ML112230639
| ML112230639 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/17/1984 |
| From: | IES Utilities, (Formerly Iowa Electric Light & Power Co) |
| To: | |
| Shared Package | |
| ML112230638 | List: |
| References | |
| NUDOCS 8408280206 | |
| Download: ML112230639 (19) | |
Text
PROPOSED CHANGE RTS 164 TO THE DUANE ARNOLD ENERGY CENTER TECHNICAL SPECIFICATION The holders of license DPR-49 for the Duane Arnold Energy Center propose to amend Appendix A (Technical Specifications) to said license by deleting current pages and replacing them with attached, new pages. A list of the affected pages is given below.
As new bundle types are added to the reactor core, the design basis Loss-of-Coolant Accident (LOCA) must be re-analyzed and the Technical Specifications updated to include the operating limits of these new bundle types. As part of the reload for Cycle 8, General Electric (GE) has reanalyzed the design basis event for the.two new bundle types (BP8DRB299 &
BP8DRB301L) being added to the core. The results of this analysis are reported in the attached updated LOCA report. This change to the Technical Specifications is to revise the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) operating limits based on this analysis for the new bundle types.
Also as part of the reload licensing process for Cycle 8, GE has reanalyzed the most limiting abnormal operating transients as specified in Table 15.0-1 in the Updated Final Safety Analysis Report (UFSAR).
The results of this analysis are reported in the attached supplement to the Reload Licensing Submittal.
This change to the Technical Specifications is to revise the Minimum Critical Power Ratio (MCPR) operating limits based on this analysis for all fuel types.
In addition, various administrative changes are being made, such as, revising figure numbers, updating the table of contents and correcting references.
The changes being made are as follows:
- 1) Change the List of Figures on page vii to delete MAPLHGR curves for the old 8x8 fuel, to add the new MAPLHGR curves for the new fuel and correct the title of Figure 3.12-7 to read P8DRB299.
- 2) Delete Table number 3.12-2 and update the List of Tables on page vi to reflect the fact that the MCPR limits will now be given in a figure format.
- 3)
Revise the scram insertion times in Section 3.3.C to be consistent with the ODYN Option-B scram times used in the GE transient analysis.
- 4) Delete the requirement to perform end-of-cycle (EOC) scram time testing, as this was required only through the end of Cycle 6, and update the Bases accordingly.
- 5) Update the reference sections of Sections 3.5 and 3.12 to reflect the new LOCA analysis.
8408280206 840817 PDR ADOCK 05000331 P
- 6)
Change the text of Section 3.12.A on page 3.12-1 to reflect deletion of the MAPLHGR curves for the 8x8 fuel types and the addition of the new BP8x8R fuel types.
- 7) Revise the text of Section 3.12.C to reflect that the MCPR limits are now in a figure format and to update the Bases accordingly.
- 8)
Revise the Bases of Section 3.12.C to reflect the procedure for determining the MCPR Operating Limits under ODYN Option-B.
- 9)
Delete Table 3.12-2, page 3.12-9a, on MCPR operating limits.
- 10)
Update the references for Section 3.12 to correct the title of NEDE-24011-P-A and to add the ODYN Option-B procedures referenced in the Bases.
- 11)
Add Figure 3.12-2, MCPR Limits versus average scram time (T).
- 12)
Delete the MAPLHGR curves on pages 3.12-16 and 17 and add new curves on pages 3.12-16 and 19.
List of Pages Affected vi 3.12-3 3.12-11 vii 3.12-5a 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.
TABLE NO.
4.2-D 4.2-E 4.2-F 4.2-G 4.2-H 3.6-1 4.6-3 4.6-4 4.6-5 3.7-1 3.7-2 3.7-3 4.7-1 4.10-1 3.12-1 3.12-2 3.13-1 3.13-2 6.2-1 6.9-1 6.11-1 6.11-2 RTS-164 W
DAEC-1 W
TITLE Minimum Test and Calibration Frequency for Radiation Monitoring Systems Minimum Test Calibration Frequency for Drywell Leak Detection Minimum Test Calibration Frequency for Surveillance Instrumentation Minimum Test and Calibration Frequency for Recirculation Pump Trip Accident Monitoring Instrument Surveillance Requirements Number of Specimens by Source Safety Related Snubbers Accessible During Normal Operation Safety Related Snubbers Inaccessible During Normal Operation Safety Related Snubbers in High Radiation Area During Shutdown and/or Especially Difficult to Remove Containment Penetrations Subject to Type "B" Test Requirements Containment Isolation Valves Subject to Type "C" Test Requirements Primary Containment Power Operated Isolation Valves Summary Table of New Activated Carbon Physical Properties Summary Table of New Activated Carbon Physical Properties Deleted Deleted Fire Detection Instruments Required Fire Hose Stations Minimum Shift Crew Personnel and License Requirements Protection Factors for Respirators Reporting Summary -
Routine Reports Reporting Summary -
Non-routine Reports vi PAGE NO.
