ML20206D848
| ML20206D848 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/14/1988 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20206D851 | List: |
| References | |
| NUDOCS 8811170301 | |
| Download: ML20206D848 (25) | |
Text
_
- ysa uer A
UNITED STATES 8.
NUCLEAR REGULATORY COMMIS$10N W ASHINGTON, D. C. 20684 E N Q ER$ POWER COMPANY f
PALISADES PLANT DOCKET NO. 50-255 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment Nn 117 License No. OPR-20 t
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Consumers Power Company (the licensee) dated December 22, 1987, as revised April 12, 1988, complies with the standards and requirements of the Atomic Energy l
Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; I
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted 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.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have M satisfied, s
P
i i
i 2.
Accordingly, the Itcense is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B. of Provisional Operating License j
No. DPR-20 is hereby amended to read as follows:
j Technical Specifications
[
t The Technical Specifications contained in Appendices A and B,
{
as revised through Amendment No.117, are hereby incorporated in the license.
The licensee shall operate the facility
}
in accordance with the Technical Specifications, j
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0694!SSION
(
-d4 w
%p Theodore Quay, Acting Director Project Directorate !!!-1 Division of Reactor Projects - !!!, IV, V
& Special Projects l
l Attachment
- Changes to the Technical l
Specifications j
Date of Issuance: November 14, 1988
(
a f
l 4
e
ATTACNMENT TO LICENSE AMENDMENT NO.117 PROVISIONAL OPERATING LICENSE NO. DPR-20 a
DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the captioned amendment n' ;ber and contain marginal lines indicating the area of change.
REMOVE INSERT 3-Ic 3-Ic 3-3 3-3
'.-4 3-4
- 5 3-5 36 3-6 37 3-7 3-1 3-8 3-5 3-9 3-10 3 10 3-11 3 11 3-12 3-12 l
l 3-25a 3-25a 3-25b 3-30 3-30 3-31 3-31 3-32 3-32 3-33 3*3 a
3-33a 4-2 4-2 4-39 4-39 4-40 4-40 4-41 4-41 d
3.1 PRIMARY COOI. ANT SYSTEM (Continued) 3.1.1 Operable Components (Continued)
(2) ';fdrostatic tests shall be conducted in accordance with applicable paragtephs of Section XI ASME Boiler to Pressure Vessel Code (1974). Such tests shall be conducted with sufficient pressure on the secondary side of the steam gene.ators to restrict primary to secondary pressure differential to a inanteus of 1380 psi. Maximus hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is sontaal operating pressure.
(3) Primary side leak tests shall be condacted at normal operettas pressure. The toeperature shall be consistent with applicable fracture toughness criteria for ferratic saterists and shall be selected such that the differtatial pressure across the steam generator tubes is not greater than 1380 psi.
(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A staisua toeperature of 100*T 14 i
required. Only ten cycles are permitted.
(5) Maxir.us seconda'y leak test pressure shall act exceed 1000 psia. A etsimum temperature of 100'T is required.
(6) In performing the tests identified in 3.1.1.e(4) and 3.1.1.e(5), above, the secondary pressure shall not exceed the primary pressure by morc than 350 psi.
f.
Nominal primary system operation pressure shall act exceed 2100 psia.
3 The reactor inlet tempertture (indicated) shall act exceed the value gives by the following equattoa at steady state 1001 power operation:
l r,1,, 3 33 0 e 0.0393. (,- us0) e 0.00004:43 c, 2060)2 e 1.0342 (v i20.2) i T,g,g a reactor talet toeperature la T*
Where:
g P = acetaal operating pressure in paia 6 W = total recirculatias mass flew ta 10 lb/h corrected to the operating temperature conditions.
Note: This equattom to shown in Figure 3 0 for a vgriety of mass flew rates.
h.
During initial pristry coolant pump starts (i.e., taittation
/
of forced circulation), secondary systes toeperature ta the
/
steam gaaerators shall be 3 the PCS cold les temperature
/
ualess the PCS cold les teeperature is 1 450'T.
/
3-Ic Aseadsent No.11,1f.85,117
3.1 PRIMARY C001. ANT SYSTI.M (Continued) allcw3d duries tsrual cporatios, so that substantial safety car 61n exists betseen this pressare differential and the pressure i
differential required for tute rupture.
t Secondary side hydrostatic and leak testing requirements are consistent with ASE SPV Section XI (1971). The differential maintains stresses in the steam g(nerator tube walls within code allowable stresses.
