ML20135E855
| ML20135E855 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/28/1997 |
| From: | Singh A, Weir L COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20135E845 | List: |
| References | |
| NUDOCS 9703100279 | |
| Download: ML20135E855 (164) | |
Text
1 ATTACHMENT !
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l VERIFICATION SCREENING OF KEY PARAMETERS FOR TWELVE RISK SIGNIFICANT SYSTEMS l
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TABLE OF CONTENTS Introduction Tab A i
Sectton Pane Number i
! Purpose and Background A1 i Comed Commitments AI j Project Personnel A-2 l Report Organization A-2 i
l Summary and Conclusions Tab B i
Section Page Number f
Introduction B-1 Summary of Results by System B-2 Conclusions B-6 Screening Methodologr Tab C Section Pane Number Screening Verification Methodology C-1 l Selected Systems C-1 Key Parameter Selection and Screening Verification C-3 i NRC Notification C-7 Table C-1: Risk Achievement Worth (RAW) of Twelve Risk Significant Systems C-8 (Dresden Unit 2)
Table C-2: Risk Achievement Worth (RAW) of Remaining Systems and PSA Model C-10 l Nodes (Dresden Unit 2) l Table C-3: Risk Achievement Worth (RAW) of Twelve Risk Significant Systems C-12 (Dresden Unit 3)
Table C-4: Risk Achievement Worth (RAW) of Remaining Systems and PSA Model C-14 Nodes (Dresden Unit 3)
Table C-5: Activities Related to Design Basis and Calculation Verification and/or C-16 Reconstitution Effort Figure C-1: Process for Identification of Key Parameters C-17
. Figure C-2: Relationship of Key Parameters to Design Basis Parameters C-18 Figure C-3: Development of Design Basis Parameters C-19 Figure C-4: NRC Notification Process C 20 ii Vef- Screening orKey Pm Rewmoe i
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f Summary of Key Parameter Screening Tab D I Section Page Number i
Summary Overview D1
} Safety-Related 125 Vdc System D-3 l Safety-Related 250 Vdc System D-10 i Low Pressure Coolant Injection (LPCI) System D 15 Containment Cooling Service Water (CCSW) System D-29
, Feedwater/ Condensate System D-37
! Turbine Building Closed Cooling Water (TBCCW) System - D-45
) Main Steam Safety and Relief Valves D-48
- Service Water (SW) System D-55 l l Automatic Depressurization System (ADS) D-59 l 4 kV Safety Related Auxiliary Power System D-63 j j 480 Vac Safety Related Auxiliary Power System D-69 i Isolation Condenser System (including Makeup Water) D-72 i j Offsite Power System D-87 l l Emergency Core Cooling System (ECCS) Initiation Logic D-91 1 High Pressure Coolant Injection (HPCI) System D-98 !
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j Discrepancy Disposition Tab E 4 Section Page Number l 3
} Discrepancy Summary E-l I
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0 iii voirama screen.no of Key Pvamm Remon 0
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INTRODUCTION Puroose and Background The purpose of this verification screening of the key parameters of the twelve risk significant systems was to determine the key parameters for the twelve most important systems from a risk perspective, determine the numerical values for the parameters, and verify that a calculational basis exists to support the key 1
parameters, if proper calculations do not exist or if the calculations are inconsistent, the discrepancy was identified and dispositioned.
1 ComFd Commitments l The key parameter screening effort was a commitment from Comed to the Nuclear Regulatory l Commission (NRC) that was made as a follow up to the Dresden independent Safety Inspection (ISI) that l
was performed in late 1996. This effort is among the actions being implemented to provide confidence that
( the key operating parameters for the twelve risk significant systems are supported by design basis calculations This activity was committed to under a letter from Mr. J. Stephen Perry to Mr. A. Bill Beach *. The follow-up Confirmatory Action Letter (CAL) was sent from Mr. Beach to Mr. Perry m, These two letters committed to the following activities for this effort:
A screening of key parameters on the twelve systems most important from a risk perspective. The j screening includes a review of key operating parameters of the twelve risk significant systems to
! verify that calculations exist to support those parameters. This screening is to be completed by the
( end of February,1997.
e The NRC will be immediately informed if critical parameters on any of the twelve systems selected i for screening are discovered to be outside of normal acceptance values. The NRC will also be informed if calculations cannot be located.
l The results of the screening performed on the twelve systems selected from a risk perspective will be provided to the NRC on a monthly basis through a meeting and docketed response.
These commitments are being tracked through Nuclear Tracking System (NTS) number 237-121 96 01609.
In addition to the key parameter screening for the 12 risk sil,nificant systems, the letter JSPLTRit96-0125 also commits to the following tasks to establish the ' Adequacy and Retrievability' of the design basis:
On an ongoing basis, as design modifications are made, validation or reconstitution of the design basis and/or calculations will be performed on the equipment and portions of systems specifically affected by the modification.
- Design basis and calculation validation and/or reconstitution will be performed for the twelve systems most important from a safety perspective per the Dresden Probabilistic Risk Assessment. This effort will ensure that design information and calculations support system functional and testing l requkemenu.
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l JSPLTR #96-0125, Letter from Mr. J. Stephen Perry to Mr. A. Bill Beach, dated November 8,1996.
