ML17329A059
| ML17329A059 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/23/1990 |
| From: | Alexich M American Electric Power, American Electric Power Service Corp |
| To: | Murley T NRC/IRM |
| Shared Package | |
| ML17329A060 | List: |
| References | |
| AEP:NRC:1130, NUDOCS 9106250243 | |
| Download: ML17329A059 (29) | |
Text
ACCELERATED DISTRIB~+ON 'DEMONS (I
I 'p TION SYSTEM
~t REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9106250243 DOC.DATE: 90/07/23 NOTARIZED: NO DOCKET FACIL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Ind'a a
0500p315 AUTH.NAME AUTHOR AFFILIATION ALEXICH,M.P.
American Electric Power Co., Inc.
PJ gg/>J i RECIP.NAME RECIPIENT AFFILIATION HURLEY,T.E.
'ocument Control Branch (Document Control Desk) ~/
R
SUBJECT:
Application for amend to License DPR-58,decreasing min
/P'glFf I measured flow requirement as found in Table 3.2-1
& revising RCS limit Figure 2.1-1
& Table 2.2-1 functional unit 12 D
fo'otnote.Rept withheld.
DISTRIBUTION CODE:
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~ ~j TITLE: Proprietary Review Distribution Operating Reactor
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3 RECIPIENT ID CODE/NAME PD3-1 PD COPIES LTTR ENCL 1
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D INTERNAL: AEOD DOA EG EXTERNAL: NRC PDR 01 1
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NOTE TO ALL"RIDS" RECIPIENTS:
D D
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAMEFROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
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ENCL 7
4
American Electric Power Service Corporation 1 Riverside Plaza ~
Columbus. OH 43~
~
61 4 223 1000 AM ERIN ELECTRIC POWER AEP:NRC:1130 Donald C.
Cook Nuclear Plant Unit 1 Docket No.
50-315 License No.
DPR-58 TECHNICAL SPECIFICATIONS CHANGE FOR UNIT 1 CYCLE ll U.
S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
T.
E. Murley July 23, 1990
Dear Dr. Murley:
This letter constitutes an application for amendment to the Technical Specifications (T/Ss) for Donald C.
Cook Nuclear Plant Uni.t, 1.
Due to the possibility of increased steam generator tube plugging, additional margin is sought to the minimum measured flow requirement referenced in T/S 3.2.5 (Reactor Coolant System Total Flow Rate),
and associated changes to Safety Limits 2.1.1 (Reactor Core) and 2.2.1 (Reactor Trip System Instrumentation Setpoints).
Specifically, we are proposing to decrease the mini.mum measured flow requirement as found in Table 3.2-1 and to revise Reactor Core Safety Limit Figure 2.1-1 and the Table 2.2-1 Functional Unit 12 footnote to reflect this change.
Currently, Unit 1 has 2.8% margin to its minimum measured flow T/S. If significant additional plugging is required during the upcoming outage, approaching our currently analyzed limit of 10%
average plugging, approximately 1.28 in flow would be lost, lea~ing only 1.6% margin for measurement repeatabili.ty.
- Thus, although the actual flow is expected to be sufficient, measurement fluctuations could result in a failure to meet the T/S requirement.
Unit 1 will undergo extensive steam generator tube testing and repair which may result in additional tube plugging during the upcoming refueling outage.
The flow is measured soon after reaching Mode 1, which could be as early as December 19, 1990.
Therefore, we request that you respond to this proposal by December 12, 1990.
9 o 2 K 05 003' 50243 900723 PDR A
P (P>i i
Dr. T.
E. Hurley AEP:NRC:1130 A description of the change and our analyses concerning significant hazards are contained in Attachment 1.
The proposed revised T/S pages are contained in Attachment 2.
Attachment 3
contains an evaluation of the proposed changes from Westinghouse.
For the reviewer's convenience, Attachment 4 contains reports referenced in the Westinghouse evaluation in Attachment 3.
Attachment 4 includes:
4A.
One copy of WCAP-12568, "Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for American Electric Power, D.
C.
Cook Unit 1 Nuclear Power Station" (PROPRIETARY).
4B.
One copy of WCAP-12569, "Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for American Electric Power, D.
C.
Cook Unit 1 Nuclear Power Station" (NON-PROPRIETARY).
4C.
A Westinghouse authorization letter, CAW-90-046, Proprietary Information Notice, and accompanying Affidavit.
