expected. Specifications 3.6;H.3.b,, 3.6.H.3.c. and 4.6.H.3. provide additional assurance that if unstable operation should occur, it will be detected and corrected in a timely manner. During the starting sequence of the inoperable recirculation pump, 1 restricting the operable recirculation pump speed below 65% of rated prevents possible damage to the jet pump riser braces due to excessive vibration. The closure of the suction valve in the idle loop prevents the loss of LPCI through the idle recirculation pump into the downcomer. Analyses have been performed which support indefinite operation in single loop provid~d the restrictions discussed in Specification 3.6.H.3.d. are implemented within 24 hours. The LSSSs are corrected to account for backflow through the idle jet pumps above 40% of rated-recirculation pump speed. This assures that the original drive flow biased rod block and scram trip settings are preserved during Single Loop Operation. The MCPR safety limit has been increased by 0.01 to account for core flow and TIP reading uncertainties which are used in the statistical analysis of the safety limit. In addition, the~CPR Operating Limit has also been increased by 0.01 to maintain the same margin to the safety limit as during Dual Loop Operation. J.. J ri
- h1~~w B 3/4.6-36
- )-
-~ -~
~1 -~- .-~. DRESDEN JI.:.~:. *_ DPR :
- _;I Amendment No. 82, 85, 95, 104
- d'
'l 1 3. 6 LIMITING CONDITION FOR OPERATION BASES (Cont'd) ~'f
)j ~y Tle_ moL-npL ~The decrease of the MAPLHGR Operat1ng l1miFte 91% ef its ePigiAalE::.. 1-h..c.TO,- +, 'C.C..T?..ie. "'iah1ec:-accou~ts for the ~ore rapid loss of c~re flow durir.g Single
,.i Co S?ec., ,e.cf *"'-;kl loop Operat1on than dur1ng Dual loop Operat1on. _j ,,. L ,., 'J i:-.,,_ ~'T"l"'O l
- G '"- 7ke... LOre. Ote.r""""'? ,...._,-n; K* r-1
' 1 "R~tr ' "rr.s The more conservative MAPLHGR reduction factors ef e. 89 f'er 8>t8 f'1:1el-E'-
_;l * .. arui Q.76 fer 9x9 ft1el=.are applied if one relief and one recirculation
*1 loop are inoperable at the same time. The small break lOCA is the j concern for one relief valve out-of-service; the large break LOCA is
- 1 the concern for Single loop Operation. Selecting the more conserva-
\ tive MAPLHGR multipliers will cover both the relief valve out-of-
~ service and Single loop Operation.
. 1 Specification 3.6.H.4 increased the margin of safety for thermal- .: ~
hydraulic stability and for startup of recirculation pumps from natural circulation conditions. I. Snubbers (Shock Suppressors) Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the prob-ability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation. Because the snubber protection is required only during low probability events, a period of 72 hours is allowed for repairs or replacements. In case a shutdown is required, the allowance of 36 hours to reach a cold shutdown condition will permit an orderly shutdown consistent with standard operating procedures. Since plant startup should not commence with knowingly defective safety related equipment, Specifi-cation 3.6.I.4 prohibits startup with inoperable snubbers. When a snubber is found inoperable, a review shall be performed to determine the snubber mode of failure. Results of the review shall be used to determine if an engineering evaluation of the safety-related system or component is necessary. The engineering evalua-tion shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the support component. or system.
- All safety related hydraulic snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures.
All safety related mechanical snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation and attachments to the piping and anchor for indication of damage or impaired operability. B 3/4.6-37
. ~ . - .. '
DRESDEN II DPR-19 Amendment* No. 82, 86, 97, 105
. 6. 0 ADMINISTRATIVE CONTROLS (Cont'd.)
* :~durJng the test program and a comparison of these va 1ues
***,~*witoc:design predictions and specifications. Any
~'"'~~~.corrective actions that were required to obtain
.~ sa~isfat::tory~operation shall also be described. Any
*.~ additional'specific details required in license conditions based on other commitments shall be included in this report.
- Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days f o11 owing resumption or commencement of commetci al p*ower. operation, or (3) 9.months followfng in.itial -*** ....
-.;;-~:.er.i.ticality, whichever is ear]iest. If the Startup Repo*rt .... ____ . _ ~-~--- . - ....... -,.
doe's riot cover all three events (i.e., initial criticality, completion of startup test program, and resumption or** commencement of commercial power operation), supplementary: reports shall be submitted at least every three months until all three events have been completed ..
. 2. A tabulation shall be submitted on an annual basis of the number of statioA, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions, (See Note); e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenahce (describe maintenance),
waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual. total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
- 3. Monthly Operating Report Routine reports of operating statistics and shutdown experienc.es shall be submitted on a monthly basis to the United States Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate Regional Administrator, to. arrive no later than the 15th of each month following
- r.:;. ~ the calendar month covered by the report.
Set-;- /- L_";> Note: This tabulation supplements the requirements of 20.407 of 10 CFR Part 20. 6-18 AMENDME~T N0. 105
- 'i
.:J *-. .. - *. -** *-- .
'.i Jr4;;~~01I?r~x;~::~r:~st0l~~~~~~~!f§t,71JJ~F1.i&{~~~'B~':r0~~1~~~1Mi~~~:.~::~~~:t1~~~rr~i;f~1~~0:~'5.c~rn~;*:;~~~fi;)~'t-ir
Insert F Core Operating Limits Report
- a. Core operating limits shall be established and documented in the Core Operating Limits Report before each reload cycle or any remaining part of a reload cycle for the following:
I) The Minimum Critical Power Ratio Safety Limit and Single Loop Operation adjustments for Specifications I.I.A and 3.6.H.3.f.
- 2) The Control Rod Withdrawal Block Instrumentation for Table 3.2-3 of Specification 3.2.C.
- 3) The Average Planar Linear Heat Generation Rate (APLHGR) Limit and associated APLHGR multipliers for Specifications 3.5.I, 3.5.D.2, and 3.6.H.3.f.
- 4) The Local Steady State Linear Heat Generation Rate (LHGR) for Specification 3.5.J.
- 5) The Local Transient Linear Heat *Generation Rate
. (LHGR) for Specification 3.5.K.
- 6) . The Minimum Critical Power Operating Limit for Specifications 3.5.L and 3.6.H.3.f. This includes rated and off-rated flow conditions, and single loop operation.
- b. The analytical methods used to determine the core operating limits shall be those ANF methods previously reviewed and approved by the NRC. The topical reports describing the base methodology used by ANF for Dresden Unit 2 and approved by the NRC are listed in References 1 through 6.
- c. The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- d. The Core Operating Limits Report, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
"R~1;;;;:.~~----------
I. XN-NF-5I2(P)(A), "XN-3 Critical Power Correlation", (latest approved revision).
- s. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel", (latest approved rev_ision).
*--~:......
- 6. XN-NF-81~2~(P)(A),. ~Generic Statistical Uncertainty Ati~lys*is.0 Methodology"; (latest approved revision).
. -~. 1.
*.-:i *./.
I
- DRESDEN STATION UNIT 3 TECHNICAL SPECIFICATION REVISIONS DPR-25 License Page 3 Appendix A Pages viii
- 1. 0-1 1.0-5 B 1/2.1-6 B 1/2.1-7 B l /2. 1-8 B 1/2.1-10 B 1/2.1-11 B 1/2.2-4 3/4.2-12 3/4.3-11
- B 3/4.3-16 B 3/4.3-17 B 3/4.3-19 B 3/4.3-20 3/4.5-9 3/4.5-15 3/4,!j-l()
3/4.5-17 3/4.5-18 3/4.5-19 3/4.5-20 3/4.5-21 3/4.5-22 3/4.5-25 3/4.5-26 3/4.5-27 B 3/4.5-36 B 3/4.5-37 B 3/4.5-38 B 3/4.5-41 3/4.6-15 3/4.6-16 B 3/4.6-36 B 3/4.6-37
. 6-18 0334T:15
- ,~j'"':~:z:::::.::~}i";:;:;~--?~:";"~'!::~~=~::;,'::=-~b_~r~z,~""'~::::~~'.~~~~=:~~~rn~~'3~~~~~~~~~~~+/-&~~==~,~~~~-~i~;i:;;tt~~:'.:;_:~:'.~:~~:2~;\(~~:~~=~
- ,]
..1 DPR-25 Am. 31 2. B. Pur1uant to the Act and 10 CFi Parta 30 and 70, to 1/30178 poa1e11, but not 1epar1te, 1uch byproduct and 1pecial nucl,at;'.. materiah u may be produced by the operation of Dresden Nuclear Power Station, Unita Noa. 1, 2 and 3.
- 3. Thia licen1e 1hall be deemed to contain and i1 1ubject to the condition1 specified in the following Commlaaio~.
regulationa; 10 CFR Part 20, section 30.34 of 10 era Part 30, Section 40.41 of 10 CPR Part 40, Sectiona 50.54 and S0.59 of 10 CPR Part 50, and Section 70.32 of 10 CPR Part 70; and la 1ubject to all applicable provi1ion1 of the Act and to the rule1, regulation* and orders of the Commha fon now or hereafter in effect; and l1 1ubjact to the additional condition1 1pecified below: A. Maximum Power Level Am. 1 3/2/71 Commonwealth Edi1on l1 authorized to operate the DRL facility at 1teady 1tat1 power lav1l1 not in e1ce11 of Authorized 2527 megawatt.SI (thermal), except that Commonwealth 6/28/71 Bdhon *hall not operate the facility at power**leveh in e1ce11 of five (5) megawatt1 (thermal), until 1ati1factory completion of modifications and final > te1ting of the station output tranaformer, the >* auto-depre11urizatioa interlock, and the f11dwater 1y1t1m, a1 deacribed in Commonwealth ldi1on 1 1 t1legram1 dated February 26, 1971, have been verified in writing by th* Commi11ioa. B. Technical Specification* - .-..;.:;.. *;:-1***- .... __'!'ha Technical Specificatioa1 contained in Appendix A 11 ravl1ed through Jmendiientlfi)2)are hereby incorporated in the licon*** ~. llcauaaa *hall , operate the facilitJ in accordance with the Technical Specificationa *
. ~{ c. lepo1"tl i
.! Coanonwealth ldi1on 1hall make certain report1 in
' accordance with the requirement1 of the Technical Speciflcation1.
