ML20213E056
| ML20213E056 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/07/1987 |
| From: | Virgilio M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213E059 | List: |
| References | |
| NUDOCS 8705140444 | |
| Download: ML20213E056 (35) | |
Text
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f jo UNITED STATES g
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NUCLEAR REGULATORY COMMISSION j
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5 IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE j
CORN BELT POWER COOPERATIVE DOCKET NO. 50-331.
DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY' OPERATING LICENSE
{
Amendment No.142 License No. DPR-49' t
1.
The Nuclear Regulatory Commission (the Commission) has found that:
4 A.
The application for amendment by Iowa Electric Light and Power.
Company, et al., dated October 31, 1986, as clarified March 20, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 1
Commission; i
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 3
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 t
been satisfied.
2.
Accordingly, the license is amended by changes to the; Technical Specifications as indicated in the attachment to this license amendment 1'
and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
f' Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.142, are hereby incorporated in the license.
The licensee shall operate the-facility in accordance with the Technical Specifications.
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May 7,1987
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The license amendment is effective as of the date of issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
?
k b%b Martin J. Virgilio, Acting Director Project Directorate III-1 3
Division of Reactor Projects - III, IV, V
& Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: May 7,1987 4
t
o ATTACHMENT TO LICENSE AMENDMENT NO.142 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331
-Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Pages vii 3.12-3a viii 3.12-4 1.2-4 3.12-5 3.3-2 3.12-5a 3.3-4 3.12-6 3.3-5 3.12-7 3.3-11 3.12-8 3.3-11a*
3.12-9 3.3-14 3.12-10 3.3-15 3.12-12 3.3-16 3.12-13 3.3-17 3.12-14 3.3-20 3.12-17 3.5-26 3.12-20 3.12-1 3.12-22 3.12-2 3.12-23*
3.12-3 5.2-1
- These pages are being deleted l
i
DAEC-1 TECHNICAL SPECIFICATIONS LIST OF FIGURES Figure Number Title 1.1 -1 Power / Flow Map 1.1-2 Deleted 2.1-1 APRM Flow Biased Scram and Rod Blocks 2.1 -2 Deleted 4.1 -1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability Vs. Test Interval
- 3. 4-1 Sodium Pentaborate Solution Volume Concentration Requirements 3.4-2 Saturation Temperature of Soditsn Pentaborate Solution
- 3. 6-1 DAEC Operating Limits 4.8.C-1 DAEC Emergency Service Water Flow Requirement 3.12-1 Flow-Dependent Minimum Critical Power Ratio (MCPR )
p 3.12-2 Power-Dependent Minimum Critical Power Ratio Multiplier (K )
l p
3.12-3 Minimum Critical Pcwer Ratio (MCPR) versus t (Fuel Types:
BP/P8X8R GE8X8EB, LTA-311 and ELTA) l 3.12-4 Limiting Average Planar Linear Heat Generation Rate (Fuel Type:
BD303A) 3.12-5 Limiting Average Planar Linear Heat Generation Rate (Fuel Type:
LTA 311) 3.12-6 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P80RB301L) 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type:
BD299A) l 3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Types:
BP/P8DRB299 and ELTA) 3.12-9 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P80RB284H)
Amendment No. 179, 142 yjj
DAEC-1 Figure Number Title 3.12-10 Flow-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC )
p 3.12-11 Power-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC )
p 3.'12-12 Flow-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC ) for SLO p
6.2-1 DAEC Nuclear Plant Staffing viii Amendment No. J/$, 142 n-
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DAEC-1 design pressure (120% x 1150 = 1380 psig; 120% x 1325 = 1590 psig).
The analysis of the worst overpressure transient, a 3 second closure of all main steam isolation valves with a direct valve position scram failure (i.e., scram is assumed to occur on high neutron flux), shows that the peak vessel pressure experienced is much less than the code allowable overpressure limit of 1375 psig (Reference 1). Thus, the pressure safety limit is well above the peak pressure that can result from reasonably expected overpressure transients.
A safety limit is applied to the Residual Heat Removal System (RHRS) when it is operating in the shutdown cooling mode. At this time it is included in the reactor coolant system.
1.2 References i
1.
Supplemental Reload Licensing Submittal for Duane Arnold Atomic Energy Center, Unit 1.*
- Refer to analyses for the current operating cycle.
1.2-4 Amendment No. J/$, 142
w-DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT b.
The control rod directional b.
control valves for inoperable control rods shall be disarmed electrically and the control (DELETED) rods shall be in such position that Specification 3.3. A.1 is met.
c.
Control rods with inoperable c.
Once per week when the plant is accumulators or those whose in operation, check status of position cannot be positively pressure and level alarms for detennined shall be considered each CRD acctsnulator.
inoperable, d.
d.