As Item 4A contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information.
The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.
Accordingly, it is requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.790 of the Commission's regulations.
ig Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse Affidavit should reference CAW-90-046 and should be addressed to R. A. Wiesemann, Manager of Regulatory
& Legislative Affairs, Westinghouse Electric Corporation, P.
O.
Box 355, Pittsburgh, Pennsylvania 15230-0355.
We believe that the proposed changes will,not result in (1) a significant change in the types of effluents or a significant increase in the amount of any effluents that may be released offsite, or (2) a significant increase in individual or cumulative occupational radiation exposure.
Dr. T.
E. Murley AEP:NRC:1130 These changes have been reviewed by the Plant Nuclear Safety Review Committee and the Nuclear Safety and Design Review Committee.
In compliance with the requirements of 10 CFR 50.91(b)(l), copies of this letter and its attachments have been transmitted to Mr. J.
R. Padgett of the Michigan Public Service Commission and the NFEM Section Chief of the Michigan Department of Public Health.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M.
P. Al ich Vice President MPA/eh Attachments cc:
D. H. Williams, Jr.
A. A. Blind - Bridgman G. Charnoff NFEM Section Chief J.
R. Padgett NRC Resident Inspector
- Bridgman A. B. Davis
- Region III
l yP I
l
ATTACHMENT 1 TO AEP:NRC:1130 REASONS AND 10 CFR 50.92 SIGNIFICANT HAZARDS EVALUATION FOR CHANGES TO THE DONALD C.
COOK NUCLEAR PIANT UNIT 1 TECHNICAL SPECIFICATIONS
ATTACHMENT 1 TO AEP:NRC:1130 Page 1
Introduction This letter requests modification to Technical Specification (T/S) 3.2.5 (Reactor Coolant System Total. Flow Rate),
and associated changes to Safety Limit 2.1.1 (Reactor Core) and 2.2.1 (Reactor Trip System Instrumentation Setpoints).
Specifically, we are proposing to decrease the minimum measured flow (MMF) requirement as found in Table 3.2-1 from 366,400 gpm to 361,600 gpm; to revise Reactor Core Safety Limit Figure 2.1-1 to reflect this change in MMF; and to revise the Table 2.2-1 Functional Unit Number 12 footnote to reflect the new MMF.
In addition, the mathematical symbols have been written out as an administrative change to Table 2.2-1.
The MMF requirement assures that reactor coolant system (RCS) flow meets the assumptions used in the NSSS design calculations and the accident and transient analyses.
For those depar'ture from nucleate boiling (DNB) transients that are analyzed using the Improved Thermal Design Procedure (ITDP) (Reference 1), the initial RCS flow is assumed to be the MMF.
For these
- analyses, uncertainties associated with the flow measurement are incorporated into the DNBR limit value.
For the NSSS design calculations, non-DNB related accident and transient
- analyses, and DNB transient analyses for which the ITDP is not used, the initial RCS flow assumed is the thermal design flow (TDF), which is 354,000 gpm for Cook Nuclear Plant Unit 1.
For these
- analyses, flow uncertainty is accounted for by the Technical Specification requirement (MMF) being larger than the TDF plus the calculated uncertainty.
The actual calculated RCS flow measurement uncertainty is 2.1%
(Reference 1)
~
The current T/S MMF value is based on the original design value of 3.5%.
This evaluation removes this margin from the uncertainty value.
Thus the new T/S will still assure the TDF is met.
Evaluations..are required for DNB calculations using the ITDP methodology.
Westinghouse has evaluated the Cook Nuclear Plant Unit 1 safety analyses and NSSS design calculations for the reduced RCS MMF (see ). It has been determined that the conclusions of the safety analyses remain valid for the reduction in RCS flow, and the NSSS design calculations are unaffected by the reduction in the MMF requirement.
10 CFR 50.92 Evaluation Per 10 CFR 50.92, a proposed amendment will not involve a significant hazards consideration if the proposed amendment does not:
ATTACHMENT 1 TO AEP:NRC:1130 Page 2
(1) involve a significant increase in the probability or consequences of an accident previously analyzed, (2) create the possibility of a new or different kind of accident from any accident previously analyzed or evaluated, or (3) involve a significant reduction in a margin of safety.
Our evaluation of the proposed change with respect to these criteria is provided below.
