ML20113H295
| ML20113H295 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/15/1992 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20113H297 | List: |
| References | |
| NUDOCS 9207290153 | |
| Download: ML20113H295 (13) | |
Text
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UNITED STATES n
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,i NUCLEAR REGULATORY COMMISSION p
a WASHING TON, D C. 20%5
'99 c.
_C_03SUMERS POWER CDMPANY EDIKET NO, 50-255 EAl 15M15 PtANT A3fRpKtli T0 iACILITLQPIBATING licit [51 Amendment fM. 150 License No. DPR-20 1,
The Nuclear Regult. tory Commission (the Commission) has found that:
A.
The application for amendment by Consumers Power Company (the licensee) dated r bruary 3, 1992, complies with the standards and e
reouirements of the Atomic Energy Act of 1954, as amended (the Act),
ar.d the Connission's rules and regulations set forth in 10 CFR Chapter 1; B.
The facu lty will operate in conformity with the application, the provisions of the Act, and the rules and regulations the Comi sion; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducts:1 without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
Tha issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E,
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Consission's regulations and all applicable requirements have been satisfied.
l 2.
Accardingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C.? of facility Operating License No DPR-20 is hereby.
i amended to read as follows:
l l
i 9207290153 920715 PDR ADOCK 05000255 1
P PDR
2-Itc.tatal sonifiutiom The lechnical Sper..ifications contained in Appendices A and B, as ravised through Araendment No.150, are herriby incorporated in the license.
The licensee shell operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR IHE NUCLEAR REGULATORY COMMISSION k
. B. Ma
. Director Project Directorate 111-1 Division of Reactor Projects Ill/lV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 15, 1992 r
~
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____.______.___._.__.__m____________._______________________m_____.__________._
____. _ _ _ _ _ _ _. _ __U
ATTA(11tiLNT TO Ll.(U!SELMEUDMRLBO.ml50 fKimLOPERAT1!!LLifI!!SLHO., OPR-20 DO M E L if M M M Revise Appendix A Technical Specifications ty removing the pages identified below and inserting the attached pages.
The revised pages are identified by the amendment number ano' contain marginal lines indicating the are of change.
RltiQ1E IffiEfd i
i 2-1 2-1
^
2-2 2-2 2-3 8 2-1 2-4 8 2-2 2-5 B 2-3 2-6 b 2-1 2-7 B 2-5 2-8 2-9 d-3 4-3 4-5 4-5 m
PALISADES PLANT TECHNICAL SPECIFICATIONS IABLE OF__ CONTENTS - APPDfDJX_8 SLCIl0S DL5CR1PTION MQLEQ l.0 DEIIHLILQRS 1-1 1.1' REACTOR OPERATING CONDITIONS 1-1 1.2 PROTECTIVE SYSTEMS 13 1.3 INSTRUMENTATION SURVEILLANCE l-3 1.4 HISCELLANEOUS DEFINITIONS 1-4 2.0 LAEU1_UnlIS _ AND LJH11[8.Q_S AF E T Y S Y S ifM_1111],RQ1 2-1 2.1 SAFE 1Y LfMITS - REACTOR CORE 2-1 2.2 SAFETY LlHITS PRIMARY COOLANT SYSTEM PRESSURE 2-3 2.3 LlHITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM 2-4 Table 2.3.1 Reactor Protective System Trip Setting Limits 2-5 B2.1 Dasis - Reactor Core Safety Limit B2-1 B2.2 Basis - Primary Coolant System Safety Limit 82-2 02.3 Basis - Limiting Safety System Settings B2-3 3.0 ljMlIIRQl0NDIT10NS FOR CPERAT103 3-1 3.0 APPLICABILITY 3-1 3.1 PRIMARY COOLANT SYSTEM 3-lb 3.1.1 Operable Components 3-lb figure 3 0 Reactor Inlet Temperature vs Operating Pressure 3-3a 3.1.2 Heatup and Cooldown Rates 34 Figure 3-1 Pressure - Temperature Limits for Heatup 3-9 Figure 3 2 Pressure - Temperature Limits for Cooldown 3-10 Figure 3 3 Pressure - Temperature Limits for Hydro Test 3 11 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Haximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 3.1.6 Maximum Primary Coolant Oxygen and Halogens Cu.centrations 3-23 3.1.7 Primary and Secondary Safety Valves 3 25 3.1.8 Overpressure Protection Systems 3-25a 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3 26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4
. CONTAINMENT COOLING 3 34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 l
3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.1 Containment Penetrations and Valves 3-40b
-3.7 ELECTRICAL SYSTEMS 3 41 3.8 REFUELING OPERATIONS 3 46 3.9 EFFLUENT RELEASE (DELETED) 3-50 l
l t
j Amendment No. JJ, JJ, J/, EE, J E, JJE, JJE,1
[,,.._
~[
2.0 EA[111.L1M1TS AND_1Di[TINO SAFEII.1Y31LtLifJIltlE 2.I 1Af11.Y 11m.it - R,gArdor Core The Minimum DNBR of the reactor core shall be maintained greater than or equal to the DNB correlation safety limit.