Once per quarter verify that:
(1) the Scram Discharge Volume (SDV) vent and' drain valves
'close within 30 seconds (DELETED) after receipt of a close signal, and (2) after removal of the close signal, that the SOV vent and drain valves are open.
Once per month verify that the SDV vent and drain valve position indicating lights located in the control room indicate that the valves are open.
e.
Control rods with scram times e.
Once per cycle verify that:
greater than those permitted by Specification 3.3.C.3 are (1) the SDV vent and drain inoperable, but if they can be valves close within 30 inserted with control rod seconds after receipt of a drive pressure they need not signal for the control rods be disarmed electrically.
to scram, and (2) open when the scram signal is reset.
)
f.
Inoperable control rods shall be positioned such that Specification 3.3. A.1 is met.
3.3-2 Amendment No. 77$,142
-DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4
c.
During each REFUELING OUTAGE l
observe that any drive which has been uncoupled from and subsequently recoupled to its
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control rod does not go to the overtravel position.
2.
The control rod drive housing 2.
The control rod drive housing support system shall be in support system shall be 4
place during REACTOR POWER inspected after reassembly and OPERATION or when the reactor the results of the inspection coolant system is pressurized
- recorded, above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.3. A.1 is met.
3.a Whenever the reactor is in 3.a. Prior to the start of control the STARTUP or RUN mode below rod withdrawal towards and prior to criticality % RATED POWER 30% RATED POWER, and the attaining 30 control rod movement is within the group notch mode after during rod insertion at 50% of the control rods have shutdown, the capability of the the Rod Rod Sequence Control System to been withdrawn,l System (RSCS) properly fulfill its function Sequence Contro shall be OPERABLE.
If the shall be verified by the system is determined to be following check:
i inoperable in accordance with checks in Specification Group Notch - Test the six be comparator circuits. Go 4.3.B.3, power may% RATED through each comparator.
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increased above 30 POWER by increasing core flow, inhibit, initiate test, verify error, and reset. After j
b.
Whenever the reactor is in comparator checks initiate test the STARTUP or RUN modes and observe canpletion of cple below 30% RATED POWER the indicated by illumination o.
Rod Worth Minimizer (RWM) test complete light.
l shall be OPERABLE or a second Reactor Operator shall verify b.
Prior to the start of control i
that the Reactor Operator at rod withdrawal towards criti-
)
i the reactor console is cality and prior to attaining following the control rod 30% RATED POWER during rod
- program, insertion at shutdown, the capability (of the Rod Worth Minimizer RWM) shall be I
i c.
If either Specifications 3.3.B.3.a or.b cannot be met, verified by the following the reactor shall not be checks:
I started or if the reactor is inthekUNorSTARTUPmodesat 1)
The correctness of the Reduced
)
less than 30% RATED POWER Notch Worth Procedure sequence controlrodmovementshallnot input to the RWM computer shall 1
be pennitted, except by a be verified.
scram. Limited control rod movement is permitted for the purpose of determining RSCS or RWM OPERABILITY and shall be i
verified by a second Reactor Operator.
Amendment No. U, 142 3.3 4 1
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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2)
The RWM computer on line diagnostic test shall be successfully performed.
3)
Proper annunciation of the selection error of at least one out-of-sequence control rod in each f0lly inserted group shall be verified.
4)
The rod block function of the RWM shall be verified by withdrawing the first rod as an out-of-sequence control rod no more than to the block point.
I 4.
Control rods shall not be 4.
Prior to control rod withdrawal withdrawn in STARTUP or in STARTUP or REFUEL modes, REFUEL modes unless at least verify that at least two Source two Source Range Monitor Range Monitor Channels have an Channels have an observed observed count rate of at least count rate equal to or greater three rounts per second.
than three counts per second.
5.
During operation with Limiting 5.
When a L'imiting ' Control Rod Control Rod Patterns, either:
Pattern exists, an Instrument Functional Test of the RBM a.
Both RBM channels shall be shall be performed prior to l
OPERABLE, or withdrawal of the designated rod (s).
b.
With one RBM channel inoper-able, control rod withdrawal shall be blocked within 24 l
hours, unless OPERABILITY is restored within this time period, or c.
With both RBM channels inoper-able, control rod withdrawal shall be blocked until l
OPERABILITY of at least one channel is restored.
Amendment No. //S, 142 3.3-5
DREC-1 maximum contribution to shut-down reactivity.
If it is disarmed electrically in a non-fully inserted position, that position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.l.
This assures that the core can be shut down at all times with the remaining control rods assuming the strongest operable control rod does not insert.
Inoperable bypassed rods will be limited within any group to not more than one control rod of a (5 x 5) twenty-five control rod array.
