ML20070D767
| ML20070D767 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/20/1991 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070D771 | List: |
| References | |
| NUDOCS 9103010159 | |
| Download: ML20070D767 (2) | |
Text
. _ _ - _ _ _ _ - - - _ _ _ _ - _ _
p1l880g j#
'k UNITED STATES j
.g NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D. C,20$55 S
e g
CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.137 License No. DPR-20 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The applications for amendment by Consumers Power Company (the licensee) dated August 31, and September 19, 1990 as amended October 3 and December 28,.1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended-(the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter 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 (i) 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 comon
- 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 Comission's regulations and all applicable requirements have been satisfied.
- 2..
Accordingly, the-license 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 No. DPR-20 is hereby amended to read as follows:
9103010159 910220 ADOCK0500g5 PDR P
. _ =
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.137, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f
7%4 L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: February 20, 1991
l i
ATTACHMENT TO LICENSE AMENDMENT NO. 137 PROVISIONAL OPERATING LICENSE NO. DPR-20 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 amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT l-2 1-2 2-1 2-1 2-2 2-2 2-4 2-4 2-6 2-6 2-9 2-9 3-Ib 3-lb 3-Ic 3-Ic 3-2 3-2 3-3 3-3 3-3a 3-3a 3-61 3-61 3-63 3-63 3-64 3-64 3-67 3-67 3-107 3-107 3-111 3-111 4-84 4 84
~..
l.1 REACTOR OPERATING CONDITIONS (Contd)
Low Power Physics Testina Testing performed under approved written procedures to determine control rod worths and other core nuclear properties.
Reactor power during these tests shall not exceed 2% of rated power, not including decay heat and primary sy' stem temperature and pressure shall be in the range of 260'F-to 538 F and 415 psia to 2150 psia, respectively.
Certain deviations from normal operating practice which are necessary to enable performing some of these tests are permitted in accordance with the specific provisions therefore in these Technical Specifications.
Shutdown Boron Concentrations Boron concentration sufficient to provide k s 0.98 with all control radsinthecoreandthehighestworthcontEolrodfullywithdrawn.
Refuelina Boron Concentration Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 5% Ap with all control rods withdrawn).
Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.
Assembly Radial Peakina Factor - F' The assembly radial peaking factor is the maximum ratio of individual fuel assembly power to core average assembly power integrated over the total core height, including tilt.
Total Interior Rod Radial Peakina Factor-- Fl l.
The maximum product of the. ratio of individual anembly power to core average assembly power times the. highest interior local peaking factor integrated over the total core height including tilt. Local peaking is defined as the maximum ratio of the power in an. individual
. fuel rod to assembly average rod power.
1-2 Amendment No. 77, (), 5%, ET, JJS,
- 137,
, _ ~. _. _. _ _
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS - REACTOR CORE Aeolicability This specification applies when the reactor is in hot standby candition and power operation condition.
Obiective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the primary-coolant.
Specifications The MDNBR of the reactor core shall be maintained greater than or equal to the DNB correlation safety limit.
l Basis 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 temperature. 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 coefficient which would result in high-cladding temperatures and the-possibility of cladding failure. Although DNB is not an observable parameter during reactor coolant flow,e observable parameters of thermal power, primary hrough operation th the use of-a DNB. Correlation. pressure, can be related to DNB tDNB Correlations h temperature and to predict DNB and the location of DNB for axially 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 DNB. The minimum value of the DNBR, during steady-state operation normal o>erational transients, 'and antici)ated transients is limited to ONB correlation safety limit. A )NBR equal to the DNB correlation safety limit corresponds to a 95% probability at a 95%-confidence level that 2-1 Amendment No.JJS,137 i-l
.=
~
2.1 SAFETY llMITS - REACTOR CORE.(Contd)
DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.
l The reactor protective system is designed to prevent any anticipated combination of transient conditions for-primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit. The DNB correlations used -in the Palsiades safety analysis are listed in the following table.