D. Record* Commonwealth Edi1on 1hall keep facilitJ opo~ating records in accordance with the requiremeat1.of th* Technical Specificatlon1. Amendment' No.
., 5039N
' 84030
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*}j Figure 2.1-3 -*~j APRM Bias Scram Relationship to Normal Operating Conditions J.~
B 1/2.1-17 Figure 4.1.1 -~ Graphical Aid in the Selection of an Adequate Interval Between Tests B 3/4.l-i8 Figure 4.2.2 Test Interval vs. System Unavailability B 3/4.2-38 Figure 3.4.1 Standby Liquid Control Solution Requirements . 3/4.4-4 Figure 3. 4_. 2 Sodium Pentaborate Solution Temperature Requirements 3/4.4-5 MAPLHGR Limit vs Bundle Average Exposure ANF 8x8 Fuel (Sheet 1 of 2) MAPLHGR Limit vs Bundle Average Exposure ANF 9x9 Fuel (Sheet 2 of 2) 3/4.5-19 Figure 3.5-lA 3/4.5-20 Figure 3.5-lB 3/4.5-21 Figure 3.5-2 MCPR LI for Reduced Total Flo heet 1 of 3) 3/4.5-25 BxB MCPR Operating Limit for Automatic Flow Control (Sheet 2 of 3) 9x9 MCPR Operating Limit for Automatic Flow Control (Sheet 3 of 3) Minimum Temperature Requirements per Appendix G of 10 CFR 50 3/4.6-23 Figure 3.6.2 Thermal Power vs Core Flow Limits for Thermal Hydraulic Stability Surveillance In Single Loop Operation .3/4.6-24 Figure 4. 6.1 Minimum Reactor Pressurization Temperature B 3/4.6-29 Figure 4.6.2 Chloride Stress Corrosion Test Results at 500°F 8 3/4.6-31 Figure 4.8-1 Owner Controlled/Unrestricted Area Boundary 8 3/4.8-38 Figure 4. 8-2 . Detail of Central Complex 8 3/4.8-39 Figure 6.1-1 Offsite Organization - Deleted Figure 6.1-2 Station Organization - Deleted Amendment No. l 00 viii
- j .- . . .. . . . .. . . . - *-* ---* -
DRESDEN III ,h _QPR-25 Amendment No. 1 ~, 93, 75 1.0 DEFINITIONS The succeeding frequently used tenns are explicitly defined so that a unifonn interpretation of the specifications may be achieved. A. (Deleted) --e___ A-%. Alteration of the Reactor Core - The act of moving any . component in the region above the core support plate; below the upper grid and within the shroud. Nonnal control rod movement with the control rod drive hydraulic system is not defined as a core alteration.
- Critical Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of th"~correlation. -{Reference XN-NF-512).-'?_
D. Hot Standby - Hot standby means operation with the reactor cr1t1cal, system pressure less than 600 psig, and the main steam isolation valves closed. E. Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action. F. Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including _ actuation, alann, or trip. Response time is not part of the routine instrument calibration, but will be checked once per cycle. G. Instrument Functional Test - An instrument functional test means the inJect1on of a simulated signal into the instrument primary sensor to verify the proper instrument response alann, and/or initiating action.
- H. Instrument Check - An instrument check is qualitative detenn1nat1on of acceptable operability by observation of J
~ ~
instrument behavior during operation. This detennination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable
- I. Limiting Conditions for Operation (LCO) - The limiting cond1t1ons for operation specify the minimum acceptable levels of system perfonnance necessary to assure safe startup and i
~ operation of the facility. When these conditions are met, the 1.0-1
Insert A B. Core Operating Limits Report CCOLR) - The Core Operating Limits Report is the unit specific document that provides core operating limits for the current operating reload cycle. These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.A.4. Plant operation within these operating limits is addressed in individual specifications.
.. .. ' * . . ~*-
-:.:~; **-
DRESDEN III DPR-25 Amendment No. 77, 87, 94
- 1.0 DEFINITIONS (Cont'd.)
AA. Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alternations are being performed. When the mode switch is placed in the shutdown position a reactor scram is initiated, power to the control rod drives is removed, and the reactor protection system trip systems are de-energized.
- 1. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212°F.
- 2. Cold Shutdown means~conditions as above with reactor coolant temperature equal to or less than 212°F.
BB. Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question. 0 CC. Surveillance Interval - Each surveillance requirement shall be performed within the specified surveillance interval with: **
- a. A maximum allowable extension not to exceed 25% of the surveillance interval.
- b. A total maximum combined interval time for any 3 consecutive intervals not to exceed 3.25 times the specified surveillance interval .
DD. Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to rated thermal power of 2527 Mwlli. EE. Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occ~r intermittently with neither type. being completely stable. FF. Fu~l Design Limiting Ratio ~~DLRX) - The fuel design limiting ratio fsr ExxG~ fye~..rfs the limit used to assure that the fuel operates within the end-of-life steady state design criteria. FDLRX assures acceptable end-of-life conditions by, among other items, limiting the release of fission gas to the cladding plenum. GG. Dose Equivalent I-131 - That concentration of I-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose
- conversion factors used for this calculation shall be those listed in Table III of TID-14844, 11 Calculation of Distance Factors for Power and Test Reactor Sites. 11
- 1. 0-5
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DRESDEN**llL';::... ,** DPR*25.
...... *... :-:*;*:: *. :~*.. *:*** '.. .. -* ... *-- ..
Amendment.No. ~'rs, 75 1.1 SAFETY LIMIT BASES j FUEL CLADDING INTEGRITY ..'1
';-~.
The fuel cladding integrity limit is set such that no calculated fuel damages would occur as a result of an abnormal operational transient. Because fuel dainage* is not directly observable, a step-back approach is used to establish a Safety Limit such that the mini mum cri ti ca1 power ratio (MCPR )* i s no 1ess than the MCPR fuel cladding integrity safety limit. MCPR greater than tha MCPR fuel cladding integrity safety limit represents a conservative* margin relative to the conditions required to maintain fuel cladding integrity by assuring that the fuel does not experience transition boiling. The fuel cladding is one of the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosions or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thennal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation is just as measurable as that from use r~lated cracking, the thennally caused cladding perforation signals a
*threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with margin to the conditions which would produce onset of transition boiling, (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fue 1 cladding integrity Safety Limit assures that duri tig nonna 1 operation and during anticipated operational occurrences, at least 99.9i of the fuel rods in the core do not experience transition boiling. -see ~efeFeAee XN NF 524.E A. Reactor Pressure greater than 800 psig and Core Flow greater than 1oi of Rated Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boiling transition, is not a directly observable parameter in an qperating reactor.
Therefore, the margin to boiling tra~sition is calculated from pl ant operating parameters such as co.re power, core fl ow, feedwater temperature, and core powe~ distribution. The margin for each fuel assembly is characterized by the critical B 1/2.1-6
\\ I I
DRESDEN III DPR-25 Amendment No. 75, 87, 94 1.1 SAFETY LIMIT BASES (Cont'd.) power ratio (CPR) which is the ratio of the bundle power which would produce the onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the Minimum Criti~al Power Ratio (MCPR). It is assumed that the plant operation is controlled to the I nominal protective setpoints via the instrumented variables. (Figure 2.1-3). The MCPR Fuel Cladding Integrity Safety Limit assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for"()peration, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (MCPR=l.00) and the MCPR Fuel Cladding Integrity Safety Limit is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state. One specific uncertainty include~ .. i~~he safety limit is the JANr t-!KC.-a.pp,....""-J. uncertainty inherent in the ~'t:':ritical power correlation~ Refer to1CMN tff 524'2-for the methodology used in determining the MCPR Fuel Cladding Integrity Safety Limit.
~L-~~ue.d ~ritical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because boundingly high radial power *peaking factors and boundingly flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by AN'F NRC.-c:pprciJe..CI '\.... the tendency of the,.~correlation to overpredict the number of rods in boiling transition. These conservatisms and the Af'IF N'RC..-c:;prro~ <. inherent.accuracy of th~correlation provide a reasonable degree of assurance that during sustained operation at the MCPR Fuel Cladding Integrity Safety Limit.there would be no transition boiling in the core. If boiling transition were to occur, however, there is reason to believe that the integrity of the fuel would not necessarily be compromised. Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach; much of the data indicates that LWR fuel can survive for an extended period in an env1ronment of transit iOn boiling. . . ... . .
~--
During Single Loop Operation, the MCPR safety limit is increased by 0.01 to conservatively account for increased uncertainties in the core flow and TIP measurements. B 1/2.1-7
DRESDEN III DPR-25 Amendment No. 42, 63, 75, 94 1.1 SAFETY LIMIT BASES (Cont'd.)