{
The minimum temperature of 100*T for pressurinns the steam generator secondary side is set by the NDTT of the aanway cover of 7
~
+40'T.
v The transient analyses vera performed assuming a vessel flow at l
0 bot zero power (532*T) of 126.9 x 10 lb/h minur 3}oaccountfor
(
flow measurement uncertainty aco core flow bypass A steady state DNB analysis was also performed (assuming 115% overpower, 50 psi for pressure uncertainty, 3% for flow seasurement uncertainty, and 3% for core flow bypass) in a parametric fashion j
to determine the core inlet temperature as a function of pressure to 1.30.(Gr which the nintaus DNBR at 115% overpower is equal and flow The result of tt',is steady state DNB analysis was the following equation for limiting reactor inlet temperature:
T,g g 1 541.0 + 0.03938 (P 2060) + 0.00004843 (P-2060)2,
g 1.0342 (W-120.2)
)
i A temperature seasurement uncertainty of 3'T was subtracted from this limit in arriving at the I,CO given in Section 3.1.1 3 The nominal full p'ver inlet temperature is 2'T less than the value given in Section 3.1.1 3 to allow for drift within the temperature control band. Thus, a tott.1 uncertsinty of 5'T is applied to the limiting reactor inlet temperature equation. The limits of validity of this equation are l
i s
1850 i Pressure 1 2250 psia 0
I 110.0 x 10 $ Vessel Flow i 130 x 10 lb/h f
The requirement that the steam generator temperature be i the PC5
/
temperature when forced circulation is initiated in the PCS
/
l ensures that an energy addition caused by heat transferred f ree
/
the secondary system to the PCS will not occur. This requirement
/
l applies only to the initiation of forced circulation (the start
/
of the first primary coolant pump) when the PCS cold les
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j tosperature is < 450'T. At or above 450'T, the PCS safety valves
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i prevent the PCS pressure froe exceeding 10CTR,50 Appendix 0 limits.
/
f References
{
() T$AR, Sections 6.1.2.2 and 14.3.2.
(3) XX NT-77 18.
I (2) TSAR, Section 4.3.7 (4) XW-NT 77 22.
(5)
(Deleted)
/
i l
l 3-3 Amendment No. JI, II, 117 t
L 3:1 PRI'tARY CCOLANT SYSTDi (Cost'd) t 3.1.2 Heatup and Cooldown Rates t
The primary coolant pressure and the system bestup and cooldown rates shall be limited in accordance with Tigura 3 1, Tigure 1 2 and as follows.
i a.
Allowable combinations of pressure and temp'.rature for awy heatup i
rate shall be below and to the right of the applicable iteit line as /
[
shown on Tigure 3 1.
The iverage haatup rate la any one hour ttee
/
l' l
period shall not exceed the heatup rate limit when one or more PCS
/
cold les is less tnan the corresponding ' Cold Les Temperature"
/
below.
/
/
- Cold Lea Tempsrature Heatup Rate Liett
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/
J 1.
< 190'T 20*T/Hr
/
2.
I 190'T and < 310*T 40*T/Hr
/
r 3.
> 310'T and i 450'T 60'T/Hr
/
i 4.
> 450*T 100*T/Hr
/
l t
b.
Alliwable combinations of pressure and tesperature for any cooldown
}
rate shall be below and to the rig' t of the applicable limit lines
/
[
a as shown on Figure 3 2.
The average cooldown race shall act exceed j
the Cooldown Rate Limit when one os more PCS cold legs to less than
/
the co.* responding "Cold Les Terperature* below:
/
7
/
- Cold Les Temperature Cooldown Rate Limit
/
/
1.
> 450*T 100*T/Hr
/
2.
I 3J0*T and ( 450*T 60'T/Hr
/
3.
< 300'T and >180'T 40*T/Mr
/
[
4.
I 180*T 20*T/Mr
/
l I
c, Allowable combinations of pressure and temperature lor inservice testing during bestup are as shown in Figure 3-3.
The 34xinua
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i heatup and cooldown rates required by Sections a and b, above,
/
are appitcable.
Interpolation between limit lines for other than
/
the noted temperature change rates is permitted in 3.1.2a. b or c.
/
d.
1.
The average heatup rates for the pressuriser shall not exceed
/
100'T/hr in any one hour time period when the PCS cold les
/
temperature is less than 450'T.
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2.
The average cooldown rate for the pre ssuriser shal} not exceed
/
200'T/br for any one hour time period.
/
- Vse shutdown cooling return temperature if the shutdown cooling system is in
/
operation and all PCP's are off.
/
3+4 Amendment No. II.4 U U I. III
3.1.2 Heatus and C %1down Rates (Cest'd) e.
Before the radiation exposure of the reactor vessel exceeds the exposure fcr which the figures apply. Tigures 31, 3-2 and 3 3 shall be updated in accordance with the following criteria and procedure:
1.
US Nuclear Regulatory Commission Regulatory Guide 1.99 has beert used to predict the increase in transition tesperature based on integratad fast neutron flux and surveillance test data.
If sessurements on the irradiated specimens show increass above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data j
points.
2.
