ML20116F054
| ML20116F054 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 04/17/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co |
| Shared Package | |
| ML20116F056 | List: |
| References | |
| DPR-49-A-117 NUDOCS 8504300540 | |
| Download: ML20116F054 (20) | |
Text
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NUCLEAR RECULATORY COMMISSION t.
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IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE
_C_0RN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 117 License No. DPR-49 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated August 17, 1984, 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 Comission; 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of facility Operating License No. DPR-49 is hereby amended to read as follows:
8504300540 950417 PDR ADOCK 05000331 P
l l 1 (2) Technical Specifications 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.
3.
The license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: April 17,1985 l
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ATTACHMENT TO LICENSE AMENDMENT NO.117 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 i
l Revise the Appendix A Technical Specifications by removing the current pages and inserting the revised pages listed below. The revised areas are identified by vertical lines.
LIST OF AFFECTED PAGES l
vi 3.12-3 3.12-11 vii 3.12-Sa 3.12-13 3.3-6 3.12-6 3.12-16*
3.3-18 3.12-7 3.12-17 3.5-26 3.12-8 3.12-19 3.12-1 3.12-9a*
- These pages are being deleted.
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DAEC-1 TABLE NO.
TITLE PAGE N0.
4.2-D Minimum Test and Calibration Frequency for Radiation 3.2-29 Monitoring Systems 4.2-E Minimum Test Calibration Frequency for Drywell Leak 3.2-30 Detection 4.2-F Minimum Test Calibration Frequency for Surveillance 3.2-31 Instrumentation 4.2-G Minimum Test and Calibration Frequency for 3.2-34 Recirculation Pump Trip 4.2-H Accident Monitoring Instrumentation Surveillance 3.2-34a Requirements 3.7-1 Containment Penetrations Subject to Type "B" Test 3.7-20 Requirements 4
3.7-2 Containment Isolation Valves Subject to Type "C" Test 3.7-22 Requirements J
3.7-3 Primary Containment Power Operated Isolation Valves 3.7-25 4.7-1 Summary Table of New Activated Carbon Physical 3.7-50 Properties 4.10-1 Sunmary Table of New Activated Carbon Physical 3.10-7 Properties 3.12-1 Deleted 2
i 3.12-2 Deleted l
3.13-1 Fire Detection Instruments 3.13-11 3.13-2 Required Fire Hose Stations 3.13-12 6.2-1 Minimum Shift Crew Personnel and License Requirements 6.2-3 6.9-1 Deleted 6.11-1 Reporting Sunmary - Routine Reports 6.11-12 6.11-2 Deleted i
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j Amendment No.1,)4,117 vi 1
1
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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 Sodium Pentaborate Solution 3.6-1 DAEC Operating Limits 4.8.C-1 DAEC Emergency Service Water Flow Requirement 3.12-1 Kf as a Function of Core Flow 3.12-2 Minimum Critical Power Ratio (MCPR) versus t l
3.12-3 Deleted 3.12-4 Deleted 3.12-5 Deleted 3.12-6 Limiting Average Planar Linear Heat Generation Rate (FuelTypeBP/P8DR8301L) l 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P8DP8289) l 3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P80RB299) l 3.12-9 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P80RB284H) 6.2-1 DAEC Nuclear Plant Staffing VII Amendment No.
, 117
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.
Scram Insertion Times C.
Scram Insertion Times 1.
The average scram insertion 1.
After each refueling outage all time, based on the de-operable rods shall be scram energization of the scram time tested from the fully pilot valve at time zero, of withdrawn position with the all operable control rods in nuclear system pressure above the reactor power operation 950 psig (with saturation condition shall be no greater temperature) and the than:
requirements of Specification 3.3.B.3.a met. This testing
% Inserted Average Scram shall be completed prior to from Fully Rod Insertion exceeding 40% power.
Below 30%
Withdrawn Position Times (Sec) power, only rods in those sequences (A and A or 8 12 3u 12 5
44 0.375 and B3%) which are fully 20 38 0.900 withdrawn in the region from 50 24 2.000 100% rod density to 50% rod 90 04 3.500 density shall be scram time tested.
During all scram time 2.
The average scram insertion testing below 30% power, the times for the three fastest Rod Worth Minimizer shall be control rods of all groups of operable or a second licensed four control rods in a 2 x 2 operator shall verify that the array shall be no greater operator at the reactor console than:
is following the control rod program.
% Inserted Average Scram from Fully Rod Insertion Withdrawn Position Times (Sec) 5 44 0.398 20 38 0.954 50 24 2.120 90 04 3.710 3.
Maximum scram insertion time for 90% insertion of any operable control rod should not exceed 7.00 seconds.