.:) If t~e r~a~tor pressure shou~d.ever:~xc~~d t~e li~it of ANF ~"RC.*ei:ppr-oo.Je..cl j ik ANF=' ~'l "'1e.~o mc.-<<f'PtooJe41 app~1cab~l1ty of the~ r1t1cal ower *correlati-on as defined lr'!f\lU. tff Slt°f1t wou- e assumedtt'!at the MCPR Fuel
- -1 ~Y ~:.-red Cladding Integrity Safety Limit had been-violated. This
.;_\
'.j "'* Spec1~c ...r1.o" "*"*"*'I appl icabi 1ity pressure 1 imit is higher than the pre? sure safety limit specified in Specification 1.2.
*.1 cJ
-I
.. I ~01 fuel fa~Fieated by Advaneed NueleaF Fuels CerpeFation 2
"~-! (ANP~!uel design criteria have been established to provide protecti'on against fuel centerline melting and 1% plastic J .j cladding strain during transient overpQ'#eT' conditions
-,.: throughout the life of the fue 1. To demonstrate comp 1 i ance 1 with these ~riteria, fuel ~~d tent~~li~e-temperatures are
-i determined at 120% overpower conditions as a check against l
calculated centerline melt temperatures. FDLR~ is incorporated to protect the above criteria at all power levels considering events which will cause the reactor power to increase to 120% of rated thermal p~wer.
- 8. Core Thermal Power Limit (Reactor Pressure less than 800 psia)
At pressures below 800 psia, the core elevation pressure drop
! (O power, 0 flow) is greater than 4.56 psi. At_ low powers and
\
I flows this pressure differential is maintained in the bypass _, region of the core. Since the.pressure drop in the bypass i region is essentially all elevation head, the core pressure drop at low powers and flows will always be 93eater than 4.56 psi. Analyses show that with a flow of 28xl0 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow i 1
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I 8 1/2.1-8
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~~H DRESDEN III DPR-25 Amendment No. 75
*--~* ...-*-**
"l 1.1 SAFETY LIMIT BASES (Cont'd.)
available for any scram analysis, Specification 1.1.C.2 will be relied on to detennine if a safety limit has been violated. During periods when the reactor is shut down, consideration must also be given to water level requirements* due to the. effect of decay heat. If reactor water 1evel should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reductio~ in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to . prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel* provides adequate margin. This level **will be continuously monitored whenever the recirculatiqn pumps are not operating.*
*Top of active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
I
- 2.1 LIMITING SAFETY SYSTEM SETTING BASES FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the I
.j units have been analyzed throughout the* spectrum of planned **! operating conditions up to the rated thermal power condition of I
- I
~l, 2527 MWt. In addition, 2527 MWt is the 1icensed maximum steady- ) -rle.. AMF' N'Kc.. *c:..i"?N~J state power level of the units. This maximum steady-state power ] :_e..~oJoLo 1y L,flnz.ci 1evel will never knowingly be exceeded. Se~R-eference XH*NF*79*11~
1
-:1 Spu,f,ce.n~ fo.f.,l't,"J l Conservatism is incorporated into the transient analyses which
! define the tcPR operating limits. Variables which inherently
] possess little or no uncertainty or whose uncertainty has little i or no effect on the outcome of the limiting transient are selected
- ~ at bounding values. Variables which possess significant
, uncertainty that may have undesirable effects on thennal margins Jit.e. ANr N~l*Af>P~~j* are addressed statistically. Statistical methods used in the
~ L transient analyse~ are described i °"XH >IF' a~ 22 f The MCPR 0
Me.: DtJy Li'f>R.d '" operating 1imi ts are established such that the occurrence of the
~ec:.,.C,CQ:J"lo"" '*'*A*'I limiting transient will not result in the violation of the ~CPR Fuel Cladding Integrity Safety Limit in at least 95% of the random statistical combinations of uncertainties. In general, the variables with the greatest statistical significance to the consequences of anticipated operational occurrences are the reactivity feedback associated with the formation and removal of coolant voids and the timing of the control rod scram.
B 1/2.1-10 "~
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DRESDEN III DPR-25 Amendment llo; ~ 87 2.1 LIMITING SAFETY SYSTEM SETTING BASES (Cont'd.) stead7-state: operation without forced recirculation will not be permitted, ezcept during startup testing. The analysis to support operation at various power and "flow relationships has considered
*operation with either one or ~~,;ecirculation pumps.
The bases for individual trip settings are discussed in the fol-lowing paragraphs. For analyses of the thermal consequences of the transients, the MCPR's stated in paragraph 3.5.1:~ -Z..S.L. limiting condition of operation bound those which are conser-vatively assumed to ezist prior to initiation of the transients. A. Neutron Flux Trip Settings
- 1. APRM Flux Scram Trip Setting (Run Mode)
*The average power-range-monitoring (APR!!) system, which is
~] calibrated using beat balance data taken during steady-
-.i
*)
1tate conditions, reads in percent of rated thermal power.
.-~
Because fission chambers provide the basic input signals,
'I
- the APR!! s7stem responds -directly to average neutron fluz.
~~
During transients, the instantaneous rate of beat transfer from the fuel (reactor thermal power) is less than the 1.
-. instantaneous neutron flux due to the time constant of the
*"1 fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will- be less than that
.1' *. ' indicated by the neutron fluz at the scram setting.
_,I
-1 - ,' Analyses demonstrate that, with a 120 percent scram trip -1 I
1etting during dual loop operation.or 116.S percent during l
) single loop operation, none of the abnormal operational
-, transients analyzed violate the fuel Safety Limit and
.*.:J there is a substantial margin from fuel damage *
.,1 Therefore, the use of flow referenced scram trip provides even additional margin.
- i
~j . ~
An .increase in the APR!! scram trip setting would decrease the margin present before the fuel cladding integrity J ' Safety Limit is reached. The APR!! scram trip setting wa~ determined by-an analysis of margins required to provide a
*~ . .'~ reasonable range for maneuvering during operation
- I Reducing this operating margin would increase the frequenc7 of spurious scrams which have an adverse effect
- i on reactor safety because of the resulting thermal l
*1' 1tresses. thus,- the APR!! scram trip setting was selected
- i because it provides adequate margin for the fuel cladding l integrity Safety Limit yet allows operating margin that
*i_, reduces the possibility of un.necessary scrams. **
B 1/2.1-1.1 3892a
- 1 3122A i
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.j 2.2 LIMITING SAFETY SYSTEM SETTING BASES DRESDEN III A~ *aPR-25 Amendment No. 7'* 9'3* 75 .*1 -~
In compliance with Section III of the ASME Code, the safety valves
".j must be set to open at no higher than 103i of design pressure, and they must limit the reactor pressure to no more than llOi of design pressure. Both the neutron flux scram and safety valve actuation are required to.prevent overpressurizing the reactor pressure vessel and thus exceeding the pressure safety limi~. The pressure scram is available as a backup protection to the direct valve position trip scrams and the high flux scram.
If the high flux scram were to fail, a high pressure scram would occur at 1060 psig. Analyses are perfonned as described in
~ PefeFeAee XN Nf-79=71-L.for each reload to assure that the pressure
\,__f'7k,._ AN~ safety lfmft fs not exceeded.
[ ~ AN F' M~c. -a:ppf'r>Ve.d
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B 1/2.2-4
DRESDEN III DPR-25 Amendment No. 77, 87, 94 "i Table 3.2.3
.:1 INSTRUMENTATION THAT INITIATES ROD BLOCK Minimum No. of Operable Inst.
Channels Per Trip System (1) Instrument Trip Level Setting 1 APRM !.!pscale (flow bias) (7) Dual Leep Operation Less than or equal to (.58 WB plus 50)/FDLRC' (See N te 2) : Single Loop Operation Less than or equal to (.58 WB plus 46;5)/FDLRC (See N te 2) . 1 APRM upscale (refuel and Less than or equal to j.
. f
-l
'j
~
2 1
- Startup/Hot Standby mode)
APRM downscale (7) Rod block monitor upscale (flow bias) (7) 12/125 full scale Greater than or equal to 3/125 full scale
! Dual Loop Operation Single Loop Operation 1 Rod block monitor downscale (7) 3 IRM downscale (3) Greater than or equal to 5/125 full scale 3 IRM upscale Less than or equal to 108/125 full ~cale 3 IRM detector not fully N/A inserted in the core 2 (5) SRM detector not in (See Note 4) startup position
.1.
.* ~
2 (5) (6) 1 SRM upscale Scram discharge volume water level - high Less 5than or equal to 10
- counts/sec .
Less than or equal to 25 gallons
*i Notes: (See Next Page) 3/4.2-12 \ *1
DRESDEN III gPR-?5 Amendment No. yi, 9'3, yo, 75
- 3. 3 LIMITING CONDITION-FOR OPERATION 4.3 SURVEILLANCE REQUIREMENT (Cont'd.) (Cont'd.)
- 2. The maximum scram 2. At 16 week intervals,********
insertion time for at least soi of the con-9°' insertion of any trol rod drives shall be operable control rod tested as in 4.3.C. l so r shall not exceed 7.00 that every 32 weeks all seconds. of the control-rods-shall have been tested. When-ever 5°' or more of the control ro.d drives have .. 1 been tested, an
'
- evaluation shall be made to provide reasonable assurance that proper control rod drive performance is being maintained.
- 3. Following completion of each set of scram testing as described above, the results will be compared against the average scram speed distribution used in i
l the transient analysis J to verify the * 'i
; applicability of the*
current MCPR Operating Limit. Refer to Specification 9.5.Kf'"
- 3. S'* .L.