If damage within the control red drive mechanism and, in particular, cracks in drive internal housings cannot be ruled out, then a generic problem affecting a nunber of drives cannot be ruled out.
Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWR's.
This type of cracking could occur in a nunber of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.
Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large nunber of rods with failed collet housings.
3.3-11 Amendment No. 79, 142
A DAEC-1 At power levels below 20% of rated, abnormal control rod patterns could produce rod worths high enough to be of concern relative to the 280 calories per gram -
rod drop limit.
In this range the RWM and the RSCS constrain the control rod
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patterns to those which involve only acceptable rod worths.
The Reduced Notch Worth Procedure for control rod withdrawt.1 allows the Group Notch RSCS plants to take advantage of the Banked Position Withdrawal Sequence (BPWS) (Ref.1).
The BPWS has the advantage of having been proven statistically to have such low individual control rod worths that the possibility of a control rod drop accident (CRDA), which exceeds the 280 cal /gm peak fuel enthalpy limit, is precluded (Ref. 2).
The Rod Worth Minimizer and the Rod Sequence Control System provide automatic l
supervision to assure that out-of-sequence control rods will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences.
They serve as a backup to procedural control of control rod sequences, which limit the maximum reactivity worth of control rods.
In the event that the Rod l
Worth Minimizer is out of service, when required, a second Reactor Operator or other qualified technical plant employee whose qualifications have been reviewed by the NRC can manually fulfill the control rod pattern conformance functions of this system.
In this case, the RSCS is backed up by independent procedural controls to assure conformance.
The functions of the RWM and RSCS make it unnecessary to specify a license l
limit on rod worth to preclude unacceptable consequences in the event of a CRDA.
At low powers, below 20%, these devices force adherence to acceptable rod patterns.
Above 20% of rated power, no constraint on rod pattern is required to l
assure that the consequences of a CRDA are acceptable.
I 3.3-14 Amendment No. 19, 142 1
DAEC-1 l
Power level for automatic cutout of the RSCS function is sensed by first stage turbine pressure.
Because the instrument has an instrisnent error of + 10% full power, the nominal instrisnent setting is 30% of rated power.
Power level for automatic cutout of the RWM function is sensed by feedwater and steam flow and is set nominally at 30% of rated power to be consistent with the RSCS setting.
The Reduced Notch Worth Procedure is progransned into the RWM and is compatible with the hardwired Group Notch RSCS, In the pre-checkerboard pattern (100% to 50% control rod density), the RWM will enforce the Reduced Notch Worth Procedure; while in the post-checkerboard pattern (50% control rod density to RSCS/RWM low power setpoint) the RSCS will enforce the rod pattern. Therefore, the RSCS is not required to be OPERABLE until the post-checker-board pattern is entered.
Functional testing of the RWM prior to the start of control rod withdrawal at startup, and prior to attaining 30% rated thermal l
power during rod insertion while shutting down, will ensure reliable operation and minimize the probability of-the rod drop accident.
The RSCS can be functionally tested prior to control rod with-drawal for reactor startup. The hardware functional test sequence is performed to demonstrate that the Group Notch mode of the 3.3-15 Amendment No. 79, 142
DAEC-1 T
l RSCS is OPERABLE prior to entering the Group Notch mode (i.e., after 50%
l control rod density). The Group Notch restraints are automatically removeo above 30% power.
During reactor shutdown, similar surveillance checks shall be made with regard to rod group availability as soon as automatic initiation of the RSCS occurs and subsequently at appropriate stages of the control rod insertion.
If the operability requirements of either the RSCS or RWM are not satisfied, i.e., RSCS is inoperable or RWM is inoperable without the second reactor operator, then further rod movement is not permitted,
/
except by a scram (manual scram or mode switch to SHUTDOWN). This is done to ensure that high rod worths, with the potential to exceed 280 cal /gm during a CRDA are not generated. However, limited rod movement shall be pennitted solely for the purpose of troubleshooting and/or testing the RSCS or RWM for OPERABILITY. Limited rod movement is defined as the movement of control rod (s) only to the extent necessary to determine that the roa timibit functions of RSCS or RWM are working properly.
In addition, if the RSCS become inoperable and reactor power is less than l
30% of rateo, but feedwater flow is above the interlock at 20% of rated feedwater flow, reactor power may be increased above the RSCS low power setpoint (30% rated power) by increasing the core flow.
Increasing the power, without moving control rods, will ensure that a e/]tential CRDA will not exceed the 280 cal /gm limit mentioned earlier, absent the automatic rod pattern constraints of the RSCS.
d.
The Source Range ' Monitor (SRM) system performs no automatic safety system f
function; i.e., it has no scram function.