Safety References 81mg Limil Correlation Acolicability XNB' l.-17 1
2 ANFP 1.154 4
5 The MDNBR analyses are performed in accordance with Reference 6.
References
(,
XN-NF-621(P)(A), Rev 1 i
XN-NF-709 I
d Updated FSAR Section 14.1.
ANF-1224 (
A)danuary1990 May 1989 ANF-89-192 I(
XN-NF-82-2
}, Revision 1 2-2 Amendment No. 71, J), JJS, 137
2.3 LTM1 TING SAPETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM Applicability This specification applies to reacter trip-settings and bypasses for-instrument channels.
Objective To provide for automatic protective action in the event that the principal process variables approach a saf ety limit.
Specification The reactor protective system trip setting limits and the permissible b> Tasses for the instritment channels shall be as stated in Table 2.3.1.
The TM/LP trip system monitors core _ power, reactor coolant maximun inlet temperature. (Th) c re coolant system pressure and axial shape index. The low pressure trip limit (Pvar) is calculated using the following equation.
Pvar = 2012(QA)(QR ) + 17.0(Th) - 9493
/
g where:
QR
= 0.M2(Q) + 0.588 Q s 1.0 Q = core power g
=Q Q > 1.0 rated power QA- = -0.720(ASI) + 1.028
-0.628 < AS'I < -0.100
/
-0.333(ASI) + 1.067
--0.100 I ASI < +0.200
/
= +0.375(ASI) + 0.925
+0.200 I-AS1 < +0.565
/
~
~
=- 1.085 when Q < 0.0625 The calculated limit (P
) is then compared to a .xed low pressure y
try liMt (P,gn). The auctioneered highest of these signals becomes the trip limit (Ptdp).
trip e mpared to the measured reactor:
P 18 coolant. pressure (P) and a trip signal.is generated when P-is less-than or equal to P A pre-trip alarm is.also generated when P erip.
is'less than or equal to the pre-trip setting P
+ 6P.
2-4 Amendment No. !!5,137
2,3 llMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd) flui t 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 Hioh Power - The variable high power trip (VHPT) 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 boron dilution transient).
The VHPT system provides a trip setpoint no more than a predetermined amount above the indicated core power.
Operator action is required to increase the setpoint as core power is increased; the setpoint is automatically decreased as core power decreases.
Provisions have been made to select different set points for three pump and four pump operations.
During normal plant operation with all primary coolant pumps operating, 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 state power at which a trip would be actuated is 115%, which was used for the purpose l
F of safety analysis.")
2.
Primary Coolant Systpm low Flow - A reactor trip is provided to protect the core against' DNB should the coolant flow suddenly decrease significantly.") Flow in each of the four coolant loops is determined from a measurement of pressure drop from inlet to outlet-of the steam generators. The total flow through the reactor core is measured-by summing the loop pressure drops across the steam generators and correlating this pressure sum with _the pump calibration flow curves.
The percent of normal core flow is shown in the following table:
4 Pumps 100.0%
3 Pumps 74.7%-
During four-pump operation, the -low-flow trip setting of 95%
insures that the reactor cannot operate when the flow rate is less than 93% of the nominal value considering -instrument errors. ")
2-6 Amendment No )J, JJS,137
3.3 LIMITING SAFETY SYSTEM SETTINGS - RF. ACTOR PROTECTIVE SYSTEM (Contd)
_ Basis (Contd)
-1 6.
Low Steam Generator Pressure - 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 event of excessively high steam flow. This setting was used in the accident analysis.IO}
- 7. -Containment High Pressure - A reactor trip on containment high pressure is provided to assure =that the reactor is shut down before the ingtion of the safety itijection system and containmert
- spray, 8.
Low Power Physics Te %ing - For low power physics tests, certain-tests will require-the reactor to be critical at low temperature (1 260*F) and low pressure (1 415 psia).