Criterion 1 Given the change, the thermal desi.gn. flow which is assumed in NSSS design and non-TTDP UFSAR Chapter 14 analyses will still be met.
Existing departure from nucleate boiling margin has been allocated to offset the change in mini.mum measured flow i,n the remaining UFSAR Chapter 14 analyses.
- Thus, the change is not expected to involve a significant increase in the probability or consequences of a previously analyzed accident.
Criterion 2
The change will not change the desi.gn or ooeration of the plant.
Thus, it would not be expected to create
=he possibility of a new or different kind of accident from any accident previously analyzed or evaluated.
Criterion 3
The original thermal design flow analyses assumptions for NSSS design and non-ETDP UFSAR Chapter 14 analyses will be met, and existing departure from nucleate boiling margin has been allocated for ITDP UFSAR Chapter 14 analyses.
- Thus, the subject evaluations have been demonstrated to comply with the licensing basis of the plant and in fact, involve no reduction in previ.ously reported analysis results.
Therefore, although the change may be construed as involving a reduction in the margin of safety, this will not be significant from a safety or licensing viewpoint.
We note that the Commission has provided guidance concerning the determination of significant hazards by providing examples (48 FR 14870) of amendments considered not likely to involve significant hazards consideration.
The sixth of these examples refers to changes that may reduce in some way a safety margin, but the results of which are within limits established as acceptable.
As di.scussed
- above, the change is consistent with our licensing basis
ATTACHMENT 1 TO AEP:NRC:1130 Page 3
and involves no reduction in results of analyses previously reported to the NRC.
Therefore, we believe the example cited is applicable and that the changes should not involve significant hazards consideration.
References 1)
"Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for American Electric Power D.
C.
Cook Unit 1 Nuclear Power Station" VCAP 12568, April 1990.
Provided as Attachment 4 for the reviewer's convenience.
ATTACHMENT 2 TO AEP:NRC:1130 PROPOSED REUISED TECHNICAL SPECIFICATION PAGES
ATTACHMENT 3 TO AEP:NRC:1130 WESTINGHOUSE EVALUATION FOR A REDUCTION IN THE TECHNICAL SPECIFICATION MINIMUM MEASURED FLOW REQUIREMENT
0
ATTACHNENT 1 DONALD C.
COOK NUCLEAR PLANT UNIT 1
SAFETY EYALUATION FOR A REDUCTION IN THE TECHNICAL SPECIFICATION MINIMUM MEASURED FLOW REQUIREMENT Attachnert 1 to AEP-90-231 l<S-OPLS-OPL-li-90-463 Page 1 of 18
BACKGROUND Technical Specification (T/S) 3.2.5 (Table 3.2-1) requires that the Reactor Coolant System (RCS) Total Flow Rate be greater than or equal to an indicated value of 366,400 gpm.
This Minimum Heasured Flow (MMF) requirement assures that RCS flow meets the assumptions used in the NSSS design calculations and the accident and transient analyses.
For those DNB transients that are analyzed using the Improved Thermal Design Procedure
( ITDP) (Table
- 1) the initial RCS flow is assumed to be the HHF.
For these
- analyses, uncertainties associated with the flow measurement are incorporated into the DNBR limit value.
For the NSSS design calculations, non-DNB related accident and transient analyses (Table 2),
and DNB transient analyses for which the ITDP is not used (Table 2), the initial RCS flow assumed is the Thermal Design Flow (TDF),
which is 354,000 gpm for Cook Nuclear Plant Unit 1.
For these
- analyses, flow uncertainty is accounted for by the Technical Specification requirement (HHF) being larger than the TDF plus the calculated uncertainty.
The purpose of this safety evaluation is to determine the impact of a reduction in the Cook Nuclear Plant Unit 1 Technical Specification HHF requirement on the NSSS design calculations and the accident and transient analyses.
The HMF requirement reduction is from 366,400 gpm to 361,600 gpm.
EVALUATION All of the Cook Nuclear Plant Unit 1 safety analyses and the NSSS design calculations have been reviewed to determine the impact of a reduction in the RCS MHF Flow Technical Specification requirement from 366,400 gpm to 361,600 gpm.
This section summarizes the effects of a reduction in the HMF requirement on the safety analyses and the NSSS design calculations.
Page 2 of 18 Attec.".-.e~t 1 to A=P-90-231 NS-QPLS-QPL-11-90-463
NSSS Desi n Calculations As stated
To assure that there is no impact on the TDF, the minimum measured flow requirement must be larger than the TOF plus uncertainties.