It does provide the operator j
with a visual indication of neutron level. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures that any transient, 3.3-16 Amendment No. U S, 142
.i
BAEC-1 should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimisn of two operable SRM's are provided as an added conservatism.
e.
The RBM provides local protection of the core; i.e., the prevention of
' boiling transition in a local region of the core, for a single rod withdrawal error from a Limiting Control Rod Pattern.
The trip point is referenced to power.
This power signal is provided by the APRMs. A statistical analysis of many single control rod withdrawal errors has been perfonnei and at the 95/95 level the results show that with the specified trip settings, rod withdrawal is blocked at MCPRs greater than the Safety Limit, thus allowing adequate margin.
This analysis asstanes a steady state MCPR of 1.20 prior to the postulated rod withdrawal error. The RBM functions are required when core thermal power is greater than 30% and a Limiting Control Rod Pattern exists.
When both RBM channels are operating either channel will assure required withdrawal blocks occur even asstsning a single failure of one channel. When a Limiting Control Rod Pattern exists, with one RBM channel inoperable for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, testing of the RBM prior to withdrawal of control rods assures that improper control rod withdrawal will be blocked (Reference 3).
Requiring at least half of l
the nonnal LPRM inputs to be operable assures that the RBM response will be adequate to protect against rod withdrawal errors, as shown by a statistical failure analysis.
The RBM bypass time delay is set low enough to assure minimtsn rod movement while upscale trips are bypassed.
A Limiting Control Rod Pattern for rod withdrawal error (RWE) exists when (a) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% < P < 90%) and the MCPR is less than 1.70, or (b) core thermal power is greater than or equal to 90% of rated (P > 90%) and tne MCPR is less than 1.40.
3.3-17 Amendment No. Up,142
DAEC-1 3.3 and
4.3 REFERENCES
1)
General Electric Service Information Letter (SIL) No. 316, Reduced Notch Worth Procedure, November 1979.
2)
General Electric Standard Application for Reactor Fuel, Mtut-240ll-P-A*.
3)
Average Power Range Monitor, Rod Block Monitor and Technical Specification Imarovement (ARTS) Program for the Duane Arnold Energy Center, NEDC-30813-P, December, 1984.
l
- Latest NRC-approved revision.
3.3-20 Amendment No. 129, 142
DAEC-1
3.5 REFERENCES
1.
Jacobs, I.M., Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards, General Electric Company, APto, April 1%8 ( APED 5736).
2.
General Electric Company, The GESTR-LOCA and SAFER Models for the Evaluation of Loss-of-Coolant Accident, NtDG-23/85-P, October 1984.
3.
General Electric, Duane Arnold Energy Center SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis, NEDC-31310-P, August 1986.
4.
General Electric Company, Analysis of Reduced RHR Service Water Flow at the Duane Arnold Energy Center, NEDL-30051-P, January 1983.
5.
General Electric Company, Duane Arnold Energy Center Suppression Pool Temperature Response, NEDC-ZZO82-P, March 1982.
1 3.5-26 Amendment No. 77E, 142
DAEC-l' LIMITING CONo!TIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12 CORE THERMAL LIr41TS 4.12-CORE THERMAL LIMITS i
Applicability Applicability The Limiting Conditions for The Surveillance Requirements Operation associated with the apply to the parameters which fuel rods apply to those monitor the fuel rod operating parameters which monitor the conditions.
fuel rod operating conditions.
Objective Objective The Objective of the Limiting The @jective of the Surveillance Conditions for 0peration is to Requirements is to specify the assure the performance of the type and frequency of fuel rods.
surveillance to be applied to i
the fuel rods.
Specifications Specifications A.
Maximum Average Planar Linear A.
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
Heat Generation Rate (MAPLHGR) 1.
During REACTOR POWER The MAPLHGR for each type of
}
OPERATION, the actual MAPLHGR fuel as a function of average j
for each type of fuel as a planar exposure shall be i
function of average planar determined daily during reactor i
exposure shall not exceed the operation at > 25% RATED POWER.
l limiting value shown in Figs.
1 l
3.12-4, -5,
-6, -7, -8 and -9 multiplied by the smaller of j
the two MAPFAC factors determined from Figs. 3.12-10 and 3.12-11.
l 2.
During SLO, the actual MAPLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value shown in Figs.
l 3.12-4, -5, -6, -7, -8 and -9 multiplied by the smaller of the two MAPFAC factors determined from Figs. 3.12-11 l
and 3.12-12.
1 i
3.12-1 Amendment No. Up,142 4
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DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.
If at any time during REACTOR POWER OPERATION (one or two loop) at > 25% RATED POWER, it is determTned by normal surveillance that the limiting value for.MAPLHGR (LAPLHGR) is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits.