Before the end of the integrated power period for which Tigures 3 1, 3 2 and 3 3 apply, the limit lines on the figures shall be updated for a new integrated power period.
l The total integrated reactor thermal power froe start-up to the end of the new power perico shall be converted to an equivalent integrated fast nutron exposure (f. t 1 MeV).
Such a conversten shall be made consistent with the dost. netty evaluation of capsule W 290(12),
i 3.
The limit lines in Tigh*es 3-1, 3 2 and 3 3 are based on the i
requirements of Eeference 9, Paragraphs IV.A.2 and IV.A 3.
These lines reflect a preservice hydrostatic test pressure of 2400 psig and a vessel flange material reference tesperatu e of 60'T(0).
Basis 4
]
All components in the priz.:ry coolant systes are designed to withstand j
the ef fects of cyclic loads due to primary syntes temperature and j
pressure cnanges.IU These cyclic loads are introduced by normal unit load transients, reactor trips and start up and shutdown operation, j
Dursos unit start up and shutdown, the rates of temperature and pressure changes are Itaited. A naximum plant heatup and cooldown rate of 100*F per hour is consistent with the design number of cycles and sattsites t ess liaits for cyclic operation.(2) 1 sector vessel plate and material opposite the core has been seed to a specified Charpy V Notch test result of 30 ft lb
- eater at an WDTT of + 10*T or less. The vessel weld has the s ent Ri of plate, weld and MAZ satorials at the flueiice to g
which the Tigures 3 1, 3 2 and 3 3 apply.
The'unirradiated RTNDT has been determined to be -56'T.IIU Aa RT of 56'T is used as an
/
NDT untiradiated value to which tiradiation effects are added.
In addition, 35 Amendment No. 21,31,33,H.91,ll?
3.1.2 Heatup and Cesidown Patee (Continued) the plate has been 100% volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system compor tats, meets the appropriate design code requirements and specific component function and has a maximum NDTT of s40'T. 0)
As a result of fast neutrun 1~ radiation in this region of the core, there vill be an increase in the RT with operation. The techniques used to predict the integratad fast neutron (E > 1 MeV) fluxes of the L
reactor vessel are describe ( in Section 3.3.2.6 of the TSAR and also l
in Amendment 13. Section 11. to the TSAR.
Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift from a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude.
The maximum exposure of the reactor vessel vill be obtained from r
the measured sample exposure by applitation of the calculated azimuthal neutron flux variattoa. The predicted RT shift for g
the base metal has been predicted based upon surveillance data and the US NRC Regulatory Guide.(
}
To compensate for any increase in the RT caused by irradiatien. limits on the pressure-tempera' ara j
relationship are periodically changed to stay within the stre9 l
limits during heatup and cooldown.
l l
l Reference 7 provides a procedure for obtaining the allowable i
loadings for ferritic pressure-retaining materials in Class I eceponents. This procedure is based en the principles of linear j
elastic fracture mechanics and involves a stress intensity factor j
predictionwhichisalowerboundofstatic,dynamicandcrach) r arrest critical values. The stress intensity factor computed t
l temperaturegradienk,operatingtemperature,andvesselvall
{
1s a function of RT i
Pressure-temperature limit calculationel procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference 7.
The limit lines of Figures 3-1 through 3-3 l
consider a 54 pst r,ssure allevance to account for the fact that t
pressure is measured in the pressuriser rather than at the vesse!
beltline. In addition, for calculattenal purposes. S'T and 30 psi were taken as measurement error allowances for temperature end l
pressure, res,$ectively. Py Reference 7. reactor vessel vall l
locations at 1/4 and 3/4 thickness are limiting.
It is at these locations that the crack propagation associated with the hypothetical flav must be arrested. At these locations, fluence i
attenuation and thereal gradients have been l
[
3-6
[
Amendment No. U.U.33.U.U.117 l
p-
-q.g gm.am--e--
%.---_r.,--,----__--.,%,,,e
,-_-,.,__.,r-yv-y
---'--*--w---
3.1.0 Heatup and Cooldevn Rates (Cont'd)
"ists (Cent'4) evaluated. During cooldovn. the 1/4 thickness locatien is always more 1tmiting in that the RT is higher than that at the NDT 3/4 thickness location and thernal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate.
Tigures 3-1 through 3-3 define stress limitations only from a f racture sechanics' point of view.
Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other inherent plant characteristics may limit the heatup and cooldown rates which can be achieved. Puep parameters and pressuriser heating capacity tends to restrict both normal heatup and cooldown rates to less than 60'T per hour.
The revised pressure-temperature limits are appiteat te to reactor vessel inner vall fluences of up to 1.8 x 10 'nyt.
The applicatien l
of appropriate fluence attenuation facters (Referesce 10) at the 1/4 and 3/4 thickness locations results in P.T.DT ehifts of 241'T 3
and 183'T. respectively, for the limiting veld esterial. The criticality condition which defines a tesperature belov vhich the core cannot be made critical (strictly based upon fracture techanics' considerations) is 371*r'.