D. Control Rod Accumulators D. Control Rod Accumulators At all reactor operating Once a shift check the pressures, a rod accumulator status of the pressure may be inoperable provided and .level alarms for each that no other control rod accumulator. in the nine-rod square array around this rod has a:
- 1. Inoperable accumulator,
- 2. Directional control
- 1 valve electrically disarmed while in a
' non-fully inserted position.
3/4.3-11
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)
'=:*.j DRESDEN III _QPR-25 Amendment No. 90, yJ, 75 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
- 3. The operability of the scram discharge volume vent and drain valves assures the proper venting and draining of the volume. This ensures that water accumulation does not occur which would cause an early tennination of control
- ~
rod movement during a full core scram. These specifications provide for the periodic verification that j~ . the valves are open and for testing of these valves under
--.d reactor scram conditions during each Refueling Outage.
- ~
i) B. Control Rod Withdrawal
*~.;1 -rle. ANr "fRC.* ¢ppf'o"j. 1. Control rod dropout accf dents as discussed in *R.efeF~Aee e.-
- l .Me;r~.cdo~y L, n-1, * ~XN=NF 80 19 1 Vel. 1,'<.ean lead to significant core damage.
~l s .+. J :. "' If coupling integrity is maintained, the possibility of a d "Pee. *c~T"' "' 0 '* '*"* Lf.,, rod dropout accident f s elf mi nated.. The overtravel j position feature provides a posi tf ve check as only
' .j uncoupled drives may reach this position. Neutron
~ instrumentation response to rod movement provides a
.'J. verification that the rod is following its drive. Absence 4 of such response to drive movement would provide cause for
- 1 suspecting a rod to be uncoupled and stuck. Restricting
- 1 recoupling verifications to power levels above 2oi
-.*~.-_l.*.*.****-., pr~ifdiest~ssuranc ve11 ca ion wou 1 ed no thattresu a ro1td .1dnropa dudridng ro ropa ace recioudplting en *
*:1 ......
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- j
- 1 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extreme*1y remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a nonnal single withdrawal increment, will not contribute to a~y damage to the primary coolant system. The design basis is given in Section 6.6.1 of the SAR, and the design eval~ation is given in Section 6.6.3. This support is not required ff the reactor coolant system is .at atmospheric pressure since there would then be no driving force to rapidly eje~t a drive housing. Additionally, th~ support is. not required if all control rods are fully inserted and if an
-'i' **.1 I adequate shutdown margin with one control rod withdrawn has been demonstrated since the reactor would remain sub-critical even in the event of complete ejection of the strongest control rod."
- 3. Control rod withdrawal and insertion sequences are established to assure that the ma~imum insequence individual control rod or control rod saqY&R,e&4which are
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)
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-.. :~..:.~::...::.:~.:-!<.~~:t~"~"--::-..~*--:~~~~~':'it°'~::.s.:~*~~-:::=~~~~.t:t"i~.-n~.£.=.i--~,;i~-J.i~0~'3..-*~~":--....i:~..'i...W1i\t~*.,.~;0.i:~Ji!~~;;,,...:;~~~-;i~-:~~:;:;-~:t:.~dltS:~.;-l~-t.:f:.':'ilS~-~~\~~.+/-.{:\:,j~'f~~..:;;£~J~~j~:;_\3~~11;.~:::;..
- DRESDEN I II . . .. DPR-25 .
Amendment No. ~, ffe{J, ~, 75
- .A
- ~J 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.}
.\ withdrawn could not be worth enough to cause the rod drop ~l accident design 1 imi t of 280 cal /gm to be exceeded if they
)1 were to drop out of the core in the manner defined for the
,, Rod Drop Accident. These sequences are developed prior to )] initial operation of the unit following any refueling
- .*~ outage and the requirement that an operator follow these
.j sequences is backed up by the operation of the RWM or a
.:J second qualified station employee. These sequences are ii developed to limit reactivity worths of control rods and, *:1 together with the integral rod velocity limiters and the *1 action of the control rod drive system, 1imit potential ~) reactivity insertion such that the results of a control j rod drop accident wi 11 *not exceed a inaxi mum fue 1 energy.
- l content of 280 ca 1I gm. The peak fue 1 entha1py of 280 *
~i cal/gm is below the energy content, 425 cal/gm, at which 'i
- rapid fuel dispersal and primary system damage have been
~ A)JF' N'R ti
- i ] found to occur based on experimental data as is 'discussed
.i IVI 1 ,
< c.-t2f7P"e"e."- ~in~efereAee XN NF 80*19, Volume l .c_
1 e.'n\octoloo l *- /
- J J 'f l.;)rea I'\..
- .: ~,f. c.Q.,.., , , A The analysis of the control rod drop acci de.nt was j
- C)"" * * *'f~ originally presented in Sections 7.9.3, 14.2.1.2 and
- J
- 14. 2. 1. 4 of the Safety Analysis Report. Improvements in
,j analytical capability have al lowed a more refined analysis
-~ of the control rod drop accident. Parametric Control Rod Drop Accident analyses have shown that for wide ranges of key reactor parameters (which envelope the operating* ranges of these variables}, the fuel enthalpy rise during a postulated control rod drop accident remains considerably lower than-the 280 cal/gm limit. For each .operating cycle, cycle-specific j parameters such as maximum control rod worth, Doppler
] coefficient effective delayed neutron fraction and maximum four-bundle local peaking factor are compared with the l
results of the parametric analyses to detennine the peak ",J _ l fuel rod enthalpy rise. This value is then compared
.-~
against the Technical Specification limit of 280 cal/gm to
~i demonstrat~ compliance for each operating cycle.
specific values of the above parameters are outside the If cycle range assumed in the parametric analyses, an extension of the analysis or a cycle specific analys1s may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of
;* . the methodology for perfonning the Control Rod Drop
'; Accident analysis are provided in refeFeflee XH NF=80*19, e..
~e. A'NF 1\JRc -a.1"'P 0 JI.el Yme l ( Sblpfll emeRts l aAa 2) .~
)
*i
- l Ple."r4odolo? y
' r
~re.ii ~ou
-pec.,..,,c.;.ne., '* 6, 11, 'l 7 B 3/4. 3.. 17.
,1
- i='~"'""""'"""'""""""'"""'""""""'"""'"'""""*~*"""'"='"'""--'""'""~'"""'=""-"""'""",,."""""""""'"""""'-'-
- ,~
.ii -~~ j c;~ DRESDEN III Ah DPR-25 Amendment No. 7 ~, ~' 75 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.} maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator who withdraws rods according to a written sequence. The specified restrictions with one.channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists. Pmendments 17/18 and 19/20 present the results of an evaluation of a rod block monitor failure. These amendments show that during reactor operation with certain limiting control rod pattern, the withdrawal of a designated single control rod could result tn one or more fuel rods with MCPRS less than the MCPR fuel cladding integrity safety limit. During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur. It is the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop-due to the occurrence Qf inoperable control rods in other than limiting patterns. _C. Scram Inser:ti on Times The perfonnance of the control rod insertion system is
.*j analyzed to verify the system's ability to bring the reactor subcritical at a rate fast enough to prevent violation of the MCPR Fuel Cladding Integrity Safety Limit and thereby avoid fuel damage. The analyses demonstrate that if the reactor is operated within the limitations set in Specification 3.5.K~...-'3 . .5".L.
the negative reactivity insertion rates associated with the
..
- observed scram perfonnance (as adjusted for statistical variation in the observed data} result in protection of the MCPR safety limit'.
In the analytical treatment of most transients, 290 . milliseconds ~re allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This
- is adequate and conservative when compared to the typically observed time delay of about 210 milliseconds. Approximately 90 mi.11 i seconds after neutron flux reaches the trip point, the pilot scram valve solenoid de-energizes and 120 milliseconds later the control rod mo~ion is.estimated to actually begin.
However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time interval in the transient B 3/4.3-19
DRESDEN III DPR-25
..J.*.~~-*'"
Amendment No * }'7, .'2' rs, 75 3.3 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
*analyses, and is also included in the allowable scram insertion times specified in Specification 3.3.C. In the statistical treatment of the limiting transients, a statistical distribution of total scram delay f s used rather than ~he bounding value described above.
The perfonnance of the individual control rod drives is monitored to assure that scram perfonnance is not degraded. Fifty percent of the control rod drives in the reactor are tested every sixteen weeks to verify adequate perfonnance. serve p an ata were use o e enn1ne e average scram
*rep~ w,rk_:* .* pe nnance used in the transient analyses, and the results I11.se" . C1. each s of control rod scram tests during the current c e are compa against earlier results to verify that
"----~ perfonnance o he control rod insertion system h not changed signific ly. If an individual test group of tests should be dete ined to fall o~tside the statistical population defining the ram perfonnan characteristics used in the transient analyses, e-det ination of thennal margin requirements is underta as required by Specification 3.5.K) unless can e shown that the number of individual drives fallfn tside the tistical population defining the nominal fonnance is less an the allowable number of inoperab control rod drives. I e number of statistically rrant drives falls within this itation, operation be a1lowed to continue w1thout redet ination of the margin requirements provided the identified aber nt ~rives are fully inser,ted into the core and nergized in the manner of an f-noperable rod drive.
The scram times for all control rods are measured at the time of each refueling outage. Experience with the plant has shown that control drive insertion times vary little through the operating cycl~; hence no reassessment of thermal margin
.,)
l
' requirements is expected under nonnal conditions. The history of drive performance accumulated to date indicates that the 90",g insertion times of new and overhauled drives ap'proximate a nonnal distribution about the mean wnich tends to become skewed toward longer scram times as operating *time is accumulated. The probability of a drive not exceeding the mean 9~ insertion time by 0.75 second is greater than 0.999 for a normal distribution.
i * ..