If the MAPLHGR (LAPLHGR) is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor l
power to < 25% of RATED POWER, or to sucW a power level that the limits are again being met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.
If the reactor is being operated in SLO and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought to the COLD SHUTDOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
5.
For either the one or two loop operating condition surveillance and corresponding action shall continue until the prescribed action is met.
8.
Linear Heat Generation Rate 8.
Linear Heat Generation Rate (LHGR)
(LHGR )
1.
During REACTOR POWER OPERATION The LHGR as a function of core the LHGR of any rod in any height shall be checked daily BP/P8X8R or ELTA fuel assembly during reactor operation at > 25%
shall not exceed 13.4 KW/ft, RATED POWER.
l while the LHGR of any rod in l
any GE8X8EB or LTA 311 fuel assembly shall not exceed 14.4 KW/ft.
1 l
3.12-2 Amendment No. 129, 142
.I DAEC-1 LIMITING COWITIONS FOR OPERATION SURVEILLANCE REQUIREPENT 2.
If at any time during REACTOR POWER OPERATION at > 25% RATED POWER it is determiiied by
-l nonnal surveillance that the limiting value for LHGR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to l
< 25% of RATED POWER, or to such a power level that the limits are again being met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Surveillance and corresponding action shall continue until the prescribed limits are again being : net.
C.
Minimum Critical Power Ratio C.
Minimum Critical Power Ratio i
(MCPR)
(MCPR) 1.
During REACTOR POWER OPERA-MCPR shall be determined daily TION, the MCPR shall be equal during REACTOR POWER OPERATION i
to or greater than the Operat-at > 25% RATED POWER and ing Limit MCPR, which is a folTowing any change in power function of core thermal level or distribution that would power, core flow, fuel type cause operation with a Limiting and scram time (t). For core Control Rod Pattern as defined thermal power greater than or in Section 3.2.C.2(a). During equal to 25% of rated and less operation with a Limiting Control than 30% of rated (25% <-P <
Rod Pattern, the MCPR shall be 30%), the Operating'LimTt MCPR determined at least once per 12 l
is given by Fig. 3.12-2.
For hours.
r core thermal power greater 1
than or equal to 30% of rated i
(P > 30%), the Operating Limit i
MCPE is the greater of 1
either:
a) The applicable flow-dependent MCPR (MCPR )
p determined from Figure 3.12-1, or b) The appropriate RATED POWER MCPR from Figure 3.12-3 [MCPR(100)]
multiplied by the applicable power-dependent MCPR multiplier 3.12-3 Amendment No. Up,142
DAEC-1
^
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT (K ) determined from n
l Figure 3.12-2.
2.
During SLO with core thermal power greater than or equal to 25% of rated, the Operating Limit MCPR is increased by adding 0.03 to the above determined Operating Limit MCPR.
3.
If at any time during REACTOR POWER OPERATION (one or two recirc. loop) at > 25% RATED POWER, it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits.
If the operating MCPR is not returned to within the prescribed limits within two hours, reduce reactor power to l
< 25% of RATED POWER, or to such a power level that the limits are again being met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.
If the reactor is being operated in SLO, and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought to a COLD SHUTDOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
5.
For either the one or two recirc. loop operating i
condition surveillance and corresponding action shall l
continue until the prescribed action is met.
\\
3.12-3a Amendment No. 170, 142
-,: n
DAEC-1 3.12 BASES: CORE THERMAL LIMITS A.
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10CFR50.46 and that the fuel desigr. analysis limits specified in NEDE-24011-P-A (Reference 1) will not be exceeded.
Mechanical Design Analysis:
NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in.a lattice operating at 'the bounding power history, meet the fuel design limits specified in Reference 1.
No single fuel rod follows, or is capable of following, this bounding power history.
This bounding power history is used as the basis for the fuel design analysis MAPLHGR limit.
LOCA Analysis: A LOCA analysis is performed in accordance with i
10CFR50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits comoly with the ECCS limits specified in 10CFR50.46.
The analysis is performed for the most limiting break size, break location, and single failure combination for the plant (Reference 2).
The Technical Specification MAPLHGR limit is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the LOCA analysis MAPLHGR limit.
The actual MAPLHGR values for the GE 8 fuel design are lattice-type dependent and are explicitly modeled by the plant process computer.
The lattice-type dependent values can be found in Reference 2.
The Technical Specification MAPLHGR limit is a nominal representation of the lattice-dependent values, (i.e.,
the most limiting lattice-type, other than the natural uranium bundle ends), which can be used to conservatively model the MAPLHGR limit if the process computer becomes unavailable.