The most limiting vall location is at 1/4 thickness. The minimur criticality temperature. 371*T is the minimum pernissible temperature for the inservice system hydrostatic prwssure test. That te=perature is calculated based upon 2310 psis inservice hydrostatic test pressure.
The restriction of average heatup and cooldown rates to 100*T/h
/
vhen all PCs cold legs are 1 450'T and the maintenance of a
/
pressure-temperature relationship under the heatup, cooldevn and inservice test curves of Tigures 3-1. 3-2 and 3-3 respectively, ensures that the requirements of References 6. 7. 8 and 9 are set.
The core operational limit applies only when the reactor is critical.
l The heatup and cooldown rate restrictions applicable when the
/
temperature of one or more of the PCS cold legs is less than 450'T
/
are coasistent with the analyses performed for lov temperature
/
overpressure protection (LTOP) (References 13. 14. 15, 16 & 17).
/
At 450'T or above, the TCS safety valves provide overpressure
/
protection for heatup or cooldown rates 1 100*T/hr.
/
3-7 Amend:ent No. 21.dl.55.f f.91.117
3;1.2 Heatup and Ceoidovn Rates (Continued)
Basis ' Continued)
The criticality temperature is deter 1sined per Refe7ence 8 and the core gerational curves adhere to the requirements of Reference 9.
'r>e inservice test curycs incorporate allevances for the thermal gradients associated with the heatup curve used to attain inservice test pressure. Theso curvas differ from heatup curves only with l
respect to margin for primary membrane stress.I ) Due to the shifts ET. NDTT requi'.' aments samociated sith nonreactor vessel in RT materials are, for all pract. cal purposes, no longer limiting.
d i
References l
(1)
TSAR. Section 4.2.2.
(2) ASME Boiler and Pressure Vessel Code.Section III. A-2000.
(3) Battelle Columbus Laboratories Report. "R11sades Pressu're Vessel Irradiation Capsule Program: Unitradiated Mechanical Properties." August 25, 1977.
(4) Battelle Columbus Laboratories Report. "Palisades Nuclu r Plant j
Reactor Vesssl Surve.11ance Progran* Capsule A-240." Harch 13 1979, submitted to the NRC by Consumers Power Company letter dated July 2. 1979, i
(5) TSAR. Section 4.2.4.
(6) US Nuclear Regulatory Commission. RegulatoryGuide 1.99 "Effects of Residual Elemen's on Predicted Radiation Damage to Reactor Vessel Haterials."."tly. 1975.
(7) ASME Boiler and Pressure Vescel Code.Section III. Appendix C.
"Protection Against Non-Ductile Taflure." 1974 Edition.
(8) US Atomic Energy Coensission Standard Review Plan. Directorate i
of Licensing. Section 5.3.2. "Pressure-Teeperature Limits."
l (9) 10 CTR Part 50. Appendix C. "Tracture Toughness Requirements."
t Hay 31. 1983.
(10) US Nuclear Regulatory Commission. Regulatory Guide 1.99. Draft j
Revision 2. April, 1984 (11) Combustion Engineering Report CEN-189. December, 1981.
l (12) "Analysis of capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance t
Program." WCAP-10637. September, 1984.
(13) EA-PAL-85-101 "Cair:ulation of TCS Pressure Increase from Adding
/
I 133 spa (3 charging pumps) Before the PORVs Open." November 4
/
1987.
/
[
(14) EA-PAL-LTOP-880119 "Calculation of Required PORY Capac!.ty to
/
l Faintain the PCS Below Appendix C." January 19, 1988.s
/
l (15) EA-PAL-LTOP-880120 Rev. A - PORY T1ow Capacity at Expected
/
l LTOP Conditions" February 15. 1988.
/
l (16) EA-PAL-LTOP-880121 "Calculation of Time for Operator to Act
/
l for HPSI and Bubble" - January 20, 1987.
/
(17) ZA-ESSR 88727-C/5-01 "Palisades Plant Primary Coolant System
/
l Pressure Te=perature Limits Per Appendix C of the ASME Boiler
/
and Pressure Vessel Cod."
Revision O.
/
(
3-8 i
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E l
3.1.3 Minimum conditions for Criticality a) Except during low-power physics test, the reactor chall not be made critical if the primary coolant temperature is below 525'F.
b) In no ' case shall the reactor be made critical if the primary coolant temper &ture is below 371*F.
c) When the primary coolant temperature is below the minimum temperature specified in "a" above, the reactor shall be suberitical by an amount equal to or greater than the potential reactivity insertion dua to depressurization.
d) No more than one control rod at a time shall be exercised or withdrawn until after a steam bubble and normal water level are established in the pressurizer.
e) Primary coolant boron concantration shall not be reduced until after a steam bubble and normal water level are established in the pressurizer.