The measurement uncertainty associated with MMF is 2. 1% (Reference 1).
The proposed change to lower the MMF to 361,600 gpm will, therefore, not impact TDF (354,000 gpm).
361,600 - 0.021 x 361,600 354,006.4
> 354,000 Since there is no impact on the TDF, the NSSS design calculations are not affected by the change to the MMF requirement.
Non-ONB Safet Anal ses and Non-ITDP ONB Safet Anal ses These safety analyses also assume a minimum RCS flow consistent with TOF.
Since there is no impact on the TOF, the non-ONB and DNB non-ITOP safety analyses are also not affected by the change to the MMF requirement.
DNB ITOP Safet Anal ses A reduction in MMF for Cook Nuclear Plant Unit 1 impacts the analysis of the DNB events analyzed with ITDP in three areas:
- 1) the DNBR design limits and/or margin will be affected,
- 2) the Reactor Core Safety Limits (T/S Figure 2. 1-1) will be affected, which can potentially affect the Overtemperature hT and Overpower 4T reactor trip setpoints (T/S Table 2.2-1) that are based on these core limits, and 3) the initial RCS flow value assumed is reduced.
Each of these areas is discussed in detail below.
Page 3 of l8 Attach;ent j to Ar.P-90-231 NS-CPLS-CPl.-j l-90-463
l.
ONBR Calculations The design limit ONBR values for Cook Nuclear Plant Unit 1 are 1.323 for the thimble cell and 1.334 for the typical cell.
These are the minimum DNBR values that are required in order to meet the DNB design basis for analyses that use the ITDP.
- However, the safety analyses conservatively use a safety limit DNBR of 1.45 for both thimble and typical cells.
This allows for DNBR penalties to be offset with available margin between the design and safety limit DNBR values.
A reduction in core flow is a
penalty with respect to the calculation of DNBR.
For Cook Nuclear Plant Unit 1, sufficient margin is available and has been allocated to accommodate the MMF reduction.
- 2. Reactor Core Safet Limits and Protection Set pints The Reactor Core Safety Limits (T/S Figure 2. 1-1) represents the loci of points of thermal
- power, RCS average temperature, and RCS pressure for which the DNBR is equal to the safety analysis limit ONBR (1.45) or the average enthalpy at the vessel exit is equal to the saturated liquid enthalpy (vessel exit boiling).
For a given RCS pressure, the lines at high power levels are DNBR limits, while the lines at lower powers are vessel exit boiling limits.
The vessel exit boiling limits are lowered from the actual saturated enthalpy line to account for flow uncertainty, since the RCS MMF requirement (T/S 3.2.5) is the same as that used as the basis of the Reactor Core Safety Limits.
The current T/S Figure 2.1-1 is based on a
MMF of 366,400 gpm.
The DNBR lines remain unchanged-for the MMF reduction, since DNBR margin has been allocated as described above.
However, the vessel exit boiling lines represent a physical limit (bi ased for flow uncertainty as described above).
A reduction in the RCS flow results in these lines becoming more restrictive.
That is, for a given pressure and power, exit boiling will occur at a lower RCS temperature.
Thus, the Reactor Core Safety Limits figure in the Technical Specifications should be revised.
Page 4 of 18 At:acht;.ent
) to AEP-90-231 NS-CPLS-OPL-I !-90-463
The Overtemperature hT and Overpower hT reactor trips are designed to prevent the Reactor Core Safety Limits from being exceeded.
Since the flow is reduced and the vessel exit boiling limits change, the current setpoints for these trip functions were examined.
Setpoint calculations were performed using the methodology described in Reference 3.
These calculations confirm that the current trip setpoints continue to provide core protection for the reduced MMF.
As noted in Reference 2, it is assumed that the reference average temperature (T'nd T) in the setpoint equations are rescaled to the full power average temperature each time the cycle average temperature is changed.
Also, the reference pressure (P') in the Overtemperature hT setpoint equation is assumed to be set to the appropriate nominal primary system pressure (2100 psia or 2250 psia).
- 3. Initial RCS Flow Assum tion The transient analysis of DNB events that use the ITDP assume the MMF for the initial RCS flow value.
The MMF reduction could affect the results of these analyses in two ways:
- 2) by affecting the calculated system transient.