3.12-4 Amendment No. 129, 142
DAEC-1 The flow-dependent correction factor (Figure 3.12-10) applied to the MAPLHGR limits at rated conditions assures that (1) the 10CFR50.46 limit would not be exceeded during a LOCA initiated from less than rated core flow conditions and (2) the fuel thermal-mechanical design criteria would be met during abnormal operating transients initiated from less than rated core flow conditions (Reference 5).
l The power-dependent correction factor (Figure 3.12-11) applied to the MAPLHGR limits at rated conditions assures that the fuel thermal-mechanical design criteria would be met during abnormal operating transients initiated from less than rated power conditions (Reference 5)..
l For two recirculation loop operation, the calculational procedures used to establish the MAPLHGR's shown on Figures 3.12-4 thru 3.12-9 are documented in Reference 1.
The reduction factors for SLO were derived in Reference 4.
B.
Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation. rate and that the fuel cladding 1% plastic diametral strain linear heat generation rate is not exceeded during any abnormal operating transient if fuel pellet densification is postulated.
The LHGR l
as a function of core height shall be checked daily during reactor operation at > 25% power to determine if fuel burnup, or control 3.12-5 Amendment No. US,142 4
4 OAEC-1 4
rod movement has caused changes in power distribution.
For LHGR to be a limiting value below 25% rated thermal power, the Maximum Total Peaking Factor (MTPF) wwld have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.
C.
Minimum Critical Power Ratio (MCPR) 1.
Operating Limit MCPR The required operating limit MCPR's at stecdy state operating conditions as specified in Specification 3.12.C are derived from the established fuel cladding integrity Safety Limit MCPR value, and an analysis of abnormal operational tran-l sients (Reference 1). For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip settings given in Specification 2.1.
1 i
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
3.12-5a Amendment No. J20, 142
\\
\\
DAEC-1 The limiting transient, which determines the required steady state MCPR limit, is the transient which yields the largest ACPR. The minimtsn Operating Limit MCPR of Specification 3.12.C bounds the sun of the Safety Limit MCPR and the largest ACPR.
l The required MCPRs at rated power [MCPR(100)] are determined using the GEMINI transient analysis methods described in Reference 1.
These limits were derived by using the GE 678 scram times, given in Section 3.3.C. which are based upon extensive operating plant data, as well as GE test data. The ODYN Option 8 scram insertion times were statistically derived from the 678 data to ensure that the resulting Operating Limit from the transient analysis would, with 95% probability at the 95% confidence level, result in the Safety Limit MCPR not being exceeded. The scram time parameter (T), as calculated by the following fonnula, is a measure of the conformance of the actual plant control rod drive performance to that used in the ODYN Option-B licensing basis:
T
- Tb ave T a ta
- Tb where:
T,y, = average scram insertion time to Notch 38, as measured by surveillance testing Tb
= scram insertion time to Notch 38 used in the ODYN Option-B Licensing Basis.
T
= 678 scram insertion time to Notch 38 a
As the average scram time measured by surveillance testing (Taye), exceeds the ODYN Option B scram time (tb), the MCPRs l
at rated power [MCPR(100)] must be adjusted using Figure 3.12-3.
3.12-6 Amendment No. J20, 142
OAEC-1 2.
MCPR Limits for Other Than Rated Power and/or dated Flow Conditions At less than 100% of rated power and/or flow tne required Operating Limit MCPR is the larger value of the flow-dependent MCPR (MCPR ) or the power-dependent multiplier (K ) times p
p the rated power MCPR [MCPR(100)] at the existing core power /
flow state. The required Operating Limit MCPR is a function of flow in order to protect the fuel from inadvertent core flow increases such that the Safety Limit MCPR requirement can be assured.
The MCPRps were calculated such that, for the maximtsn core flow rate and core thermal power along a conservative load line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit MCPR.
Using this relative bundle power, the MCPRs were calculated at different points along this conservative load line corresponding to different core flows. The resulting MCPRps are given in i
Figure 3.12-1.
For operation above 30% of rated thennal power, the core power-dependent MCPR operating limit is the rated power MCPR
[MCPR(100)], multiplied by the factor given in Figure 3.12-2, l
1.e., K. For operation below 30% of rated thermal power,-
p where the direct scrams on turbine control valve fast closure and turbine stop valve closures are bypassed, absolute MCPR 4
limits are established.
This limit is taken directly from 3.12-7 Amendment No, J20, 142 e
DAEC-1 l
Figure 3.12-2.
This limit protects the fuel from abnormal operating transients, including localized events, such as a rod withdrawal error, other than those resulting from inadvertent core flow increases, which are covered by the flow-dependent MCPR limits.
This power-dependent MCPR limit was developed based upon bounding analyses for the most limiting transient at the given core power level.