Basis At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be sli h ly negative at operating temperatures with all control rods withdrawn.
However, the uncertafnty of the calculation is such that it is possible that a slightly positive coefficient could exist.
The moderator coefficient at lower temperatures will be less negative or more positive than at operati'g temperature.II'2) It is, t'herefore.
3-12 Amendment No. 27,4),5),5), 97, 117 t
3.1.8 OVERPRESSURE PROTECTION SYSTEMS
/
/
l LIMITING CCNDITIONS FOR OPERATION
/
/
3.1.8.1 REQUIREMENTS
/
/
a.
When the temperature of one or more of the primary coolant..
/
system cold legs is 1 300*F, or whenever the shutdown cooling
/
isolation valves (MOV-3015 and MOV-3016) are open, two power
/
operated relief valves (PORVs) with a lift setting of 1 310 psia
/
shall be operable, or a reactor coolant system vent of
/
1 1.3 square inches shall be open, or both PORV pilot valves
/
and both PORV block valves shall be open.
/
/
b.
When the temperature of one or oore of the primary system cold
/
legs is < 430*F, two power operated relief valves (PORVs)
/
with a lift setting of 1575 psia shall be operable except
/
as specified in section c. below.
/
/
c.
When a tubble is formed in the pressurizer and the actual
/
pressurizer level is < 60 percent and the temperature of all
/
the primary coolant system cold legs is > 385'F, PORV
/
operability is not required.
/
/
APPLICABILITY: When the temperature of one or more of the
/
primary coolant system cold legs is less than 430*F.
/
/
ACTINJ:
/
/
a.
With one PORV inopera',.e, either restore the inoperable PORV to
/
^
operable status within 7 days or depressurize and within the
/
next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either vent thJ PCS through a 1 1.3 square inch
/
vent or open both PORV pilot valves and both PORV block valves.
/
/
/
b.
With both PORVs inoperable, depressurize and within the next
/
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either vent the PCS through a 1 1.3 square iach vent or
/
open both PORV pilot valves and both PORV block valves.
/
/
The provisions of Specifications 3.0.3 and 3.0.4 are not
/
c.
applicable.
/
Basis
/
/
There are three pressure transients which could cause the PCS
/
pressure to exceed the pressure limits required by 10CFR50
/
Appendix G.
They are:
(1) a charging / letdown imbalance, (2) the
/
start of a high pressure safety injection (NPSI) pump, gnd
/
(3) initiation of forced circulation in the PCS when the steam
/
generator temperature is higher than the PCS temperature.
/
/
Amendment No. 51,1/, II7 3-25a i
p
'3.1.8 OVERPRESSURE PROTECTION SYSTEMS
/
/
1.IMITING CONDITIONS FOR OPERATION
/
/
/
3.1.8.
Basis (continued)
Analysis (Reference 1, 4 & 5) shows that when three charging pumps
/
are operating and letdown is isolated and a spurious HPSI occurs.
/
the PORV setpoints ensure that 10CFR50 Appendix G pressure limits
/
vill not be exceeded. Above 430*F the pressurizer safety valves
/
prevent 10CFR Appendix G limits from being exceeded by a charging /
/
letdown imbalance (Reference 2).
/
/
The requirement that steam generator temperature be < the PC3
/
comparature when forced circulation is initiated in the PCS
/
ensures that an energy addition caused by heat transferred from the
/
secondary system to the PCS will not occur. This requirement
/
applies only to the initiation of forced circulation (the start of
/
the first primary coolant pump) with one or more of the PCS cold
/
leg temperatures < 450'F.
/
/
Requiring the PORVs to be operable vnen the shutdown cooling
/
system is not isolated (MO-3015 and MO-3016 open) from the PCS
/
ensures that the shutdown cooling system will not be pressurized
/
above its design pressure.
/
/
The requirement for the PCS ta be depressurized and vented by an
/
opening >,1.3 square inches (Reference 3), or by opening both
'/
PORV pilot valves and both PORY block valves when one or both
/
PORVs are inoperable ensures that the 10CTR50 Appendix G pressure
/
limits vill not be exceeded when one of the PORVs is assumed to
/
fail per the "single failure" criteria 10CTR50 Appendix A.
/
Criterion 34.
/
Section 3.3.2.g(6) requires a dedicsted operator when the PORVs
/
are inoperable as allowed in Section 3.1.8.1.c.
Analysis justifying
/
this condition is referenced in the Besis for Se: tion 3.3.
/
I References
/
/
1.
EA-PAL-85-101 "Calculation of PCS Preosure
/
Increase From Adding 133 spm (3 chargine pumps) Before the
/
I PORVs Open," November 4, 1987.
/
2.