For these-certain tests-only, the thermal margin / low pressure, primary coolant flow and low steam generator pressure trips may be bypassed in order tM t I
reactor power can be increased for improved data acquisition._
Spe:ial operating' precautions will be in effect during these tests in accordance with approved written testing procedures. At reactor power levels below 10'I% of rated power, the *,hermal margin / low-pressure trip and low flow trip are not required to prevent fuel
/
rod themal limits f rom being exceeded.,The low steam generator-
. pressure trip is not requiredfoecause the low steam generator pressure will not allow a severe reactor cooldown, should-a steam line break-occur during these. tests.
References (1) ; ANF-90-078 Table 15.0.7-1
/:
(2)_ deleted l(3) _ Updated FSAR, Section 7.2.3.3.
(4) ANF-90-078, Section 15.0.7 /
i (5) XN-NT-86-91(P)
(6) deleted (7) deleted (8) ANF-90-078, Sectior. 15.1.5
/
(9) ANF-87-150(NP), Volume 2, Section 15.2.7 (10) Updated FSAR, Section 7.2.3.9.
'(11)- ANF-90-078. Section 15.2.1
/;
/
2-9 Amendment No 31, tyg,137
3.1 PRIMARY COOLANT SYSTEM Acolicability Applies to the operable status of the primary coolant system.
Obiective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.
Specifications 3.1.1 Ooerable Comoonents a.
At least one primary coolant pump or one shutdown cooling pump with a. flow rate greater than or equal to 2810 gpm shall be in operation whenever a char.ge is being made in the boron concentration of the primary coolant and the plant is operating in cold-shutdown or asove, except during an emergency loss of coolant _ flow situation.
Under these circumstances, the boron concentration may be increased with no primary coolant pumps or shutdown cooling pumps running, b.
Four primary coolant pumps shall be in operation whenever the reactor is operated above hot shotdown, with the following exception:
Before removing a pump from service, tTermal power shall be reduced as specified in Table.2.3.1 an; appropriate corrective action implemented. With one pump on' of service, return the pump to service within 12 hoss (return to four-pump operation) or be in hot shutdown (or-below) with the reactor tripped (from the C-06 panel, opening the 42-01 and-42-02 circuit breakers) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Start-up (above hot shutdown) with less than four pumps is not permited and power operation with less than three pumps is not permitted, The measured four primary coolant pumps operating reactor vessel c.
flow shall be 140.7 x 10 lb/hr or greater, when correc' ed to l
532'F.
d.
Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325'F.
e.-
Maximum primary system pressure differentials shall not exceed.
the following:
(1) Deleted 3-lb Amendment No JJ, S5, JJS, JJ9,137
3.1 PRIMARY COOLANT SYSTEM (Continued) 3.1.1 Doerable Comconents (Continued)
(2) Hydrostatic tests shall be conducted in accordance with applicable paragra Vessel Code (1974)phs of Section XI ASME Boiler & Pressure Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus-50 psi where Po is nominal operating pressure.
(3)- Primary side leak tests shall be conducted at normal operating pressure.
The temperature shall be consistent-with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.
(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia.
A minimum temperature of 100*F is required.
Only ten cycles are permitted.
-(5) Maximum secondary leak test pressure shall not exceed 1000 psia.
A minimum temperature of 100*F is required.
(O In performin 3.1.1.e(5), g the tests identified in 3.1.1.eJ4) andabove, the secondary pressure the primary pressure by more than 350 psi.
f.
Nominal primary system operation pressure shall not exceed 2l00 psia.
g.
The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state power operation:
T,,,,, s 542.99 +.0580(P-2060) + 0.00001(P-2060)**2 + 1.125(W-138) -
0.0205(W-138)**2 Where: T,,,,,
- reactor inlet temperature in *F P
- nominal operating pressure -in psia W
= total recirculating mass flow in.10' lb/h corrected to the operating temperature conditions.-
When the ASI exceeds the limits specified in Figure 3.0 within 15 minutes-initiatecorrectiveactionsto-restorethekSItothe acceptable, region.
Restore the ASI to acceptable values within-one hour or be at less than 701 of rated power within the following two hours.