Correlation Safety Li il f!L XHB 1.17 ANfP 1.154 acclicabilitv Safety limit 2.1 is applicable in H0T STANDBY and POWER OPERATION.
M11pli 2.1.1 li a Safety Limit is exceeded, comply with the requiron,ents of Specification 6.7 2.2 16fel.YliStit Primary Coolant System Pres.t.ur.c (PCS)
The PCS Pressure shall not exceed 2750 psia.
Mplicability Safety Limit 2.2 is applicable when there is fuel in the reactor.
Action 1
' 2. 2 '.1 -
If a Safety Limi't is exceeded,' comply with the rcquirements of I
Specificetion 6.7 2.3 Limiting _Jafety System Settingt - Rea.gipr Proteq.tive Syster) (RPS)
The RPS trip setting limits shall be as stated in Table 2.3.1.
hplicability Limiting Safety System Settings of Table 2.3.1 are applicable when the associated RPS channels are required to be OPERABLE by Specification 3.17.1.
Action 2.3.1 If an RPS instrument setting is not within the allowab'e settings of
'3 Table 2.3.1, immediately declare the instrument inoperable and complete corrective action as directed by Specificatioa 3.17.1.
s Amendment No. JJ, 25, 47, JJE, JJ/,150 2-1
\\
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i
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IbBLE 2.3.d i
I f!f611QR_PR0_TECTLYLS1111M TRtP SETTING titil1$
four Primary Coolant Three Primary Coolant RPS Trio Unit Pumos Onetplina Pumps Open, tina 1.
Variable High s15% above core power, sl5% above core power Power with a minimum of with a miniinum of s30% RATED POWER sl5% RATED POWER and a maximum of and a maximum of
$106.5% RATED POWER.
549% RATED POWER.
2.
260% Full PCS Flow.
3.
High Pressure 52255 psia.
s2255 psia.
Pressurizer 4.
Thermal Margin /
(a)
(a)
Low Pressure i
S.
Steam Generator Above the feedwater Above the feodwater Low Water Level ring certer line.
ring center line.
6.
Steam Generator 2500 psia.
2500 psia.
Low Pressure 7.
Containment High s3.70 psig, s3.70 psig.
Pressure (a)'
The. )ressure setpoint for the' Thermal Mar in/ Low. Pressure Trip, P,l(p', is '
s thc" 11gher of two values, P,in and P,,,, bo h in psia:
P) = 1750= 2012(QA)(QR ) + 17.0(T ) - _9493 P
3 in where:
= -0.720fASI + 1.028; when -0.628 <'ASI < -0.100 QA
= -0.333/ASI
't 1.067; when -0.100 s ASI < +0.200 QA
+0.375dASI + 0.925; when +0.200 s ASI s +0.565 Measured ASI when Q k 0.0625 ASI
=
0.0 when Q < 0.0625 ASI
=
0.412(Q) + 0.588; when Q s 1.0 QR
=
3 Q;_
when Q > 1.0 OR,
=
Core Power / Rated Power 0
=
Maximum primary coolant inlet temperature, in
'f.
.T in ASI, T, and Q are the existing values as measured by the associated in instrument channel.