3.2-29 3.2-30 3.2-31 3.2-34 3.2-34a 3.6-33 3.6-42 3.6-44 3.6-48 3.7-20 3.7-22 3.7-25 3.7-50 3.10-7 3.13-11 3.13-12 6.2-3 6.9-8 6.11-12 6.11-14 08/84
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 6.2-1 DAEC Nuclear Plant Staffing 3.12-1 Kf as a Function of Core Flow 3.12-2 Minimum Critical Power Ratio (MCPR) versus 3.12-3 Deleted 3.12-4 Deleted 3.12-5 Deleted 3.12-6 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P8DRB301L) 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P8DPB289) 3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P8DRB299) 3.12-9 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P8DRB284H) 08/84 RTS-164 v11i
DAEC-1 LIMITING CONDITION.FOR OPERATION C. Scram Insertion Times
- 1. The average scram insertion time, based on the de energization of the scram pilot valve at time zero, of all operable control rods in the reactor power operation condition shall be no greater than:
% Inserted from Fully Withdrawn 5
20 50 90 Rod Position 44 38 24 04 Average Scram Insertion Times (Sec) 0.375 0.900 2.000 3.500
- 2. The average scram insertion times for the three fastest control rods of all groups of four control rods in a 2 x 2 array shall be no greater than:
% Inserted from Fully Rod Withdrawn2 Position 44 38 24 04 Average Scram Insertion Times (Sec) 0.398 0.954 2.120 3.710
- 3. Maximum scram insertion time for 90% insertion of any operable control rod should not exceed 7.00 seconds.
SURVEILLANCE REQUIREMENT-C. Scram Insertion Times
- 1. After each refueling outage all operable rods shall be scram time tested from the fully withdrawn position with the nuclear system pressure above 950 psig (with saturation temperature) and the requirements of Specification 3.3.B.3.a met. This testing shall be completed prior to exceeding 40% power. Below 30%
power, only rods in those sequences (A12 and A34 or B12 and B34) which are fully withdrawn in the region from 100% rod density shall be scram time tested. During all scram time testing below 30% power, the Rod Worth Minimizer shall be operable or a second licensed operator shall verify that the operator at the reactor console is following the control rod program.
08/84 RTS-164 5
20 50 90 3.S-6
DAE C-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 Specifications and before the 40% power level is reached. The requirements for 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.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity.change exceeds 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 reactor system.
RTS-164 3.3-18 08/84
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), NEDO-21082-03, June 1984.
RTS-164 08/84 3.5-26
DAEC-1 LIMITING CONDITION FOR OPERATION 3.12 CORE THERMAL LIMITS Applicability The Limited Conditions for Operation associated with the fuel rods apply to those parameters which monitor the fuel rod operating conditions.
Objective The Objective of the Limiting Conditions for Operation is to assure the performance of the fuel rods.
Specifications A. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
During reactor power operation, the actual MAPLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value shown in Figs.
3.12-6, -7, -8 and -9. If at at any time during reactor power operation it is determined 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 power to < 25% of Rated Thermal Prwer 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.
SURVEILLANCE REQUIREMENT 4.12 CORE THERMAL LIMITS Applicability The Surveillance Requirements apply to the parameters which monitor the fuel rod operating conditions.
Objective The Objective of the Surveillance Requirements is to specify the type and frequency of surveillance to be applied to the fuel rods.
Specifications A. Maximum Average Planar Linear Reat Generation RateJM(MAPEHGR)
The MAPLHGR for, each type of fuel as a function of average planar exposure shall be determined daily during reactor operation at > 25% rated thermal power-and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.2. During operation with a limiting control rod pattern, the MAPLHGR shall be determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hours.
RTS-164 LIMITING CONDITION FOR OPERATION 08/84 3.12-1
LIMITING CONDITIONS FOR OPERATION C. Minimum Critical Power Ratio During reactor power operations, MCPR shall be >
values as indicated in Figure 3.12-2 at rated power and flow. If at any time during reactor power operation 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 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 Power 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.