.,'f B 3/4.3 .:; . - -- . .... ---*-----
. :*. ~: *_: T"'J~.:.::~;:~~'.:Sr~2~:~1:r~~\I~~!Jlif@J:;~~~~§1f3~:;;~:.;:}~.f~~~i~*~5mSff~;!f.0Fl:~:;W;:WWf~l$~%~i~~i~1i'7'.1'.~0~f(;,~'Q)Gr~3~10~{~jm 1 0
Insert c, Continued
- 2. XN-NF-524 (P) (A), ,}'.E-xxon Nuclear Critical Power Methodology for Boiling Water Reactors", (latest approved revision). *
- 3. XN-NF-79-7l(P), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors", (latest revision).
- 4. XN-NF-80-19(P)(A), "Exxon Nuclear Methdology for Boiling Water Reactors", (latest approved revision).
- s. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel", (latest approved revision).
- 6. XN-NF-81-22(P)f~), "Generic Statistical Uncertainty Analysis Methodolog.y", (latest approved revision)~ - *
.. ~ i -. {
~
j
-:j
,f
- I
*.~
Insert c
- 4. Core Operating Limits Report
- a. Core operating limits shall be established and documented in the Core Operating Limits Report before each reload cycle or any remaining part of a reload cycle for the following:
I) The Minimum Critical Power Ratio Safety Limit and Single Loop Operation adjustments for Specifications I.I.A and 3.6.H.3.f.
- 2) The Control Rod Withdrawal Block Instrumentation for Table 3.2-3 of Specification 3.2.C.
- 3) The Average Planar Linear Heat Generation Rate (APLHGR) Limit and associated APLHGR multipliefi fa~
Specifications 3.5.I, 3.5.D.2, and 3:6.H.3.f.
- 4) The Local Steady State Linear Heat *Generation Rate (LHGR) for Specification 3.5.J.
- 5) The Local Transient Linear Heat Generation Rate (LHGR) for Specification 3.5.K.
- 6) The Minimum-Critical Power Operating Limit for Specifications 3.5.L and 3.6.H.3.f. This in~ludes rated and off-rated flow conditions, and single loop operation.
- b. The analytical methods used to determine the core operating limits shall be those ANF methods previously reviewed and approved by the NRC. The topical report describing the base-methodology used by ANF for Dresden Unit 3 and approved by the NRC are listed in References I through 6.
- c. The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear .* -:=:.*.
limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- d. The Core Operating Limits deport, including any mid-cycle
- 1
.*j revisions or supplements thereto, shall be provided upon
.*i issuarlce to the NRC Document Control Desk with copies to
-.,.,------------::: R~g fon al Admi;n is t ra tor and Resident Inspector.
l"\e..fc.rc.Ace.s \ I I. XN-NF~SI2(P)(A),/"XN-3 Critical Power Correlation", (lates~ approved revision).
; I .
I
Insert Cl Observed plant data-or Technical Specification limits (Specification 3.3.C) were used to determine the average scram performance used in the transient analyses, and the results of each set of control rod scram tests performed per Specification 3.3.C during the current cycle are compared against earlier results to verify that the performance of the control rod insertion system has not changed significantly. If a test performed per Specification 3.3.C should be determined to fall outside of the statistical population defining the scram performance characteristics used in the transient analyses, a re-determination of thermal margin requirements is undertaken as required by Specification 3.5.L. A smaller test sample than that required by Specification 3.3.C is not statistically significant anQ should not be used in the re-determination of thermal margins. Control rod drives with excessive scram times can be fully inserted into the core and deenergized in the manner of an inoperable rod drive provided the allowable number of inoperable control rod drives is not exceeded. In this case, the scram speed of the drive shall not be *used as a basis in the re-determination of thermal margin requirements. ,l l 1 'l
- 1 i
*~
DRESDEN III DPR-25 Amendment No. 19,28,40,63,75,94,102 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) (Cont'd.)
."".>~'!""op~rati on is permi s-
,.,,,:. __ \ $ i ble~ __ only_ during the
.. <*~succeeding seven (7) days provided that during such time the HPCI subsystem is operab 1e. If the ~--el-1-e*'"",r:e:--.:=::-- a..7"?,...0 7" 11 tJ.-*e..
4A§!-MAPLHGR reduction factors (multipliers) are
--* . applied tolf"Fi§l:Jl"e 3. §-1~ -,k, rnffPL#GK. t,.....,,,T.S
... *'the* Automatic Pressure
**::::>-Re 1 i ef Subsystem of ECCS
- sh~ll be considered operab 1e\. fl) O aq fore_
8*8 ftlel, 01 (2) 0. 76 feP e_ ** * * ** - * *. 9*9 ft1e1 ;e..... ~l:..,L_ IY1HPUU?:i1?. L,...,.,/!:; ~ a. ... d -rke..
"-.._ a."f?l"'opraa..n::._. Ynl1PLH6-R.. r-e.c/ ~
- 3. From and after
- the date a.t- r *d e.. We>1'1.
.._J
'""- ~c:.. Co,...e... Ope..
ucnov..s ror:; that two rel1~f valves [,..,,,,r:, Ke oir c;.,-r""'? are found or made to be P inoperable, reactor operation is permissible only during the succeed-ing seven days provided that during such time the HPCI subsystem is operable and the multipliers speci-fied in 3.5.D.2 are j applied.
- 4. If the requirements of
... 3.5.D.l cannot be met, an
~
l orderly shutdown shall I be initiated and the
- I l reactor pressure shall l
l be reduced to below
'1 150 psig within 24 hou~s.
1 E. Isolation Condenser System E. Surveillance of the Isolation Condenser System shall be performed as follows: 3/4.5-9
- j
-l
_;! -.-.. --- ***- - - ...... *** . *.- *-.::--- --;*---* ........ *--.~-*-*:-.**"' *_:-* ** :-~.-.-.**- -**-:** -* *--~----~--.-*:--*--1~--- . --~-,.1-* ,....,-;.,.,:~~ -*.-.1-;-.~-:-- .*~~---,..~~.~--*r*--~~.--:-***~
;:-.:_::'.~:Tif~~T~~vw~~:r~;75~~!~J0:fi~~rs*r~~f:.%~w~r}f:p:.~r§rf:~gii~w;:~G18:-1,>.i;.j?:':~J0;~~s0:i:~~E~t0J~,I~F?f;;:,0:~?~:?::.~0:~'~~~~~
DRESDEN III DPR-25 Amendment No. 75, 87, 94 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont 1 d.) (Cont 1 d.) I. Average Planar LHGR I. Average Planar linear Heat Generation Rate (APLHGR) The APLHGR for each type of fuel -a-s a funct io11 or average.bbndle exposure-<- shall be determined daily during reactor operation at greater than
-or equal to 25% rated thermal power.
'f:r S°,""7lc. lttrpOpef'Cl--no-i_ (SLO) ~ ~
IYIPiPLJ-1(,.K_ Lu'11T.5 Ska.LI,. be. Je.c.~ad by .-he.. 5LO r>111Pt.H-61?. mvLrrpl1c.c.11<1"e. k-r-or-(~). r+ COt'\CLJrre ....r t..71rh. Slo I (!)r'\e.. A-..,ro,._.a..,-1 C... Pr-e..$SU\e: Kel 1e.t Svb~re""' R~lref Vo.Lue 15 0..,.,..- 0\- ~e.tv 1 ee_ ( "F,V005) 7t.~ (Yl{WLJ.16-R t1o'Vlir5 sl..a..lt he_ 1 deLt-ec..:;ed b~ ,-le. KV00.5 tylffPLrlGf<_
/111.11..npLi l.c..,-,~ .fo..c:rorcs). 11..e- P'lf.1-Pi/IGi?
luV111S a. ..... rl mvll17>L1u:,.--n..Je.. -h:lc...--or-'5
,., a..r-e..
Spec.,..(:,eJ I"'- nv._ c (')_ Or-c.. '("ert:-"T'l"'1 Lli>I~- Ke.:por-r.
- j i
- .:.i
.i j .j J
- 3.5 LIMITING CONDITION FOR OPERATION (Cont'd.)
4.5 DRESDEN III SURVEILLANCE REQUIREMENT (Cont'd.) DPR-25 Amendment No. 75, 87, 94 J. LOCAL STEADY STATE LHGR J. Linear Heat Generation Rate ( LHGR) During steady state power The Fuel Design Limiting operation above 25% rated Ratio fop ExxeA F~e-'hrfFDLRX) thermal power, the linear* shall be checked daily heat generation rate (LHGR) during reactor operation of any rod in any fuel at greater than or equal to assembly at any axial loca- 25% rated thermal power. tion shall not exceed its maximum steady state LHGR (SLHGR) value shown in..........-----
~u.r-e-3-. §=.JA-(-GGR~ '---- f'Ae.