Technical Specificati3. 3.1.2.
/
3.
"Palisades Plant Overpressurization Analysis," June 1977 and
/
"Palisades ?lant Primary Coolant System OverpressuTiration
/
Subrystem Description," Octooer, 1977.
/
4.
EA-PAL-LTOP-880119 "Calculation of Required PORY Capacity
/
to Maintain the PCS Below Appendix G Curves." January 19, 1908.
~/
5.
EA-PAL-tTOP-880120 "Palisades LTOP PORY Flovrate Capacity
/
When PCS Temperature is 300'T or Greater." January 20, M88.
/
/
)
3-25b Amendment No.117
3'. 3 EMERGENCY CORE COOLING SYSTEM (Continu:d) 3.3.2 g.
HPSI Pump operability shall be as follows:
//
1)
Both HPSI Pumps shall be rendered inoperable wherever
/
PCS temperature is < 300'F unless the reactor vessel
/
head is removed.
/
/
/
Temperature is 3 300'F but < 350'F.
/
/
?) One, and only one, HPSI Pump shall be operable whenever
/
PCS temperature is 3 350'F but < 430*F.
/
/
/
temperature is 1430'F but <460*F.
/
/
/
temperature is 3 460*F.
/
/
- 6) One HPSI pump may be made inoperable when the reactor
/
is suberitical and the PCS temperature is 1 460'F,
/
provided the requirements of Section 3.3.2.c are met.
/
/
- 7) Whenever PCS temperature is between 385'F to 430'F and
/
LTOP system is not armed, then a dedicated licensed
/
operator shall be stationed in the control room to
/
terminate an inadvertent HPSI Pump statt and stop Charging
/
Pumps as necessary to limit PCS pressure.
/
/
- 8) Safety Injection Actuation System (SIAS) testing shall not
/
be performed while the PCS is between 300'F and 430'F.
/
HPSI pump testing may be performed below 430'F provided
/
the HPSI pump manual discharge valve is closed.
/
3.3.3 Prior to returning to the Power Operation Condition af ter every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and testing of Specification 4.3.h has not been accomplir.hed in the previouc 9 months, or prior to returning the check valves in Table 4.3.1 to service after maintenance, repair or replacement, the following conditions shall be mett a.
All pressure isolation valves listed in Table 4.3.1 shall be functional as a pressure isolation device, except as specified in b.
Valve leakage shall not exceed the amounts indicated, b.
In the event that inteprity of any pressure teolation# valve specified in Table 4.3.1 cannot be demonstrated, at least two valves in each high pressure line having a non-functional valve must be in and remain in, the mode corresponding to the isolated condition.(I) c.
If Specificationsa and b. cannot be met, an orderly shutdown shall be initiated and the reactor shall be in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3-30 Amendment No. 51, 101, 117 i
I
- 3.3 EMERGENC*. CORE C00LENG SYSTEM (Con 2'd)
Basis The normal procedure for starting the reactor 1.=,
first, to heat the primary coolant to near operating te=perature by running the prinary coolant pumps. The reactor is then made critical by withdrawing control rods and diluting boron in the primary coolant.II) With this mode of start-up, the energy stored in the primary coolant during the approach to criticality is substantially equal to that during power operation and, therefore, all engineered safety features and auxiliary cooling systems are required to be fully operable.
During low-temperature physics tests, there is a negligible amount of stored energy in the primary coolant; therefcre, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards' systems are not recuired.
The SIRW tank contains a mini =um of 250,000 gallons of water containing 1720 ppm boron. This is sufficient boren concentration to provide a 5% shutdown margin with all control rods withdrawn and a new core at a temperature of 60'F.
Heating steam is provided to maintain the tank above 40*F to prevent freezing.
The 1: boron (1720 ppm) solution vill not precipitate out above 32*F.
The source of steam during normal plant operation is extraction steam line in the turbine cycle.
The limits for the safety injection tank pressure and volu=e assure the required amount of water injection during an accident and are based on values used for the accident analyses.
The =inimum 186-inch level corresponds to a volume of 1103 ft and *.te maximum 198-inch level corresponds to a volume of 1166 ft Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked.
Since the system is used for shutdown cooling, the valving vill be changed and cust be properly aligned prior to start-up of the reactor.
The operable status of the various systems and components is to be demonstrated by periodic tests. A large fraction of tilese tests vill be performed while the reactor is operating in the power range.
If a component is found to be inoperable, it vill be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. For a single component to be inoperable does not negate the ability of the system to perform its function, but it reduct.s The redundancy provided in the reactor design and that by limits the I Motor-operated valves shall be placed in the closed position and power supplies deenergized.
3-31 CKddid/t6l/l Amendment No. Il7
'3.3 EMERGENCY CORE COOLTSC SYSTEM Basis (continued) l ability to tolerate additional equipment failures. To provide maximum assurance that the redundant cceponent(s) vill operate if required to do so the redundant component (s) is to be tested prior to initiating repair of the inoperable component.