If the measured primary-coolant system flow rate is greater than 150 M lbm/hr. the maximum inlet temperature shall be less than or equal to the T.i.,1.00 at 150 M lbm/hr.
i 3-Ic Amendment No JJ, 5J,SE, JJ7, JJS,137
3.1 PRIMARY COOLANT SYSTEM (contd) h111 (Cont'd)
The FSAR safety analysis was performed assuming four primary coolant pumps. vere operating for accidents that occur during reactor operation.
Therefore, reactor startup above hot shutdown is not permitted unless all four primary coolant pumps are operating.
Operation with three priii.ary coolant pumps is permitted for a limited time to allow the restart of a stopped pump or for reactor internals vibration monitoring and testing.
Requiring the plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdraul will not be initiated by the control room operator.
Both steam generators are required to be operable whenever the temperature of the primary coolant 11 greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.
Calculations have been performed to demonstrate that a pressure dif ferential of 1380 psi'" can be withstood by a tube uniformily thinned to 36% of its original nominal wall thickness (64%
degredation), while maintaining:
(1) A factor of safety of three between the actual pressure differential and the pressure differential required to cause bursting.
(2)
Stresses within the yield stress for inconel 600 at operating temperature.
(3) Acceptable stresses during accident conditions.
Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV section XI (1971).
The differential maintains stresses in the steam generator tube walls within code allowable stresses.
The minimum temperature of 100'F for pressurizing the steam generator secondary sido is set by the NDTT of the manway cover of + 40*F.
4 The transient analyses were performed assuming a vessel flow at hot zero power (532'F) of 140.7 x 10' lb/hr minus 6% to account for flow l
measurement uncertainty and core flow bypass. A DNB analysis was performed in a pare. metric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum 1
DNBR is equal to the DNB correlation safety limit.
This analysis l
includes the following uncertainties and allowances: 2% of rated power for power 3-2 Amendment No M, SJ, JJS, 137, 137
h 3.1 PRIMARY _ COOLANT SYSTEM (Cont'd)
Basis (Cont'd) measurement; 30.06 for ASI measurement; e50 psi for pre:4urizer pressure; 47'F for inlet temperature; and 3% measurement and 3%
bypass for core flow.
In addition, transient biases were included in the derivattor. of the following equation for limiting reactor inlet tempertture:
T m., s $42.99 +.0580(P-2060) + 0.00001(P-2060)**2 + 1.125(W-138) -
.0205(W-138)**2 The limits of validity of this equation are:
1800 s pressure s 2200 psia 100.0 x 10' s Vessel Flow s l's0 x 10' lb/h l
ASI as shown in Figure 3.0 With measured primary coolant system flow rates > 150 M lbm/hr, l
limiting the maximum allowed inlet temperature to the T,,N,ow rates.
LCO at 150 M lbm/hr increases the margin to DNB for higher PCS l
The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles.
The signal representing core power (0) is the auctioneered higher of the neutron flux power and the Delta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape annealing (Y,) and the core power constitute an ordered pair (0, Y,).
An alarm eegnal is activated before the ordered pair exceed the boundaries specified in Figure 3.0.
The requirement that the steam generator temperature be s the PCS temperature when forced circulation u ;nitiated in the PCS ensures that an energy addition caused by her,t 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) when the PCS cold leg temperature is < a10*F.
However, analysis (Reference 6) shows that under limited conditions when the Shutdown Cooling System is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature, References (1) Updated FSAR, Section 14.3.2.
(2) Updated FSAR, Section 4.3.7.
(3) Palisades 1983/1984 Steam Generator Evalur tion and Repa.r Program Report, Section 4, April 19, 1984 (4) ANF-90-078 Section 15.0.7.1 (5) ANF-90-078 (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3-3 Amendment No 31, SJ, JJ7 JJE, J)J, 137
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W3 Mod 031WW JO N0110WWJ 3-3a Amendment No. 137
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3.10 CONTROL ROD AND POWER DISTRIBUTION tIMIT$ (Contd) 4 3.10.6 Shutdown Rod limits l
l a.