..tw-0 C'lr"e~ That is, the
- Fuel Design Limiting Ratio fe1 Exxo11 Fueh(FDLRX) shall not be greater than 1.0 where FDLRX = LHGR SLHGR Delete-I-
If at any time.* during operation I above 25% rated thermal power, it is determined by normal surveillance that FDLRX for any fuel assembly exceeds 1.0, action shall be initiated within 15 minutes to restore operation to within the pre-scribed 1imits. If the FDLRX is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours. Surveillance and corresponding action shall continue until react6r operation is within the prescribed limits. 3/4.5-16
DRESDEN III DPR-25 Amendment No. 94 3.5 LIMITING CONDITION FOR OPERATION 4. 5 .. SURVEILLANCE REQUIREMENT (Cont'd) (Cont'd) K. Local Transient LHGR K. Transient Linear Heat Generation Rate (LHGR)
- J At any time during power The fuel design limiting i operation, above 25% rated ratio for centerline
~i thermal power the fuel ~ . melt (FDLRC) shall be design limiting ratio for checked daily during centerline melt (FDLRC)
- reactor operation at shall not be greater greater than or equal to than 1. 0, where 25% rated thermal power FDLRC = LHGR) 1.2)
TLHGR) (FRP)
*i .,I *J 1
i If du.ring operation, the FDLRC exceeds 1.0 when operating above 25% rated thermal power, either:
- a. The APRM scram and rod block settings shall i be reduced to the values given by the equations in Specifi-cations 2.1.A.1 and 2.1.B. This may be accomplished by increasing APRM gains as described therein.
- b. The power distribution shall be changed such that the FDLRC no longer exceeds 1. 0.
3/4.5-17
- 1) e. l ~e..- /-PA-(;~
I
*-*~..... -~~***
DRESDEN III DPR-Amendment No. 75, 87, ~ MAPLHGR LIMIT VS. BUNDLE AVERAGE EXPOSURE ANF 8x8 FUEL f 13* .5
' /
/
13.9 I I'\ I/
.,... 12.* .5 ' "
..-" ' ' I I
J 12.0
/
I
'\.
5 11 * .5 I'\. I'\
,/
/
I "' 2 u.o i'\ I'\ I/ I/ 10~.5 I/
\ " r'I.
'\
I/ J 10.0 ' a* 10,~2 Z0,000 30. t>oo ~o.ooo JmmL! A *f!'W Dl'OSUJlI (HW'O ) The above graph is base on the following MAPLHGR summary or ANF 8x8 fuel design: Bundle Average MAPLHG Exposure (MWD/MTU Limit (kw/f 0 13.0 10,000 13.0 15,000 13.0 . 18,000 12.85 20,000 12.60 25,00 11.95 30,0 0 11.20 35 00 10.45 (Figure 3.5-1) _ (Sheet 1 of 2) I 3/4.5-i8 I
-';J:i~'*~:;;;;~;Ei;;:;:,,~*~~~?;s::~¥s.fl'1?.J.r/s;.~~£.:,:'>,,;',;*.*c:;{'.i.::;..6s;;~:1;.B<Jl':B.~~offiG.1::'i:ZS::w.i>.:;,z:~:i;J06~~il~~~;J;:i':~'"K*'C.\ilS&ti1if:.{j,_~f:~i[iJiflj;.fi.J/!f2f,i;/;£;~.7i.LJ~t*;-;,{.}fr..'f.,_'<:'.\YJ"':.'i-;.\'?' ~ -*. -*- ,'J
.:*~ *~
DRESDEN III DPR-2 Amendment No., 75, 76, 8 , 94 MAPLHGR LIMIT VS. BUNDLE AVERAGE EXPOSURE ANF 9x9 FUEL
- 11.5 u.o 10.5
~ 10.0 9.5
; 9.0
- 3 e.5 I e.o 7.5 1.0 6.5 6.0 0 10,000 20. 30,001) 40,000 1-l IONDL! G! mosuu < /'t!W) lJ
*i The above graph is bas ij on the following MAPLHGR summa for ANF 9x9 fuel l design:
l i Bundle Avera MA LHGR i Exposure (MWD TU) Limit kw/ft)
.,i Q 11. 0 J
*1 5,0 11.7
*~ 10 00 11.40
*l 1 1 ,000 10.55
*I J
0,000 9.70 l l
. 25 ,000 8.85
,j 30,000 8.00 35,000 7.15 40,000 6.30 Figure 3.5-1 (Sheet 2 of 2) 3/4.5-19
~
j . - *-*-*-- ... -*--. -.. . *- - -- ---- . ... . ..,_.._ . ---. *--------. **-*- -- . ----*. --- .
~~;tiid2Si3~J;t:":,:;.fu"i.1.:;~~,~~~:;~~gg~~~~;;;~\d*'*'~~'{i/:'iilf!!'"f/;,,~ili;~~""<'-';:,,.~,'?:<~~~J\-~~*~$5is:?tt~:N;'T~%;~:;?.&:~*sr,\:;~1?I<ri£.f.i~u~!rg::~~~
-~
DRESDEN III Amendment No. 75, STEADY STATE LINEAR HEAT GENERATION RATE LIMIT (SLHGR) VS. NODAL EXPOSURE
-=*-
~
PLANAR EXPOSURE 9x9 Fuel LHGR Exposure 16.00 0.00 14.50 14.10 5.00 14.50 9.30 25.20 10.80 48.00 . 7.20 Figure 3.5-lA 3/4.5-20
'::~i '~ . .. ****-*-*** - .. .. **- . _* ----"* _________ .... _ *--~~*-~~~~-~~~,.~~:?.}T:J::~~~~~W&~1'~~~~~~-];r~T~r;~~;m1?-B!}~~~~~B}~~~~~~%.~~~~~W:~~~~~~$'.t~t~15~f):?f#5-:~*
l""'"*""'§~~""",,;;-~A_~\,.-=_,11>~1liii'~~~ l)EL~ PA-~t:: ...... .. . .
) DRESDEN III
'*j Amendment No. 94
~}
TRANSIENT LINEAR HEAT GENERATION RATE LIMIT (TLHGR) VS. +/-f FOR ALL FUEL TYPES
*1 1 '.~
- ~1
~t~ ~i
-~*: {~
*J ~~ <1
- J
-:;12r------i~----1-~~-----1--..;._-~..J-------'
~
!11r---------ir------,+---~--+--.....;..--4-~----1
- ~--- -----
LHGR " *\ . 0.00 19.20 25.40 16.90 43.20 10.80 48.00 10.00 Figure 3.5-18 3/4.5-21
- r==-~':'.-:".~"."'-=~"'"~:"':::=~~::~~~~"-*:":'~~~~~~'!*-~~~""'"~...
~ ..
- f*'
- .~ . c*
DRESDEN III DPR-25 d*f Amendment No. 75, 87, 94
- ~? ..
3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT (Cont'd.) (Cont 1 d.) L. Minimum Critical Power L. Minimum Critical Power Ratio {MCPR) Ratio (MCPR)
'2.LL Core
~..,~ or-
-Floc.J~
a.uro "
'f I. During steady state
~rated core-MCPR shall be determined daily during a reactor
.flow co,.._Tt'ol ffcwj.. MCPR shall be
- power operation at greater
-'* greater than or equal t:o than or equal to 25% rated I i. 39. FeF Gore fl O\tSe.. thermal power and following
~ther thaR rated~ any ~hange in power level the MCPR OperatiRg 0 or distribution* that would bi.mit sl:lal 1 be as~ cause operation with a felle.,..s: * *limiting control rod pattern as described in the bases for Specification 3.3.8.5.
Manual Flow Control - the MCPR Operating
-Limit shall be the alue from Figure 3 --2 s et 1 or the ab e rat d flow value whic ever is gr. ater.
- __._/_ - - - - - - - ~- ---- - - - - - -
__2._j. During Single Loop Operation, the rated flow MCPR operating 1imi t shall be increased by an additive factor of 0.01. 3/4.5-22
,':[;4:-.~::,_:c:...-.:i:,,;...;:,.;;;,;:,._~~~-ilr:li:..;:;0..:.,: * ~ . . ~~*
. . , ** ~... ( '"'....)~.;.l~~::....t'~**;,...~*:~.~...li~~*.-.~~~;;.."io-"...*..o.~~~~'.i..~.$1.W"~='L-..'""":t".J.~>i:.'.l..!.&'~:.l#V-::?...."i"-!.!!;.'_;.~1'!;"~~~:i~~'2.;:;--"~**ct"' ....c ~-;;:;...-.Ol:.Z.
.*~ - . - ---*
-'?~
)~
A DRESDEN III Amendment No. 75,
~1.
0
-=
e1.&ot--~--+----r----+----+----+---+---r--+-----1 0
- 1. 50t-----i~--
c u 51.Jo .....---+------.. . . .--.. . . .---+---~.......---+----+----1
.a....
0
!1.20t----+------+--......~---+--r---+----_,;,,.;:~~-~---~
a u et.10 .......---+---+----+~~7'--+----+---+------f--~~
*u **
.1. o~o 30 40 80 90 100
*1he above curves re based on the follow*ng MCPR Limit summary for re ced Total Core Flow:
MCPR 8x8 1.10 1.15
*1.21
'\' ..
1.28 1.36 1.46 1.60 Figure 3.5-2 (Sheet 1 of 3) . MCPR Limit for reduced Total d\ore Flow
\
/.
**.*i 3/4. 5-25 I!
I
/
DRESDEN III DPR-25 Amendment No. 75, 87, 94 1.39 -- 1.36 ~1.70
=-
- 1. 60 u
3 1.50 0
~ 1.40
-~ 0 5 1.30 u 0 .: 1. 20 c u
~ 1.10
- I c
1.0~0 JO 40 10 80 90 100 The above 8x curves are based on the follo ing MCPR operating l"mit summary for Automatic Flow C MCPR Operating Limit for 8x8 fuel* 1.28 1.32 1.36 1.39 1.28 1.32 1. 36 1.39 1.31 1.35 1.39 1.43 1.34 1.39 1.43 1.46 1.39 1.44 1.48 1.52 1.45 1.49 1.54 l.58 1.52 1.56 1.61 1.65 1.66 1. 71 1.76 l.80
- Column headers are MCPR Operating Limits at rated flow.