For DNB events analyzed with the
- ITDP, DNBR margin has been allocated to offset the penalty associated with the MMF reduction.
Thus the conclusion that the DNB design basis is met for these events remains valid.
See above for details on the use of DNBR margin.
The system transient response
.is primarily governed by the initiating event and any subsequent control or protection system actuations.
The transient results of interest would not be affected by the initial RCS flow assumption unless the reduction in flow was sufficiently large to significantly affect the steady-state core or steam generator heat transfer capability.
This is not the case for the small reduction of about 1.3% being considered for Cook Nuclear Plant Unit 1.
Therefore, it can be concluded that the transient results previously calculated for the Page S of 18 Attach".ent 1 to AEP-90-231 WS-OPLS-QPL-II-90-463
events analyzed assuming HHF remain valid for the reduction in HMF to 361,600 gpm.
To confirm this conclusion, sample sensitivity cases were run for an Uncontrolled RCCA Bank Withdrawal at Power analysis.
The sensitivity cases represented a range of reactivity insertion rates, varying only the initial RCS flow assumption (361,600 gpm vs. 366,400 gpm}.
The results showed insignificant differences in the transient results in parameters such as RCS temperatures, pressures, and time of reactor trip.
The safety analyses that assume the HMF are specifically addressed below.
Uncontrolled RCCA Bank Withdrawal at Power FSAR 14. 1.2 1990 u date This analysis is documented in Section 3.3.4.4 of Reference 2.
ONB margin has been allocated to offset the penalty associated with the MMF reduction.
The Overtemperature hT reactor trip setpoint is not changed.
In addition, the calculated system transient would not be affected by the relatively small flow reduction.
The results and conclusions of the previously applicable safety analysis remain valid.
Rod Cluster Control Assembl Misali nment FSAR 14.1.3 1990 u date This analysis is documented in Section 3.3.4.5 of Reference 2.
ONB margin has been allocated to offset the penalty associated with the HMF reduction, such that the dropped rod limit lines used in the cycle specific ONB evaluation remain valid.
In addition, the calculated system transient would not be affected-.by the relatively small flow reduction.
The results and conclusions of the previously applicable safety analysis remain valid.
Loss of Reactor Coolant Flow -
FSAR 14.1.6 1990 u date The loss of flow and the locked rotor rods-in-ONB analyses are documented
'n Section 3.3.4.7 of Reference 2.
The normalized flow coastdown is determined by the pump characteristics and not the initial flow Page 6 of 18 Attachment 1 ta AEP-90-231 NS-OPLS-GPL-11-90-463
assumption.
The system transient is unaffected by this relatively small reduction in HHF.
DNB margin has been allocated to offset the penalty associated with the HHF reduction, such that the calculated DNBR at the limiting point in the transient is unaffected.
The results and conclusions of the previously applicable safety analysis remain valid.
Loss of External Electrical Load FSAR 14.1.8 1990 u date This analysis is documented in Section 3.3.4.8 of Reference 2.
DNB margin has been allocated to offset the penalty associated with the HHF reduction.
In addition, the calculated system transient would not be affected by the relatively small flow reduction.
The results and conclusions of the previously applicable safety analysis remain valid.
Excessive Heat Removal Due to Feedwater S stem Malfunctions-FSAR 14.1.10 1990 u date This analysis is documented in Section 3.3.4. 10 of Reference 2.
DNB margin has been allocated to offset the penalty associated with the HHF reduction.
In addition, the calculated system transient would not be affected by the relatively small flow reductions The results and conclusions of the previously applicable safety analysis remain valid.
Note that the zero power case was recently reanalyzed.
The analysis documented in Reference 4 is applicable to Cook Nuclear Plant Unit 1.
The system transient analysis used to calculate the maximum reactivity insertion rate would not be significantly affected by small changes in the initial RCS flow.
The DNB evaluation for the zero power case did not use the ITDP.
Excessive Increase in Secondar Steam Flow -
FSAR 14. 1. 11 1990 u date This analysis is documented in Section 3.3.4. 11 of Reference 2.
DNB margin has been allocated to offset the penalty associated with the HHF reduction.
In addition, the calculated system transient would not be affected by the relatively small flow reduction.
The results and conclusions of the previously applicable safety analysis remain valid.