3.3-6 Amendment No. 1
,117
DAEC-1 After initial fuel loading and subsequent refuelings when operating above 950 psig, all control rods shall be scram tested within the constraints imposed by the Technical Spec'fications and before the 40% power level is reached. The requirements f or the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
4.
Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as.any burnable poision in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data ' relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.
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Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceed,s 1%aK.
Deviations in core reactivity greater than 1% AK are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the react,or system.
Amendment No, 117 3.3-18
DAEC-1
3.5 REFERENCES
1.
Jacobs, I.M., " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards", General Electric Company, APED, April 1968 (APED 5736).
2.
General Electric Company, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NEDO-20566, 1974, and letter MFN-255-77 from Darrell G. Eisenhut, NRC, to E.D. Fuller, GE, Documentation of the Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-lead Plants, dated June 30, 1977.
3.
General Electric, Loss-of-Coolant Accident Analysis Report for Duane Arnold Energy Center (Lead Plant), NED0-21082-03, June 1984.
l I
4.
General Electric Company, Analysis of Reduced RHR Service Water Flow at the Duane Arnold Energy Center, NEDE-30051-P, January 1983.
5.
General Electric Company, Duane Arnold Energy Center Suppression Pool Temperature Response, NEDC-22082-P, March 1982.
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Amendment No.
,117
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Applicability Applicability The Limited 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 Objective of the Conditions for Operation is to Surveillance Requirements is to assure the performance of the specify the type and frequency fuel rods.
of surveillance to be applied to the fuel rods.
Specifications Specifications A.
Maximum Average Planar Linear A.
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
Heat Generation Rate (MAPLHGR)
During reactor power The MAPLHGR for each type of operation, the actual MAPLHGR fuel as a function of average for each type of fuel as a planar exposure shall be function of average planar determined daily during reactor exposure shall not exceed the operation at > 25% rated limiting value shown in Figs.
thermal power and following any l
3.12-6, -7, -8 and -9.
If change in power level or at any time during reactor distribution that would cause
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- power operation i H s operation with a limiting determined by normal control rod pattern as surveillance that the limiting described in the bases for value for MAPLHGR (LAPLHGR) is Specification 3.3.2.
During being exceeded, action shall operation with a limiting then be initiated within 15 control rod pattern, the minutes to restore operation MAPLHGR shall be determined at to within the prescribed least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hours, 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 power to < 25% of Rated Thermal If6wer 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 met.
Amendment No.
, 117 3.12-1
I DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT C.
Minimum Critical Power Ratio C.
Minimum Critical Power Ratio (MCPR )
(MCPR) i During reactor power MCPR shall be determined daily operations, MCPR shall be >
during reactor power operation l
values as indicated in Fig 7re at > 25% rated thermal power and 3.12-2 at rated power and folTowing any change in power flow.
If at any time during level or distribution that would reactor power operation it is cause operation with a limiting determined by normal control rod pattern as described surveillance that the in the bases for Specificatic.i limiting value for MCPR is 3.3.2.
During operation with a being exceeded, action shall limiting control rod pattern, then be initiated within 15 the MCPR shall be determined at minutes to restore operation least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I to within the prescribed limits.
If the operating MCPR is not returned to within the prescribed limits l
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce I
reactor power to < 25% of l
Rated Thermal Power within l
the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I Surveillance and corresponding action shall continue until the prescribed i
limits are again being met.
For core flows other than t
rated the MCPR shall be >
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values as indicated in FTgure l
3.12-2 times K, where Kr f
e is as shown in Figure 3.12-1.
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Amendment No. pd',117 3,12 3
. =.-
..e DAEC-1 derived from the established fuel cladding integrity l
Safety Limit MCPR value, and an analysis of abnormal l
operational transients (2). 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.
To assure that the fuel cladding integrity Safety Limit is i
not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in criticalpowerratto(CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
The limiting transient, which determines the required steady state MCPR limit, is the transient which yields the l
largest ACPR. The minimum operating limit MCPR of l
Specification 3.12.C bounds the sum of the safety limit MCPR and the largest ACPR.
The required minimum operating limit MCPRs are determined by the methods described in References 11 and 12. These Amendment No.[.117 3.12-Sa
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e DAEC-1 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 Ootion B 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 formula, is a measure of the conformance of the actual plant control rod drive performance to that used in the ODYN Option-8 licensing basis:
- ave
- b t
-T a
b where: tave = average scram insertion time to Notch 38, as measured by surveillance testing T b
= scram insertion time to Notch 38 used in the 00YN Option-8 Licensing Basis.
T
= 678 scram insertion time to Notch 38 a
As the average scram time measured by surveillance testing (tave), exceeds the 00YN Option 8 scram time (t ), the b
Operating Limit MCPRs must be adjusted using Figure 3.12-2.