Figure 3.5-2 (Sheet 2 of 3) 8x8 MCPR Operating Limit For Automatic Flow Control 3/4. 5-26.
DRESDEN III Amendment No. 87,
- .1.10~--~
a
!1.401--~~-+-_:.;;.~~+-~~-+,..;_~~+-~;.._-?f:O.......:::::-~;;pi--..:::::::=:i__;;=---~
!1.20~~~-1-~~~1--~~-+-~~~r-~~-r-~~--r~~~-r-~~-.
- I C
.*;~ .j 30 40 70 80 90 100 FLOW (% RATED, 98 M~ /HR) ves are based on the follo 'ng MCPR summary for Automatic Flow Co trol:
MCPR Operating Limit for 9x9 fuel*
- 1. 31 1. 35 1.39 1.31 1.35 1.39
--~
1.34 1.38 1.42
-l ... 1.36 1.41 1.45 1.41 1.45 1.49 1.46 1.51 1.55 1.53 1.58. 1. 62 1.68 1.73 1.78
- e
- Column headers are MCPR Operating Limits at rated flow.
Figure 3.5-2 (Sheet 3 of 3) 9x9 MCPR Operating Limit For Automatic Flow Control 3/4.5-27
- 1
-1 ij ...-- . ... .. -** - ... -* - . *- - - . - - - .... - - .. - ..
j . .*--~:1&::~~TS~@}~:~N.::;;.::::.t~~s~~~~~~~~~G::~f?l=?~---~~?~~2*;~:;"1;~~:~~0~:r.::*~~.c;::;:~~~~~?F~~~1~~~~~£<:7:_:'i,:~:~f.~1~?.:~~.~t~t-:;1i::f:~;:.~~?~:;*:0.(.,J:;:~~~A*1*~~;1F.
DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) temperatures ANF has analyzed the effects Single Loop Operation has on LOCA events (RefeFeRee 4)1i! For breaks in the idle loop, the above Dual loop Operation results are conservative (RefereAee 1)~ For breaks in the active loop, the event is more sev~ere prim;:il~ due to a more rapid loss of core flow. By applying ~ultip ica- a..V\ a.Yfrornc . tive 8.9l'reduction factor to the results of the previous analyses, all applicable criteria are met. J. local Steady State LHGR This specification assures that the maximum linear heat genera- - tion rate in any fuel rod is less than the design linear heat generation rate even if fuel pellet densifjcation is postulated. This provides assurance that the fuel end-of-life steady state criteria are met.
- 11 Loss of Coofaht Ac£ld1:!11L A11c:dys~! Report for On~,".i~n ll_pi.
Quad-Cities Units 1, 2 Nuclear Power Stations --'NED0-24146 , vision 1, April 1979. (2) (3) XN-NF-85-63 "Dresden . 3 LO E.CCS Analysis MAPLHGR results for 9x9 fuel", Septebmer 1985~ ~ * (4) ANF , "LOCA-ECCS Analysis for Dresden Uni Operation with ANF Fuel," September 1987 . l . 'i B 3/4.5-36 I
** .,..-........._ . __ .,, _*.* - * * -:-"": .. ~-**--* - -*--.-*.~ *~ ..,~ ..... *-..
* .* 7'
. -. ..... *-.--***. -*--*-.-- ~---*-* .... *-* *-* . *-** ..... ---,-.
-~------**I I
_*--1~TT:;:~];:;\~'.:~;(*~~{fil~)}~:,,~~1!t~3~f,§7'%~~~;~&~~+/-:§'f.i:~Wg;~~~'f!}'.'f\{i.'?~R*}~~~1~7~0({~~?:;0f;~?@:~0:'~~?~':t~A~;~:*1*
_-i*
~1 *J i *i '.J
- '._,j I *j
. DRESDEN III DPR-25 Amendment No. 75, 87, 94, 102 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
K. Local Transient LHGR This specification provides assurance that the fuel will neither experience centerline melt nor exceed 1% plastic cladding strain for transient overpower events beginning at any power and termi-nating at 120% of rated thermal power. L. Minimum Critical Power Ratio (MCPR)
- ~:~ .The steady-state values for MCPR specified in the Specification
~\j Tue.. <lNF N'Rc-~rc.Jfd~Were determined using -e-l=lel'TllERMEP"thermal limits methodology---... ANF rrKc.-a:ppr-ov"'4.
- .-~j rv.e:ri..cdoL=, 1.,~-recl Jdescribed in'})(N NF 88 19, Volame 3~ The safety limit implicit
'.*-.:*!* '"'- °'S?e.c.1-t,'L._T,.,o'\_ c., A in the Operating limits is established so that during sustained
- 1 1 * * *"I* operation at the MCPR safety limit, at least 99. 9% of the fuel .. -.
rods in the core are expected to avoid boiling transition. The * ** -**- Limiting Transient delta CPR implicit in the operating limits was calculated such that the occurrence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combina-
*i tions of uncertainties.
'1
' ), Transient events of each type anticipated during operation of a
-1 BWR/3 were evaluated to determine which is. most restrictive in l terms of thermal. margin requirements.* The generator load
~j rejection/turbine trip without bypass is typically the limiting
-1 j
ANF Nl?c.-~ Me.~ r-<>ve j event. *The thermal margin effects of the event are evaluated
~~l=le TMERMEX Hetheaele§:5t...and appropriate MCPR limits con-j 0
L°?y sistent with t~*N Je.critical power *correlation are determined. _1 fSeveral factors influence which transient results in the largest
-.l -rle.. ANF- 1\JKC.-<<rpro11"""- reduction in critical power ratio, such as the cycle-specific
~ fuel loading, exposure and fuel type. The current cycle's reload
~ licensing analyses identifies the limiting transient for that i cycle.
*1
'i 1i As described in ~pecification 4.3.C.3 and the associated Bases, observed 1ant data.":\were used to determine the average scram per ormance used in the transient analyses for determin-
;) or T""e..c...\,."=L ~ ing the MCPR Operating Limit. If the current cycle scram time 1 ~-pec.1.f1 cc.:r1<>""-. l1M1"T"f, performance falls outside of the distribution assumed in the analyses, an adjustment of the MCPR limit may be required to
-:'l 4~c.1.f1C;.T/00"\. 3, "3.C.) maintain margin to the MCPR Safety Limit during transients.
l Compliance with the assumed distribution and adjustment of the
,'.~
MCPR Operating Limit will.be performed as directed by the
*.i nuclear fuel vendor in accordance with station procedures.
. ~
B 3/4.5-37
" (
j,~~;;~f~;;~~~~;~~~~~;;;2ri~~;;~~~~~,~~;~~;~~t\'f~,~,;~;;;;;;;;;
r,*,~="t"~~"""~""""~::'.:"'""'""'~~""~ille".'_'i?""-":"""""'"'":'.".'""~""' '2:'"' "'"'1"'"'"*'*"'*:*~,,,
*~
- ~. DRESDEN III DPR-25
-~~ ":,~
Amendment No. 75, 87, 94, 102
~1 3.5 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
- J i For core flows less than rated, the MCPR Operating Limit
** established in the specification is adjusted to provide ,\1 protection of the MCPR Safety Limit in the* event of an
- '.;}, uncontro 11 ed reci rcul at ion fl ow increase to the physical d- limit of pump flow. - This protection ii provided for manual
- ~*j --Je. Co a~d, automatic fl ow contro ~ by choosing the MCPR operating
-:1 re C>~Tl"'-7 )jl lml t as the va 1ue from\fi ga1 e 3. 5 f Sheet 1 01 the 1 e:tee e...
- j l,.,,,,,n Re;:-..l'T. _ / ffPE! flgw.valwe! '*fhieheoe1 is g:eate~ For Automati7 ~low
'*l
- Centro l , in add1 t l on to protecting the MCPR Safety Ll m1 t
~ * . _ _ during the flow run-up event, protection is provided against
-4 ** violating the rated flow MCPR Operating Limit during an *
~!1 automatic flow increase to rated core flow. This protection ~ is rovided b the reduced flow MCPR limits shown in
-~
- eet or 3 where the curve correspondi
_:j -. the curre ed flow MCPR limit is used (l i . nterpola-
. tion between the "mit lines de
- is permissible) .
.:j Therefore, for Automatic rol, the MCPR Operating 'J! Limit is chosen as a ue from l 5-2 Sheet 1, -~ Sheet 2, or the rated flow value, wh1c is Analyses have demonstrated that transient events in Single Loop Operation are bounded by those at rated conditions; however, due to the increase in the MCPR fuel cladding integrity safety limit in Single Loop Operation, an equiva-lent adder must be uniformly applied to,.&l f!- ~CPR LC~o __J ed. -f~
maintain the same margins to the MCPR fuel c adding /)...e_..r~~ ..., integrity safety limit. M. Flood Protection Condensate pump room flood protection will assure the availability of the containment cooling service water system (CCSW) during a postulated incident of flooding in the turbine building. The redundant level switches in the condenser pit will preclude any postulated flooding of the turbine building to an elevation above river water level. The level switches provide alarm and circulating water pump trip in the event a water level is detected in the condenser pit. B 3/4.5-38 "l ,- - .-:- *. *:**-*'"'**'"",'* _.-.,*-.* -:-*------* .*-*------- *..,..,....-... -:~*-*-**----- .. ......... *--*** ......... ----.. *-------~* -*-- ..... ,. ... .._ . ,.--* -* -*-**-. -- r**--~_ . . .**-*.* -': <("*'""*--
~*\~-~~¥f~~:~~tJr}f~;:::~y:!:~7:!~Zt~r;~~:~i~r:r.~=~~.~:!*~7.{:~*::.;r0 ~i:S~~r;;.~~~;~I~i.~.:iTI:~5W!~~J::~J:~3~~:~:~~r}-~.~~~~~~S;~:.~~*:~;'.~~~~~~?{~;-r;~:::*>~.r~,:c~r~~0;:~t~~~~~~~~-~~~\~~?l1'-*\.?*~~-:*?.~N~
DRESDEN III DPR-25 _. . .Amendment No~ 75; 87, 94, 102 4.5 SURVEILLANCE REQUIREMENT BASES {Cont'd.) i
~}
- ) evaluation of the average planar LHGR below this power level is
- j not necessary. The daily requirement for calculating average
.l **l planar LHGR above 25 percent rated thermal power is sufficient since power distribution shifts are slow when there have not been significant power or control rod changes.