Amendment No. JJ, E, JJE,17E.150 2-2 l
2.0 fgSIS - 54fsly limittap1Limitina Safeu System Settinas 2.1 Enit - Reactor Core Safety limit To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperatcre. The upper boundary of the nucleate boiling regime is termed
- departure from nucleate boiling" (DNB?
At this point, there is a sharp reduction of the heat transfer cec'ft:ient which would resultirhiahcladdingtemparaturesandthepossibilItyofcladding i
failure. Although DNB is not an observable parameter during reactor coolant flow,e observable parameters of thermal power, primary operation th the use of a OliB Correlation. pressure, can be related to ONB through temperature and DNB Correlations have been developed to predict OtiB and the location of DNB for uially uniform and nonuniform heat flux distributions.
defined as the ratio of the heat flux that would cause DNB at a particular core location-to the actual heat flux, is indicative of the margin to OfiB.
The minimum value of the DNBR,'during steady-state operation, normal operational transients, and anticipated transients is limited to ONB correlation safety limit.
A DNBR equal to the DNB correlation safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur which is considered an appropriate margin to DNB for all operating conditio ts.
The reactor protective system is designed to prcvent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a OliBR of less i.han the DNB correlatten safety limit. The Palisades safety analyses uses two DNB correlations.
The XNB correlation discussed in References 1 and 2 determines tre safety limit for those fuel assemblies initially loaded in Cycle 8.
The ANFP correlation discussed in References 4 and 5 determines the safety limit for those fuel assemblies initially loaded in Cycle 9 and later.
Fuel construction than l'y loaded in Cycle 8 are of a differentater assemblies which utilize a H assemblies initiall Performance design.
The minimum DNBR analyses are in accordance with Reference 6.
References I? XN-NF-621(P)(A), Rev 1 2 i XN-NF 709 i l'ndated FSAR, Section 14.1.
AhF-1224
)ANF-89-19.{
, tiay 1989 A) January 1990
)
) XN-NF-82 2
), Revision 1 Amendment No. JJ, /3, JJE, J//,150 8 2-1
l 2.0 DASIS - Safety Limits and limiting Safety _1y11ntit(11nn 2.2 B1111 - Primary Coolant System Safety Limit The primary coolant system"8 serves as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere.
In the event of a fuel cladding f ailure, the primary coolant system is the foremost barrier against the release of fission products.
Establishing a system pressure limit helps to assure the continued integrity of both the primary cooiant system and the fuel cladding.
The Primary Coolant System design pressure is 2500 psia.
The maximum allowable Primary Coolant System transient pressure is limited by the pressure vessel limit (ASME Code, Section 111) of 110% of design pressure and by the piping, valve, and fitting limit (ASA Section B31.1)ted at 125f,of design pressure (3125 psia) ydrostat',c test was of 1207,of design pressure. The initial h conduc to verify the integrity of the primary coolant system.
Thus, the saf< sty limit of 2750 psia (Igy, of the 2500 psia design pressure) Pressure Trip, has been established.
The settings of the reactor High primary safety valves, and secondary safety valves have been established to assure never reaching the primary coolant system safety limit. Additional assurance that the nuclear steam supply.
by the norma? pressure does not exceed the tafety limit is provided system (NSSS l settirig of the atmospheric steam dump and turbine bypass valves of 900 psia.
References 1 Vodated FSAR, Section 4.
2 Updated FSAR, Section 4.3.
O Amendment No /E, JJE,150 I
I B 2-2
7.
2.0 BASIS Safa.ty_l.imits and LimLtlng_lafety Systtm_leltdngs 2.3 M is - Limitina Safety S.yltem Settinas The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached.
1.
Variable Hich Power - The Variable High Power Trip (VHP1) is incorporated in the reactor protection system to provide a reactor trip for transients exhibiting (a core power increase starting from any initial power level such as the l'oron dilution The VHPT system provides a trip setpoint no more transient) determined amou..t above the indicated core power with a than a pre specified upper limit. Operator action is required to increase the setpoint as core power is increased; the setpoint is automatically decreased as core power decreases.
Provisions ha/e been made to select different set points for three pump and four 5
pump operations.