For core flows other than rated the MCPR shall be >
values as indicated in FTgure 3.12-2 times Kf, where K is as shown in Figure 3.12-1.
DAEC 1 I SURVEILLANCE REQUIREMENT C.
Minimum Critical Power Ratio (MC PR)
MCPR shall be determined daily during reactor power operation at > 25% rated thermal power and folTowing any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.2.
During operation with a limiting control rod pattern, the MCPR shall be determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
RTS-164 08/84 3.12-3
SDAEC-1 derived from the established fuel cladding integrity Safety Limit MCPR value, and an analysis of abnormal 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 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.
The limiting transient, which determines the required steady state MCPR limit, is the transient which yields the largest ACPR. The minimum operating limit MCPR of 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 3.12-5a 08/84 RTS-164
DAEC-1 limits were derived by using the GE 67B 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 B scram insertion times were statistically derived from the 67B 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-B licensing basis:
ave -b T
Ta T b where:
Tave = 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.
Ta
= 67B scram insertion time to Notch 38 As the average scram time measured by surveillance testing, (Tave),exceeds the ODYN Option B scram time (Tb), the Operating Limit MCPRs must be adjusted using Figure 3.12-2.
RTS-164 08/84 3.12-6
DAEC-1
- 2. MCPR Limits for Core Flows Other than Rated Flow The purpose of the Kf factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the Kf factor. Specifically, the Kf factor 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.
For operation in the automatic flow control mode, the Kf factors assure that the operating limit MCPR of values as indicated in Figure 3.12-2 will not be 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 Kf factor curves shown in Figure 3.12-1 were developed generically and are applicable to all BWR/2, BWR/3 and BWR/4 reactors. The Kf factors were derived using the flow control line corresponding to rated thermal power at rated core flow.
For the manual flow control mode, the Kf 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 Kf.
08/84 RTS-164 3.12-7
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 Kf factors shown in Figure 3.12-1 are conservative for Duane Arnold 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.
RTS-164 08/84 3.12-8
DAEC-1 DELETED RTS-164 08/84 3.12-9a
DAEC-1 3.12 REFERENCES
- 1. Duane Arnold Energy Center Loss-of-Coolant Accident Analysis Report, NEDO-21082-03, June 1984.
- 2. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A**.
- 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 (NEDO-10802).
- 7.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR5O, Appendix K, NEDE-20566, August 1974.
- 8. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, NEDO-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), "ODYN Adjustment Methods for Determination of Operating Limits," January 19, 1981.
- Approved revision number at time reload fuel analyses are performed.
3.12-11 08/84 RTS-164
DAE C-1 Option B
1.30 1.26 1.25 1.20 1.15 1.10 -
0.0 T = 0.75 01.2 (based on tested measured 01.4 T
scram time 01.6 S01.8 Option A
1.30 1.28 1.25 1.20 1.15 1.10 1.0 as defined in Reference 11) 08/84 3.12-13 RTS-164 DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO (MCPR)
VERSUS t FUEL TYPE:
BP/P8X8R FIGURE 3.12-2
DAEC-1 DELETED RTS-164 08/84 3.12-16
12
-11 i
30.0 35.0 40.0 45.0 1/ When core flow not exceed 95%
Planar Average Exposure (GWd/t)*
is equal to or less than 70% of rated, of the limiting values shown.
the MAPLEGR shall
- 1 GWd/t = 1000 MWd/t 3.12-17 8 "
10*0 15.0 20.0 25.0 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: BP/P8DRB301L FIGURE 3.12-6 12.3 12.41
- 11. 9 11.5
\\
11.3 9 9.9 8.7 5.0
15.0 1 20.0 25.0 9
13 11 0
10
> a 10
.9 30.0 Planar Average Exposure (GWd/t) 1/ When core flow is equal to or less than 70% of rated, the MAPLEGR shall not exceed 95% of the limiting values shown.
- 1 GWd/t = 1000 MWd/t 3.12-19 1212 3
1
- 12. 1 11.5 11.5
-111.
0
.9 10.3 9.0
~
,~I 8
0 5.0 10.0 DUANE ARNOLD ENERGY CENTER IOWA EECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE FUEL TYPE: BP/P8DRB299 FIGURE 3.12-8 t
0 45.0 35.0U 140.0