If it develops that (a) the inoperable component is not repaired within the specified allovable ti=e period; or (b) a second component in the same or related system is fcund to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of the decay heat from the fuel and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. This vill also permit improved access for repairs in some cases. After a limited time in hot shutdown, if the malfunction (s) is not corrected.
the reactor vill be placed in the cold shutdown condition utilizing normal shutdown and cooldown procedures.
In the cold shutdown condition, release of fission products or damage of the fuel ele =ents is not considered possible.
The plant operating pro:edures will require immediate at. tion to effect repairs of an inoperable component and, therefore. in most I
cases, repairs will be completed in less than the specified allovable repair times. The limiting times to repair are intended tos (1) Assure that operability of the component will be restored promptly and yet. (2) allow sufficient time to effect repairs using safe and proper precedures.
The requirement for core cooling in case of a postulated less-
)
of-coolant accident while in the hot shutdown condition is significantly reduced below the requiremente for a postulated loss-of-coolant accident during power operation.
Putting the I
reactor in the hot shutdown condition reduces the consequences of a loss-of-:oolant accident and also allows more f ree access to some of the e agineered safeguards components in order to ef fect repairs.
Tailure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indicative of a requirement for major maintenance and, therefore, in such a case the reactor is l
to be put into the cold shutdown condition.
Withrespecttothecorecoolingfunction,there{g) functional redundancy over most of the range of break sizes.
Adequate core cooling for the break spectrum up to and including l
the 42-inch double-ended break is assured with the minimum safety injection which is defined as follows:
For the system of four passive safety injection tanks, the entire contents of one tank are assumed to be unavailable for energency core
- cooling.
In addition, of the two high-pressure safety injection pumps and the i
two low-pressure safety injection pumps, only one of each type is
[
assumed to operates and also that 25% of their combined discharge i
rate is lost from the primary coolant system out the break. The transient het spot fuel clad temperatures for the break sites considered are shown on FSAR Tigures 14.17.9 to 14.17.13. These 3-32 Amendment No. ff. I17
3.3 EMERGENCY COPE COOLING SYSTEM Basis (continued) demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below the melting temperature of zirconium (3300*F).
Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injectior feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation. This action assures that it vill not block flow during Safety Injection.
The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has not been analyzed.
To provide as,urance that this vill not occur, these valves are electrically locked open by a key switch in the control room. In addition, prior to critical the valves are checked open, and then the 480 volt breakers are opened. Thus, a failure of a b eaker and a s'vitch are required for any of the valves to close.
Insuring one HPSI pump is inoperable eliminates una.talysed PCS
/
mass additions due to inadvertent two pump starts.
Both RPSI
/
pumps starting in conjunction with a charging / letdown imbalance
/
may cause 10CFR50 Appendix C limits to be exceeded when the PCS
/
temperature is < 430'F.
When the PCS temperature is 1 430'F,
/
the pressurizer safety valves ensure that the PCS pressure vill
/
not exceed 10CFR$0 Appendix G limits when one or both HPSI
/
pumps are started.
/
The requirement to have one HPSI train operable above 350'F
/
provides added assurance that the effects of a LOCA occurring
/
under LTOP conditions vould be mitigated.
If a LOCA occurs when
/
the pri=ary system temperature is less than or equal to 350'F,
/
the pressure would drop to the level where low pressure safety
/
injection can prevent core damage.
/
Analysis (Reference 3) further shows that if the PCS temperature
/
is 2 385'T and there is a bubble in the pressurizer and the
/
actual pressurizer liquid level is < 60%, and LTOP is not armed.
/
operator action, within 2.9 minutes of the time letdown is
/
isolated concurrent with HPSI, can prevent the PCS pressure from
/
exceeding 10CFR$0 Appendix G pressure limits. A dedicated
/
operator is required under these conditions to ensure that
/
micigatory action is initiated within 2.9 minutes.
/
s 3-33 Amendment No. 21, 51, 10!,117
3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued)
HPSI pump testing with the HPSI pump manual discharge valve
/
closed is permitted since the closed valve eliminates the
/
possibility of pump testing being the cause of a mass addition
/
to the PCS.
References (1) TSAR, Section 9.10.3.
(2) FSAR, Section 6.1 (3) EA-PAL-LTOP-880121 "Calculation of Time for Operation to Act
/
for HPSI and Bubble". January 20, 1988
/
4 J
3-33a Amendment No f t, jf, ygg, ))7
- d b.
The PCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the vent (s) is tsing used for overpressure protection except when the vent pa6nvay is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once pe? 31 days.
c.
When both open PORV pilot valves are used as an alternative
/
to venting the PCS, then verify both PORV pilot valves and
/
both PORV block valves are open at least once per 7 days.