All shutdown rods shall be withdrawn before any regulating a
rods are withdrawn.
b.
The shutdown rods shall not be withdrawn until normal water L
level is established in the pressurizer.
c.
The shutdown rods shall not be inserted below their exercise j
limit until all regulating rods are inserted.
5 3.10.7 Low power Physics Testino i
may be deviated fr. 3.10.1.b. low powe,r physics, testing and CRDM Sections 3.10.1.a 3.10.3 3.10.4 b 3.10.5 and 3.10.6 om during exercises if necessary to perform a test but only for the time j
necessary to perform the test.
3.10.8 Center Control Rod Misalionment The requirements of Specifications 3.10.4.1 3.10.4.a. and 3.10.5 may be suspended during the performance of hysics tests to determine the isothermal temperature coeffi ient and power coefficient provided that only the center control rod is j
misaligned and the limits of Specification 3.23 are maintained.
i Basis Sufficient control rou shall t,e withdrawn at all times to assure that the reactivity decrease from a reactor trip provides-e adequate shutdown margin. The available worth of withdrawn rods i
must include the reactivity defect of power and the failure of-a the withdrawn rod of hichest worth to insert.
The requirement i
for a shutdown maroin of 2.0% in reactivity with 4-pump operation, and of 3.75% in reactivity with less than 4-) ump is consistent with the assumptions used in tie operation,f accident conditions (including steam line brea analysis o as reported in Reference 1 and additional analysis.
Requiring boronconcentrationtobeatcoldshutdownboronconcentrationatl less than hot shutdown assures adequate shutdown margin exists to ensure a return to power does not occur if an unanticipated-cooldown accident occurs. This requirement applies to normal operating situations and not daring emerge cy conditions where it of an accident. perform operations to miti ate the consequencesminimum shu is necessary to rate of 2810 gpm, y imposing 6 B
sufficient. time is provided for the o terminate a boron dilution under asymmetric conditions.perator to For operation with no primary coolant pumps operating and a recirculating flow rate less than 2810 gpm.the increased shutdown margin and controls on charging pump operability or alternately the surveillante of the charging pumps will ensure that the acce)tance crite for an inacvertent boron. dilution event will not )e violated.bjaIhechangeininsertionlimitwithreactor l
j power shown on Figure 3-6 insures that the shutdown 3-61 Amendment No JJ, pf, 57, SS, JJS, 137 9
R n
---tww m-w s - vw + v-,-tw-si-e,re.,r.en-e,,,,e..vew. eww,i---n--w
,m.r
..r e t-m ir-isy w =-tw ev'.
.me e-y
,--v--
- ywve ee v-y+ + ++-v v
,,www-*o-w+e--*
3.10 CONTROL ROD AND POWtR DISTRIBUTION LIMITS (Continued) flitilt (Continued) margin requirements for 4-pump operation is met at all power levels.
The 2.5-second drop time specified for the control rods is the drop time used in the transient analysis.'"
l The insertion of part-length rods into the core, except for rod exercises or physics tests, is not permitted since it has been demonstrated on other CE plants that design power distribution envelopes can, under some circumstances, be violated by using part-length rods.
Further information may justify their use.
Part-length rod insertion is permitted for physics tests, since resulting power distributions are closely monitored under test conditions.
Part-lengt1 rod insertion for rod exercises (approximately 6 inches) is permitted since this amount of insertion has an insignificant effect on power distribution.
For a control rod misaligned up to 8 inches from the remainder of the banks, hot channel factors will be well within design limits, if a control rod is misaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met.
if in-core detectors are not available to measure power distribution and rod misalignments >8 inches exist, then reactor power must not exceed 75% of rated power to insure thai, hot channel :onditions are met.
Continued operation with that rod fully inserted will only be permitted if the hot channel factors, shutdown margin and ejected rod worth limits are satisfied, in the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be maintained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations permitted by Specification 3.10.7 are required in order that the control rod worth values used in the reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified.
These deviations will only be in effect for the time period required for the test being performed.