During normal plant' operation with' all primary coolant pumps operatina, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximum actual steady.itate power at which ofsafetyanalysis.gtedis1157.,whichwasusedforthepurpose l
a trip would be act A reactor trip is Erimary Coolani lystep (P1Sl12w O_oy B should the coolant flow 2.
providedtoprotectthecoreagigstDN suddenly decrease significantly.
Flow in each of the four coolant loops is determined from pressure drop from inlet to outlet of the steam generators.
The total flow through the reactor core is determineds for the RPS flow channels, by summing the loop pressure drops across the steem generators and correlatino this pressure s'am with the sum of steam generator at 100% flow four pump The normal flow wi(th three aumps differential pressures which exist:
operation at full power t,K).
operatina is 74.7% of fulf S Flow.
Full PCS flow is tha; flow which exists at RATED POWF.R, at full power T with four pumps m,
operating.
During four pump operation, the Low flow Trip setting of 95%
insures that the reactor cunot operate wher the flow rate is less thg 93% of the nominal value considering instrument errors.
Provisions are made in the reactor protective system to permit operation of the reactor at reduced power if ont coolant pump is taken out of service. These low-flow and high-fiux settinas have beenderivedinconsiderationofinstrumenterrorsandresponse p'
times of equipment involved to assure that thermal margin and flow stability will be gintained during normal operation and anticipated transients.
For reactor operation with one coolant pump inoperative, core power must be reduced and then the Variable High Power and Low Flow setpoints must be adjusted to the three pump values before the pump may be stopped.
Amendment No. JJ, JJE, JJJ,150 B 2-3 b
2.0 DASJS - Sjtfety limits and_Limitino Safetv ErttrLSitija91 2.3 B331s - Limittnq_1gfgly_.Sylt.em.ic11]J)g (continued) 3.
litgh_2,3nyr.inttf~runtte - A reactor trip for high pressurizer pressere is provided in conjunction with the primary and secondary safety valves to prevent primary system overpressure (Specification 3.1.7).
In the event of loss of load without reactor trip, the temperature and pressure of the primary coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators.
This setting is consisten(withthetrippointassumedintheaccident analysis.
4.
JAennJtL11tr.qtaflew Preswe (TM/LP1 Trio The TM/LP trip system monitors core power, reactor coolant maximuli inlet temperature (T ), core coolant system in pressuraandixialshapeIndex.
The Low Pressure Trip limit is calculated using the equations given in Table The calculated limit (P,7). is than compared to a fixed Low The auctioneered highest of Pressure Trip limit (pD,ie trip limit (P,id,).
these signals becomes P
is gSignal is compared to the measured PCS pressure ar a tr generated when the ru asured pressure for that tiennel is less a
than or equal to P g A pre trip alarm is also generated when P 4s less tha,n 8r. equal to the pre-trip setting P,g, +
AP.
The TM/LP trip set points are derived from the-4 pump operation corethermallimitsthroughapplicationofappropr:ateallowances for measurement uncertainties and processing errors.
A pressure allowance of 165 pst is assumed to account for instrument drift in both power and inlet temperatures, calorimetric power a
measurement, inlet temperature measurement,.nd primary system pressure measurement. Uncertainties accounted for that are not a part of the 165 psi torn include ellowances f or assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temocrature measurement time delays, and ASI measurement.
Each M these allowances and uncertainties are included in the development of the TM/LP trip set point used in the accident analysis, t
k' Amendment No EJ, EE, JIE,150 t
B 2-4
l 2.0 MS.15 - Safety limits Bnd limiting lafety Sylltm_SettinQs 2.3 Djuis - Limitinq 531cly__Snlem Sttijsgi (continued) 5.
{pw Steam Gfneratsn,93_tuLleyql - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly p!antshutdownandtopreventsteamggeratordryoutassuming minimum auxiliary feedwater capacity The setting listed in Table 2.3.1 assures that the heat transfer critical (. tubes) is covered with water when the reactor is surface 6.
Low Stgam Generator Etengrg - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant.
The setting'of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the evtnt of excessively high sgam flow.
This setting war used in the accident analysis 7.
fgniainmgnt High presigt.q - A reactor trip on containment high pressure is provided to assure that the reactor is shutdown beforetheinitiatgnofthesafetyinjectionsystemand containment spray, R?ferences
(? EMF-91-176, Table 15.0.7-1 l
p Updated FSAR, Section 7.2.3.3.