J. Local Steady State LHGR fDLT-<><~LllGR-Lfor all fuel shall be checked daily during reactor operation at greater than or equal to 25 percent power to determine if fuel burnup or control rod movement has caused changes in power distribution. A limiting LHGR value is precluded by a considerable margin when employing a permis-sible control rod pattern below 25% rated thermal power. K. Local Transient LHGR The fuel design limiting ratio for centerline melt {FDLRC) shall be ~hecked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution. The FDLRC limit is designed to protect against centerline melting of the fuel during anticipated operational occurrences. L. Minimum Critical Power Ratio (MCPR) i l At core thermal power levels .less than or equal to 25 percent,
',J the reactor will be operating at minimum recirculation pump l !
speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at
*( this point, operating plant experience and thermal hydraulic
.*1 analysis indicates that the resulting MCPR value is in excess j of requireme.nts by a considerable margin. With this low void l! content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.
j ; The daily requirement for calculating MCPR above 25 percent j rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant
- power or control rod changes.
In aidition, the reduced flow correction applied to the LCO provides margin for flow increase from low flows. M. Flood Protection The watertight bulkhead door and the penetration seals for pipes and cables penetrating the vault walls have been designed B 3/4.5-41
- 3.6 LIMITING CONDITION FOR OPERATION 4.6 DRESDEN III DPR-25 Amendment No. 75, 78, 87, 94 SURVEILLANCE REQUIREMENT (Cont 1 d.) (Cont 1 d.)
- e. The suction valve in the idle loop shall be closed and electrically isolated except when the idle loop is being prepared for return to service; and
- f. If the tripped pump is
.out of service for more than 24 hours, imple-ment the following additional restrictions:
- i. The flow biased RBM Rod Block LSSS shall be reduced by 4.0%
(Specification 3.2.C.1); ii. The flow biased APRM Rod Block LSSS sha 11 be reduced by 3.5%
.* (Specification 2.1.B);
iii. The flow biased APRM scram LSSS shall be reduced by 3. 5% (Specification 2.1. A. l); The MCPR Safety
.I Limit shall be increased by 0.01 (Specification 1.1. A);
- v. ThefMCPR Operating Limit shall be increased by 0.01
{Specification
- 3. 5. L.t:.;
z. 3/4.6-15
*---~. ~ -*- _ __:____ _____________________
DRESDEN III DPR-25 . Amendment No. 75, 78, 87, 94 3.6 LIMITING CONDITION FOR OPERATION 4.6 SURVEILLANCE RE UIREMENT
**(Cont'd.) Cont'd.)
vi. The MAPLHGR Operating Limit shall be reduced by -ae- . ~ a.'"fpr?r1<*-Te-multi pl i cat i Ve factor M e. 91e ~l~f"'o*- (Specification '------ r-- I~ <!.ore. -r It-.~
/").....e.r~.,..,~*-~,
**"1 L.-.~c r- ....
3.5.I). If 1?epo~r concurrently, orie ..
'* Automattc: Pressure
<Re 1i ef Subsystem ..:* *:* ****-**- ***-
relief valve is out-of-service, the MAPLHGR Operating Limit shall be reduced by ~ ~-----rk. CI:fPi'CSf/"l"-T"e-multi pl icat i Ve. f c': tor'R C. 39 f&P~ [. ,..l~ C.Ore... ~.,......., lu,,,,rs 17___ .,_...,....
'8~:... ftte:. aRe o. 76 <- ....,....... , ~*,
fer 9>t9 fttel.e...
- 4. Core thermal power shall not exceed 25% of rated without forced recircu-lation. If core thermal
'power is greater than 25%
of rated without forced recirculation, action shall be initiated within 15 minutes to restore operation to within the prescribed limits and core thermal power shall be returned to within the prescribed.limit within two (2) hours. I. Snubbers (Shock I. Snubbers (Shock Suppressors) Suppressors) The following surveillance requirements apply to safety related snubbers.
~-;
3/4.6-16
- 1
- -1
'l
- 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)
DRESDEN III DPR-25 Amendment No. 78, 87, 94 In addition, during the start-up of Dresden Unit 2, it was found that a flow mismatch between the two sets of jet pumps caused by a difference in recirculation loops could set up a vibration until a mismatch in speed of 27% occurred. The 10% and 15%'speed mismatch restrictions provide additional margin before a pump vibration problem wiil occur. Reduced flow MCPR Operating Limits for Automatic Flow Control are not applicable for Single Loop Operation. Therefore, sustained reactor operation under such conditions is not permitted.
- Regions I and II of Figure 3.6.2 represent the areas of the power/flow map with the least margin to stable operation.
Although calculated decay r~tios at the intersection of the natural circulation flow line and the APRM Rod block line indicate that substantial margin exists to where unstable operation could be expected. Specifications 3.6.H.3.b., 3.6.H.3.c. and 4.6.H.3. provide additional assurance that if unstable operation should occur, it will be detected and corrected in a timely manner. During the starting sequence of the inoperable recirculation pump, restricting the operable recirculation pump speed below 65% of rated prevents possible damage to the jet pump riser braces uue to excessive vibration. The closure -0f*the suction valve in the idle loop prevents the loss of LPCI through the idle recirculation pump into the downcomer. Analyses have been performed which support indefinite operation in single loop provided the restrictions discussed in Specification 3.6.H.3.d. are implemented within 24 hours. The LSSSs are corrected to account for backflow through the idle jet pumps above 20-40% of rated recirculation pump speed. This assures that the original drive flow biased rod block and scram trip settings are preserved during Single Loop Operation. The MCPR safety limit has been increased by 0.01 to account r for core flow and TIP reading uncertainties which are used in the statistical analysis of the safety limit. In addition,
~. thlWMCPR Operating Limit has also been increased by 0.01 to ~atedflow maintain the same margin to the safety limit as during Dual Loop Operation.
B 3/4.6-36
DRESDEN III DPR-25 Amendment No. 78, 87, 94 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.) cdec:..rea..se- ee. tby 'fle.... r\t()l*rrr l,~, The llH:llpHeatioe 8.91 redt:1etierf--of MAPLHGR Operating Limit~ .+ctc.n:>r Sp .C,e.cJ ~ accounts for the more rapid loss of. core flow during Single L Tk Co ~ '""'" Loop Operation than during Dual Loop Operation. L*hrl-rs kro c:.rc.,-,-,"'1 The more conservative MAPLHGR reduction factors cf 9.89 fep4e.. 8MB fwel aAa Q,7i fe~ g*g f~el2.are applied if one relief valve and one recirculation loop are inoperable at the same time. The small break LOCA is the concern for one relief valve out-of-service; the lar_ge break LOCA is the concern for Single Loop Operation. Selecting the more conservative MAPLHGR
*multipliers will cover both the relief value out-of-service and Single Loop Operation.
**':Specification 3.6.H.4 increased the margin of safety for thermal-hydraulic stability and for startup of recircula-tion pumps from natural circulation conditions.
I. Snubbers (Shock Suppressors) Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety sys~em or compo-nent be operable during reactor operation. *
. Because the snubber protection is required only during low probability events, a period of 72 hours is allowed for repairs.
or replacements. In case a shutdowri is requiTed, the allowance of 36 hours to reach a cold shutdown condition will permit an orderly shutdown consistent with standard operating procedures. Since plant startup should not commence with knowingly defec-tive safety related equipment, Specification 3.6.I.4 prohibits startup with inoperable snubbers. When a snubber is found i noperab 1e, a review sha 11 be performed to determine* the snubber mode of failure. Results of the review shall be used to determine if an engineering evaluation of the. safety-related system or component is necessary. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the support component or system. All safety related hydraulic snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures. B 3/4.6-37
DRESDEN III DPR-25 Amendment No. 75, 79, 92, 100 6.0 ADMINISTRATIVE CONTROLS (Cont'd.) additional specific details required in license conditions based on other commitments shall be included in this report. Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.
- 2. A tabulation shall be submitted on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions, (See note); e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at leQst 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.
- 3. Monthly Op~rating Report Routine reports of operating statistics and shutdown experiences shall be submitted on a monthly basis to the United States Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate Regional Administrator, to arrive no later than the 15th of each month following the calendar month
~ covered by the report.
- 8. *Reportable Events Reportable events will be submitted as required by 10 CFR 50.73.
Note: This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
- 6-18 *- Amendment No. 100
ENCLOSURE E EXAMPLE CORE OPERATING LIMITS REPORTS
- 1. DRESDEN UNIT 2 CYCLE 12
- 2. DRESDEN UNIT 3 CYCLE 12 0334T:l6}}