The testing interval during which these deviations will be in effect will be kept to a minimum and special operating precautions will be in effect during these deviations in accordance with approved written testing procedures.
1 3-63 Amendment No. 31, J), 57, SS, JJS, 137
1-i t
l l'
3.19 CONTROL ROD AND POWER DISTRIBUTION llMlVS (Continued)
B1111(Continued)
Violation of the power dependent insertion limits, when it is J
necessary to rapidly reduce power to avoid or minimize a l'
situation harmful to plant personnel or equipment, is acceptable 1
due to the brief period of time that such a violation would be expected to exist, and due to the fact that it is unlikely that i
j core oserating limits such as thermal margin and shutdown margin would se violated as a result of the rapid rod insertion. Core thermal margin will actually increase as a result of the rapid i
a i
rod insertion.
In addition, the required shutdown margin will most likely not be violated as a result of the rapid rod insertion because present power dependent insertion limits result 3
1 in shutdown margin in excess of that required by the safety analysis.
References (1) ANF-90-078 l
J l
J 3-64 Amendment No. $S, JJS, 137
3.1'2 tiQ9ERATOR TEMPERATURE COEFFICIENT OF REAITIVITY Aeolicability Applies to the moderator temperature coefficient of reactivity for the core.
Objective To specify a limit for the positive moderator coefficient.
Soecifications The moderator temperature coefficient (MTC) shall be less positive than +0.5 x 10 Ap/*F at s 2% of rattd power.
hiti The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the safety analysis'" remain valid.
Referent _e (1) ANF-90-078, Section 15.0.5 l
3-67 Amendment No. JJS,137 (next page is 3-69) l
TABLE 3.23-1 LINEAR HEAT RATE LIMITS No. of fuel Rods Assembly 208 216 Peak Rod 15.28 kW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, f t Peaking Factor No of Fuel Rods in Assembly 208 216 Assembly f",
1.48 1.f7 Peak Rod F',-
1.92 1.92 l
l 3-107 AmendmentNo.Sf,JJS,
i i
g E,0WER DISTRIBUTION LIMITS 3.23.2 MDIAL PEAKING FACTORi llMITING CONDITION FOR ODERATION W
A, and FT The radial peaking factors F[he fo115 wing quantity.
shall be less than or equal to the l
value la Table 3.23-2 times Thequantityis(1.0+
0.3 (1 - P)) for P a.5 and the quantity is 1.15 for P <.5.
P is the core thermal power in fraction of rated power.
APPLICABILITY:
Power operation above 25% of rated power.
ACTION:
1.
For P < 50% of rated with any radial peaking factor exceeding its limit, be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
2.
For P a 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to less than the lowest value of:
(1 - 3.33( r - 1) ) x Rated Power F
L A
T Where F is the measured value of either F, or F and F l
5 r
L istheSorrespondinglimitfromTable3.232.
B3.111 A
T The limitations on F, and F are provided to ensure that assumptions used in l
the analysis for est5blishinh DNB margin, LHR and the thermal margin / low-pressure and variable high-power trip set points remain valid during operation.
Data from the incore detectors are used for determining the a
measured radial peaking factors.
The periodic surveillance requirements for t
determining the measured radial peaking factors provide assurance that they remain within prescribed limits.
Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded.
The radial-peaking is limited to those values used in the LOCA analysis.
Since the LOCA analysis limits the magnitude of radial peaking, Table 3.23-2 explicitly contains these-limits.
3-111 AmendmentNo.gJJS, w-
4.19-EQELtDISTRIBUTIONLIMITS-4.19.2 RADIAL PEAKING FACTORS SURVEILLANCE RE0VIREMENTS A
4.19.2.1 The measured radial peaking factors (F, and FT ) obtained by using l
r the incore detection system, shall be determined to be less than or equal to the values stated in the LCO at the following intervals:
a.
After each fuel loading prior to operation above 50% of rated power, and b.
At least once per week of power operation.
4-84 Amendment No SS, JJS, 137
.