1 f g EMF-91-176 l
JpXNNf-86-91(Section15.0.7-1 P)
J p ANF-90-078, Section 15.1.5
( p ANF 87-150(NP), Volume 2, Section 15.2.7 i pUpdatedFSAR,Section7.2.3.9.
( ) Adf-90-078, Section 15.2.1 Amendment No 7J, E2, JJE, JU,
29 Jp,150 B 2-5
n s
Tfa 2 4.I.1 Minisua Freoxncies for checks, Ce*ibratione and Testing of Reactor Protective System Surveii1 ente m t%f surveittence e
Function _
Freo m CNemet De*.cr,igt f on
- e. Corporison of fcur-power charrwrt readings.
- a. Check"'
S 1.
Power Range Safety Chamels
- b. Chamel sciustme st to agree with fseet betence itcutetler.
- b. theck
- O Repeet thenr<w f tta-Ar gewer corser.ters eter.
- c. Internal test signet.
- c. Test
- 3. Chamet allywnt th ough measuret/edjus'mer t cf
- d. Cattbrate*
4 intemet test points.
- e. Comarison cf chemet Indications,
- s. Check 5
2.
Wide-Range
- t. Test P
- b. Internet test signet.
I weutron Monitors
- c. Chemet ellgrismt through swesurementfoojustet ef
- c. Ceilbrate R
internet test pointa.
- e. Corverison of four eeperate totet ?!w indicatims.
- e. Check S
3.
eeector Coctent 7 tow
- b. Knonce dif ferentiel pressure e;ctled to se sers.
f
- b. Calibrate R
M*
- c. Sistebte trip tester. #
- c. Test
- e. Check:
5
- s. Check:
4.
YhL val Nerginflote (1) teeperisert of four separate calcutste) trip pressure
- 11) Te v ersture Pressuriter Pressu-e set point Indicatiers.
frput (2) Cercertson ci four pressurizer pressure indications.
(2) Pressure Some es Sie) tietow.)
frput e
- 5. Calibrate:
1:. Cetibrate Knewn resistence ssbstituted foe stD coinct h t with (1)
(1) Temerature known pressure and pw f rp;t.
I'iput (2) Port of Ste) belcme.
(2) Pressure irpat M
- c. Bistable trip tester.
C. Test l
- e. Comerlson of four seperste pressure indicatiere,
- e. check
- 5 S.
High-Pressuriter Pressure A
tu h pressure egetled te eersors, M*
- c. Sistable trip tester.N j
- b. Stibrate
- c. Test J
l 43 aue damt so.38. 44, JJr. 73s, f34 150 I
I l
w y
H h
'~'
TutE %.1.1 (contirwed)
Miniamm Frequercles for Checks, Cal *bratiers and Testing of Reactor Protective Syst=m Surveillarce Sunelltence Net W Chewt Dascription Foretien Freemecy
- s. Check Q
- e. Ve-ify constants.
14 Thermal T cgin Calculat,r CCTF5; i
(1) The bistable trip tester injects a signet It to the bistable and prevides e precisicn resdout of the trip set po nt.
(2) Att emnthiy tests will be oone on only one of four cherwwts st a tine to prevent reacter trio.
(3) Adju=t the meteer power or fai power until readout egrees with heet belance calculations sen above 15% c' rated power.
.I (4) Deleted ti (5) It is not nacessary to perform the specified testbeg d.rirs pretenged periods in the refuellrg shutdown cauf t on if this occurs, omitted testing wilt be cerformed pelor to returning the ptact to service:
(6) Also includes testirs variab4, high power tation Ir. the Thersel Percin Calculator.
(?) Required if t! e. <a ' tor is critiest.
I
(;') Required when PCS is >1500 psia.
FRECPJETCT sotetion 8
Frerum betat :n At least om.e per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ~
s 4t least cree eer 24 hears.
J At least once per i days.
b At f eest erae prr 31 days.
M At least orce per 92 deys.
O At least ence per 6 men'.hs.
SA At least once per 18 months.
R Prler to esca start-p if ret done P
previous teeek.
kA Wat ervilcable.
l 4-5 Ammckumt no 11,118,119,15g j
I i
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