Page 7 of 18 Attachment 1 to AEP-90-231 NS-OPLS-OPL-11-90-463
CONCLUSIONS r
The Cook Nuclear Plant Unit 1 safety analyses and NSSS design calculations have been evaluated for the reduction in the RCS MHF given in the Technical Specifications from 366,400 gpm to 361,600 gpm.
It has been determined that the conclusions of the safety analyses remain valid for the reduction in the RCS flow and the NSSS design calculations are unaffected by the reduction in the MHF requirement.
Based on the above conclusions, the reduction in the HHF requirement does not involve an increase in the probability or consequences of an accident previously evaluated or involve a reduction in a margin of safety.
The MMF requirement reduction is an operational relaxation, not a plant hardware modification.
As concluded
- above, the MMF requirement reduction does not affect the NSSS design calculations, Therefore, the HMF requirement reduction does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Based on the above, the proposed Cook Nuclear Plant Unit 1 Technical Specification changes described below related to the MHF requirement reduction do not involve a significant hazards consideration.
RECOMMENDEO TECHNICAL SPECIFICATION CHANGES 1.
Safety Limit 2. 1. 1 - The Reactor Core Safety Limit Figure (2. 1-1) was revised to reflect the change in Reactor Coolant System (RCS)
Minimum Measured Flow.
2.
Safety Limit 2.2 '
- The Reactor Trip System Instrumentation Setpoints were revised to address the change in Minimum Measured Flow
[specifically, Table 2.2-1 Functional Unit 12's footnote (Loss of Flow) was revised to reflect the new Minimum Measured Flow].
3.
Tech Spec 3/4.2.5
- Table 3.2-1 was revised to reflect the reduced HHF requirement.
Marked up Technical Specification pages are attached.
Page 8 of 18 Attach;,ent 1 to AEP-90-23l HS-OPLS-OPL-lt-90-463
REFERENCES 1.
"Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for American Electric Power D.
C.
Cook Unit 1 Nuclear Power Station,"
WCAP-12568 (Proprietary),
WCAP-12569 (Non-Proprietary),
April 1990.
2.
"Reduced Temperature and Pressure Operation for Donald C.
Cook Nuclear Plant Unit 1 Licensing Report,"
WCAP-11902, October 1988.
Submitted to NRC by AEP:NRC: 1067, October 14, 1988.
3.
"Design Bases for the Thermal Overpower hT and Thermal Overtemperature hT Trip Functions,"
WCAP-8746-A, March 1977 (original version),
September 1986 (approved version).
4.
90AE*-G-0089, W/AEP2-0083, "Reanalysis of Feedwater Flow Malfunction at Zero Power for Donald C.
Cook 2
VANTAGE 5 Fuel Transition,"
May 17, 1990.
Page 9 of 18 Attach-..eat 1 to AEP-90-231 HS-OPLS-OPL-Il-90-463
TABLE I DNB TRANSIENTS ANALYZED USING ITDP 1.
Uncontrolled RCCA Withdrawal at Power 2.
Rod Cluster Control Assembly Misalignment / Dropped Rod 3.
Loss of Reactor Coolant Flow 5.
6.
Loss of External Electrical Load Excessive Heat Removal due to Feedwater System Malfunctions Excessive Increase in Secondary Steam Flow Page 10 of 18 Attac..rent I to AiP-90-231 AS-OPLS-OPL-II-90-463
TABLE 2 NON-DNB RELATED ACCIDENTS AND TRANSIENTS AND DNB RELATED TRANSIENTS WHICH DO NOT USE ITDP 2.
Large Break LOCA Small Break LOCA 3.
LOCA Hydraulic Forcing Functions 4.
5.
6.
Post-LOCA Hot Leg Recirculation Time Post-LOCA Long Term Core Cooling Steam Generator Tube Rupture 7.
Steamline Break Mass/Energy Releases
- B
- 9 Startup of an Inactive Loop Uncontrolled RCCA Withdrawal from a Subcritical Condition 10.
Chemical and Volume Control System Halfunction ll.
Locked Rotor 12.
Loss of Normal Feedwater Flow 13.
Loss of All AC Power to the Plant Auxiliaries
- 14.
Rupture of Steam Pipe 15.
Rupture of Control Rod Drive Mechanism Housing 16.
ContainmeBt Analyses
- 17.
Excessive Heat Removal due to Feedwater System Malfunctions (zero power case)
- DNB related events Page ll of 18 Attach-.ent
} to ASP-90-231 AS-GPLS-GPL-lI-90-463