/
Basis Tailures such as blown instrumer.t fuses, defective indicators, and faulted amplifiers which recult in "upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action and a check supplements this type of built-in surveillance.
Based on experience in operation of both conventional aad nuclear i
plant systems when the plant is in operation, a checkffs frequency i
of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to insure the presentation and acquisition of accurate information.
The power range safety channels are calibrated daily against a heat balance standard to account for errors induced by changing red patterns and core physics parameters.
Other channels are subject only to the "drift" errors induced within j
the instrumentation itself and, consequently, can tolerate longer l
intervals between calibration. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at each refueling shutdown int e rval.
Substantial calibration shif ts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies of one-per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.
l t
The minimum testing frequency for those instrument channels connected to the reactor protective system is based on an estimated average
(
unsafe failure rate of 1.14 x 10-3 failure / hour per channel. This estimation is based on limited operating experience at conventional and nuclear plants. An "unsafe failure" is def'ined as one which negates channel operability and which, due to its nature, is revealed i'
only when the channel is tested or attempts to respond to a bonafide signal.
i i
4-2 Amendment No. II, II,II7 r
,,,. - _ _. - _ _ _ - _ - _ _. _ _ _. _ _.. _ _ _ _ _ _ _. _ _ _. _ _ _ _. _ ~. _ - _ _. _ _ _ _ _ _ _ _ - _ _
.e i.
4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS Applicability Applies to the safety injection system, the containrent spray system, chemical injection system and the containment cooling system tests.
Objective To verify that the subject systems will respond promptly and perform their intended functions, if required.
Specifications 4.6.1 Safety Injection System System tests shall be performed at each reactor refueling a.
interval. A test safety injection signal vill be applied to initiate operation of the system. The safety injection and shutdown c9oling system pump motors eay be de-energized for this test. The system vill be considered satisfactory if
/
control board indication and visual observations indicate that all components have received the safety injection signal in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel).
b.
Both high pressure safety injection pumps. P-66A and P-663
/
shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
/
whenever the temperature of one or more of the PCS cold legs
/
is < 300'T unless the reactor head is removed.
/
+
4.6.2 Containment Spray System System test shall be performed at each reactor refieling a.
int e rval. The test shall be performed with the isolation valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation.
b.
At least every five years the spray nostles shall be verified to be open.
l c.
The test will be considered satisf actory if visual l
l observations indicate all components have operated satisfactorily.
s 4-39 Amendment No. 31.73.96.117
4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Contd) 4.6.3 Pumps a.
The safety injection pumps, shutdown cooling pumps, and containment spray pumps shall be started at intervals not to exceed three months. Alternate manual starting between control room console and the local breaker shall be practiced in the test program.
b.
Acceptable levels of performance shall be that the pumps start, reach their rated shutoff heads at minimum recirculation flow, and operate for at least fifteen minutes.
4.6.4 Valves Deleted 4.6.5 Containment Air Cooling System a.
Emergency mode automatic valve and fan operation vill be checked for operability during each r4 fueling shutdown.
b.
Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months.
Basis The safety injection system and the containment spray system are principal plant safety features that are normally inoperativo during reactor operation.
Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes containment isolation and a containment spray system test requires the system to be temporarily disabled. The method of assuring operability of these systems '.s therefore to combine systems tests to be parformed during annual plant shutdowns, with more frequent component tests, which can be performed during reactor operation.
The annual systems tests demonstrate proper automatic operation of the safety injection and containment spray systems. A test signal is applied to initiate automatic action and verification made that the compouents receive the safety injection in the proper sequence.
The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.
(1, 2) 4-40 Amendment No 59,7J.77. 117
4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS (Contd)
Basis (continued)
/
During reactor operation, the irstrumentation which is depended on to initiate safety injection and contain=ent spray is generally checked daily and the initiating circuits are tested monthly.
In addition, the active coeponents (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judg ent that more frequent testing would not significantly increase the reliability (ie, the probability that the component would operate when required). yet more frequent test would result in increased wear over a long period of time. Verification that the spray piping and nozzles are open vill be made initially by a smoke test or other suitably sensitive method, and at least every five years thereafter.
Since the material is all stainless steel, normally in a dry condition, and with no plugging mechanism available, the ratest every five years is considered to be more than adequate.
Other systems that are also important to the emergency coolinr function are the SI tanks, the component cooling system, the service water system and the containment air coolers. The SI tanks are a passive safety feature.
In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance.
When the PCS cold leg temperature is less than 300'F. the start
/
of ors HPSI pump could cause the Appendix G limits to be
/
exceeded; therefore, both pumps are rendered inoperable.
/
References (1) TSAR. Section 6.1.3.
(2) TSAR. Section 6.2.3.
s 4-41 Amendeent No. 117
__