3.12-6 Amendment No JE57 117 d
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DAEC-1 l
1 2.
MCPR Limits for Core Flows Other than Rated Flow The purpose of the K factor is to define operating limits at other than rated f
flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the K factor. Specifically, the K factor f
f provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.
1
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For operation in the automatic flow control mode, the K factors assure that f
the operating limit MCPR of values as indicated in Figure 3.12-2 will not be l
violated should the most limiting transient occur at less than rated flow.
In the manual flow control mode, the Kf factors assure that the Safety Limit MCPR will not be violated for the same postulated transient event.
The K factor curves shown in Figure 3.12-1 were developed generically and are f
applicable to all BWR/2, BWR/3 and BWR/4 reactors.
The K factors were f
derived using the flow control line corresponding to rated t'h'ermal power at j
rated core flow.
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For the manual flow control mode, the Kg factors were calculated such that at the maximum flow state (as limited by the pump scoop tube set ' point) and the corresponding core power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the value of K..
f Amendment No.
, 117 3.12-7
' o..,
DAEC-1 For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow.
The K factors shown in Figure 3.12-1 are conservative for Duane Arnold f
l operation because the operating limit MCPR of values as indicated in Figure 3.12-2 is greater than the original 1.20 operating limit MCPR used for the generic derivation of K.
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64 gT 6
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9 Amendment No.
117 3.12-8 s
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DAEC-1 l
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DELETED M
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Amendment flo.
,117 3.12-9a
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DAEC-1 3.12 REFERENCES 1.
Duane Arnold Energy. Center Loss-of-Coolant Accident Analysis Report, NEDO-21082-03, June 1984.
l 2.
General Electric Standard Application for Reactor Fuel, NEDE-240ll-P-A**.
l 3.
" Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-19735, August 1973.
4.
Supplement 1 to Technical Reports on Densifications of General Electric Reactor Fuels, December 14,1973 (AEC Regulatory Staff).
5.
Communication:
V.A. Moore to I.S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27,1974.
6.
R.B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NED0-10802).
7.
General Electric Company Analytical Mo' del for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NEDE-20566, August 1974.
8.
Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, NED0-24087, 77 NED 359, Class 1, December 1977.
9.
Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NEDO-24087-2.
- 10. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of Feedwater Heating, NEDO-24987-5.
- 11. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5,1980.
}
- 12. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment l
Methods for Determination of Operating Limits," January 19, 1981.
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- Approved revision number at time reload fuel analyses are performed.
Amendment No.
,117 3.12-11
'.. 'e DAEC-1 Option B
Option A
1 1.30 --
- 1.30 1 28
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1.26 1.25 --
- 1.25 T = 0.75 1.20 --
- 1.20 1.15 --
- 1.15 1.10 --
- 1.10 0.0 I
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O;2 O.4 0.6 O.8 1.0 T
(based on tested measured scram time as defined in Reference 11)
DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO (MCPR)
VERSUS T FUEL TYPE: 8P/P8X8R FIGURE 3.12-2 Amendment No.
,117 3.12-13 e
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.. t DAEC-1 e
DELETED
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AmendmentNo.[,117 3.12-16
1 s. -E 1
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12.3 12.2
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gw 12-11.9 r 11.5 o"
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5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Plana: Average Exposure (CWd/t)*
1/ When cera flov is equal to or less than 70I of rated, the MAPI2GR shall not exceed 931 of the 11=1 ting values shown.
- 1 GWd/t r1000 mwd /t DUANE ARNOLD EiIRGT CENTIR ICWA 7'TCRIC LIGHT AND POWER CCMPAh7 TICENICAL SPICIIICATION3 LDiITING AVI?. AGE PLANAR LDtIAR REAT GENI?JLTION RATE AS A TUNC"" ION OF FIMAR AVIRAGE ECP05UP2 FUIL TYPI: BP/P8DRB301L TICURI 3.11-6 Amendment No.
,117 3.12-17 w- - -
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12.3 12.2 j-N 12.1 h5 3
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9 9.0 E
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5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Fl em-Average Exposure (G%'d / t)
- 1/ When core flow is equal to or less than 70% of rated, the MAPLEGE shall not exceed 95% of the 'd-' ting values shown.
- 1 GWd/t i_1000 mwd /t
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DUANI ARNOD ENIRGY CIh m IC"A ELECTR.IC LIGE:' AND PCL'ER CCX? ANT TECENICAL S?ECITICA~ ION 3 LIMITING AVERAGE PLA.LG. LLT.AR HEAT
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GINERATION RATI AS A TCNC"'!CN OF PLA.'LU.
AVIRAGE EI?OSURE FUIL TI?E: 3p/pscR3299 l
FIGURE 3.12-8 Amendment No.
117 3*12-19 I