Further information on the MCPR operating limits for l
off-rated conditions is presented in Reference 5.
i At thermal power levels less than or equal to 25% of rated thermal power, operating plant experience indicates that the resulting MCPR value is in excess.of the requirements by considerable margin. Therefore, monitoring of MCPRs below this power level is unnecessary.
The daily monitoring of MCPRs above 25% of rated thermal power is sufficient, since power distribution shifts are very slow, provided that no significant changes in core flow or control rod pattern have taken place.
2 During SLO, the Operating Limit MCPR must be increased by 0.03 to account for the increased uncertainty in the core flow and Transversing In-core Probe (TIP) readings used in the statistical analyses to derive the Safety Limit MCPR l
(see Reference 4).
3.12-8 Amendment No. J20, 142
DAEC-1 1
4.12 BASES: CORE THERMAL LIMITS C.
Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative state relative to MCPR.
The daily requirenent for calculating MCPR l
above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or co'ntrol rod changes. The requirement for calculating MCPR when operating with a Limiting Control Rod Pattern assures that Safety Limit MCPR will not be violated given a single rod withdrawal error (Reference 5).
)
1 3.12-9 Amendment No. J20, 142 l
DAEC-1 3.12 REFERENCES 1.
General Electric Standard Application for Reactor Fuel, MtUL-24011-P-A*.
2.
Duane Arnold Ener;y Center SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis, NEDC-31210-P, August 1986.
3.
Supplemental Reload Licensing Submittal for Duane Arnold Atomic Energy Center, Unit 1.**
4.
Duane Arnold Energy Center Single Loop Operation, NE00-24272, July 1980.
5.
Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for the Duane Arnold Energy Center, NEDC-30813-P, December 1984.
- Approved revision number at time reload fuel analyses are performed.
- Analysis is cycle-dependent; see the report for the current operating cycle / reload.
l 3.12-10 Amendment No. J20, 142
_.______..-.___...._.m.
t; i
?
]
2.4 l
RATED MCPR MULTIPLIER lKpl c
i OLMCPR t
t 2.3 a
> S05 CORE FLOW OPER ATiglT MCPR iP) = Kp 'OPER ATING LIMIT MCPRil008 2.2 l
FOR P < 25%: NO THERMAL LIMITS MONITORING REOutRED e.y, NO LIMIT 5 SPECIFIED j
1 E
2.1 l
l FOR 25% 5,P < PSYPASS: IPBYPASS = 30% FOR THE DAEC) m j
g l
l Kp = IKaVP + 0 02130% - Pil tOL MCPR 19001 K9YP = 1.90 FOR S 50% CORE FLOW
)
2.0
<50% CORE FLOW
= 2.15 FOR > 50% CORE FLOW g
b FOR 30% $ P < 45%:
I.
0
~
Kp = 1.29 + 0 01341455-PI e.
i I
I FOR 49% $ P < $0%:
- s
- ]
FOR 00% < P:
1.5 Ep = 1.9
- 0.00375 (100%-P) w 7
M
?
1 I
I 4
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I I
l g
i e
i.2 I
l a
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=
l g
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e i
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20 25 30 40 60 so 70 SO 90 100
)
y POVPASS POWER l% estedl j
DUM8E AflNOLD ENERGY CENTER DOWA ELECTRIC LIGHT AND POWER COPANY i
]
TECHN9 CAL SPECIFICATIONS 4
POIER-DEPEteENT MSNH4UM CRITICAL POWER RATDO NULTDPLIER (K )
F)GtRE 3.12-2 p
l 1
DAEC-1 Option B
Option A
1.30 -
- 1.30 g
g 1.25 -
- 1.25 5
a
~
~
1.21 -
~
g 1.20 -
- 1.20 E
g l.15 -
- 1.15 l.10 -
- 1.10 I
Ok2 I
O!4 I
OI8 I
0.0 O6 1.0 T
(based on tested measured scram time as defined in Reference 11)
J
.I DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO AT RATED POWER [MCPR(100)] VERSUS T FUEL TYPES:
BP/P8XER, GE8X8EB, LTA-311 and ELTA FIGURE 3.12-3 3.12-13 Amendment No. J28, 120, 142 4
3
DAEC-1 MAPLHGR vs FUEL EXPOSURE BD303A 14.0 12.2 12.2 12M N,1 I
.9 23 o uy.
1:
- 3 cL 12.0 i
I1*
cn g
$c 11.0 cn55 e
10.0
$i s
a b
9.0 1
3 8.0 O.0 0.2 2.0 4.0 6.0 8.0 10.0 12.5 15.0 20.0 42.0 50.0 Pianor Average Exposure (GwD/ST)*
- 1 GWd/t = 1000 mwd /t These values are nominal values to be used for manual calculations. The actual lattice-type dependent values are modeled in the process computer.
DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE FUEL TYPE: BD303A FIGURE 3.12-4 3.12-14 Amendment NO. J20, 142
DAEC-1
~
4 l
MAPLHGR vs FUEL EXPOSURE BD299A 14.0
'5:
'5:
0 0
II"'
y""
13.0 r
11 g
1:.3 g
E6 12.0 h
11 De 11.1 11 dj 11.0
Ei 3o j
10.0 "I
(
- 3 9.0 8.0 0.0 0.2 2.0 4.0 6.0 8.0 10.0 12.5 15.0 20.0 42.0 50.0 Planar Average Exposure (GWD/ST)*
- 1 GWd/t = 1000 mwd /t
- These values are nominal values to be used for manual calculations. The actual lattice-type dependent values are modeled in the process computer.
DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE FUEL TYPE: BD299A l
FIGURE 3.12-7 3.12-17 Amendment NO. J20. 142
_ __=
4 j
t i
i i
i l
i 10 1
i
/.
/*/
- ,s
- [
i l
0 90
/s
)
I
'l aI O to 4
MAPLHGRy = MAPF ACp
- MAPLHGRSTO i
d E
MAPLHGRgro = STANDARD MAPLHOR LIMITS i
a j
O MAPF ACp = MINIMUM IMFRPOp, MAPRAULTp*]
.V 4
w OM MFRPOplFI = MINIMUM (1.0. Ap + 5pFI g
j
'g i
N I
FFt k
0, F = FM ACTION OF M ATED CORE FLOW, p
ANO Ap.Sp ARE FUEL TYPE DEPENDENT j
o i
o CONSTANTS GIVEN BELOW:
5 3
FM MM, a
FOR 7s7, GE8X E,B,-
j g
050 MAXIMUM
.......R tw3 4
CORE FLOW 4
(% v4esiel Ap Sp
'Ap~ -~~ ~ Sp 102 5 04698 065S7 0.4869 0 8704 p
m 1070 0 4421 0 6533 0.4574 0 SPSS i
e 112 0 0 4074 06581 04214 06907 m
]
g 0 50 Il7 0 0 3709 05656 03828 06806 3
.o e+
1 2
- MArneULTy FROM ECCS CONS 80ERATIONS IS DOUNDEO P
SYhtFRPDp i
i i
I s
I l
w J
N 30 40 60 60 70 80 90 100 t10 t
]
5 CORE F LOW t% eseedl DUANE ARNOLO ENERGY CENTER N
i 10WA ELECTRIC LIGHT AND POWER C0pFANY TECHNICAL $PECIFICATIONS 1
FLOW-DEPENDENT NAXINIM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (NFLHGR)
MJLTIPLIER (l#PFAC )
s C
_HgPrTUkgq.gq__F
1.
4 10 1
I 1
1 0 90 (IP/P8X8R, GE8X8EB, LTA-3tl & ELTA) 0.8 )
u 0 a0 g.d '
- 1 i
p+
p.
MAPLHGRp = MFFACp
- M@LHGt
=
STD 0
3 MAPLHGR
= STANDAfW MAPLHIR LIMITS STD e
MFFACp = MINIMUM IMFFDF, MMMULT t 4
+
p w
u 2
0 70 MFFD (F) = MINIMtm (1.0, Ap + BrF) j F
m e
k 0,
F = FRACTION OF RATED CORE FLOW, N
Ale Ay, BF ARE FUEL TYPE DEPEIOENT E
CONSTANTS GIVEN BELOW 3
FOR P84, g
r0R 7x7,
- GE8x8EB, 2
MAX 198M 8x8, 8x8R LTA-311 CORE FLOW (5 rated)
Ap By Ap Sp g
102.5 0.4696 0.6557 0.4861 0.6784 d) 107.0 0.4421 0.6533 0.4574 0.6758 5
112.0 0.4074 0.6581 0.4214 0.6807 0 50 117.0 0.3701 0.6656 0.3828 0.6886 g
acv 5
U l
I I
I I
I I
0 40 7
30 40 50 60 70 80 90 100 110 CORE FLOW (% eeedl DUANE ARNOLD ENERGY CENTER 10WA ELECTRIC LIGHT AND POWER COWANY TECHNICAL SPECIFICATIONS FLOW-OEPENDENT MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
ORTf1Rt0F0 GTOcc ) N Sta e
L.
o DAEC-1 t
5.2 REACTOR 1.
The core shall consist of not more than 368 fuel assemblies of an approved fuel design.
2.
The reactor core shall contain 89 cruciform shaped control rods of an approved design.
5.2-1 Amendment No. 17, 33, 142