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- CAL No. Rill-96-016, Letter from Mr. A. Bill Beach to Mr. J. Stephen Perry, dated November 21,1996.
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l The information developed during the performance of the key parameter screening form an important set ofinputs to the design basis and calculation validation and'or reconstitution program. Therefore, the information developed from the key parameter screening will be transferred for use in the design basis and calculation validation and/or reconstitution program.
Project Personnel
! The project was executed under the general direction of E. Conmil, the Comed Project Owner. As l committed to in Reference (I), the screening was performed F y a dedicated team of senior experienced engineering personnel brought from outside of CemFd, and supported by appropriated Comed System and l Design Engineering personnel.
The following is a listing of the team members involved:
i Name Position /Discioline E. Connell Comed Project Owner A. K. Singh Key Parameter Screening Team Lead F. Fischer Electrical L Uddin Electrical R. Ka:ita Instmmentation and Control J. Neurauter hiechanical A. Ostenso Instrumentation and Control N. Weber hiechan? cal C. Johnson hiechanica:
R.Engel hiechanical J.Reda hiechanical(Support) l J. htatthews Electrical (h.nport) l 1
The project was supported by the following Comed lead engineers:
Name Position /Discioline B. Tsai Comed Nuclear Fuel Service (NSF) Support l L. Raney/ R. Johnson PRA Twelve System Selection J. Dawn hiechanical Design Engineering Lead S. Tutich Electrical Design Engineering Lead J. Tietz System Engineering Lead D. Evans System Engineering Lead D. Spencer System Engineering Lead J. Sword System Engineering Lead Reoort Organization This report is organized into five main sections. These are the introduction, the summary and conclusioas, the methodology for the screening review, the result of the screening and the discrepancy disposition.
j The introduction provides the reason why this project was undertaken and the objectives of the project stemming from the commitments that were made during the Independent Safety Inspection at the Dresde i Station.
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Section B contains the summary and conclusions of the verification screening of key parameters for the twelve risk significant systems.
Section C identifies the twelve systems selected for the verification screening of the key parameters, the method used to identify the key parameters for the systems selected, and the NRC notification process. 1 Also described are calculation review activities which are being performed to verify that calculations '
support the key parameters. Calculation revalidation and reconstitution activities which will be included in the longer term Design Basis and Calculation Verification and'or Reconstitution effort are also listed for l
completeness. ;
I In Section D, the screening summary presents the results of the key parameter screening for the twelve l
systems determined to be the most significant from a risk perspective. The section is broken into a I I
summary for each of the twelve systems. Each system summary contains a system description, a list of the system major components, the system operating modes, the system key parameters, the value for that parameter, and the documents that provide the basis for that value. The summary also identifies any potential discrepancy that is found during the review process.
The final section of this report contains a listing of the discrepancies that were identified during the review and a disposition for these discrepancies. It is the intent to use this listing as a means to track the resolution l of open items identified in this report to ensure that closure of these items will be completed.
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SUMMARY
AND CONCLUSIONS Introduction JSPLTR #96-0125 commits to two interrelated programs: (1) actions to verify adequacy of key safety systems and (2) actions to verify adequacy and retrievability of design basis. This report summarizes the results of the actions to verify the adequacy of key safety systems through screening of key parameters for the twelve systems most important from a risk perspective.
, The second program, actions to verify the adequacy and retrievability of design basis, is intended to revalidate and reconstitute the design basis for the same twelve risk significant systems. He results from the screening verification of key parameters for the twelve risk significant systems will be an important set ofinputs to this program. The second program will establish the detailed system design parameters necessary to demonstrate the acceptable performance of the system. As a part of this program, an acceptable set of system design basis calculations will be revalidated, revised or reconstituted, as appropriate.
The screening verification of key parameters fv .welve risk significant systems, was intended to demonstrate compliance with the safety analysis bases for twelve selected risk significant systems. There were three primary tasks associated with this program. He first task was to identify the key parameters derived from the safety analysis for these twelve systems. Three principal sources ofinformation were used to identify these parameters: (1) the current safety analysis;(2) station TSUP/ Technical Specifications; and (3) the DATR. The second task was to quantify the values for the key parameters consistent with the safety analysis bases. The relevant safety analysis documentation, TSUP/ Technical Specification and bases, and DATR were used in the process for quantifying the key parameters. The third task was to retrieve the calculations that support the values of the key parameters consistent with their use in the safety analysis process. As a part of the third task, any supporting calculationi not found were l identified for subsequent action in the system design basis program. Any discrepancy noted was identified and these discrepancies will be tracked to closure through the Nuclear Tracking S ystun (NTS).
The twelve systems selected for the key parameter screening to verify that existing calculations support the key parameters are based on the twelve most risk significant systems as determined by the system Risk Achievement Worth (RAW) determined as part of the Dresden Individual Plant Examination (IPE). The twelve systems selected for the verification screening of key parameters were:
- 1. Safety-Related 125 Vdc System la. Safety-Related 250 Vdc System
- 2. Low Pressure Coolant Injection (LPCI) System 2a. Containment Cooling Service Water (CCSW) System
- 3. Feedwater/ Condensate System
- 4. Turbine Building Closed Cooling Water (TBCCW) System
- 5. Main Steam Safety and Relief Valves
- 6. Service Water (SW) System
- 8. 4 kV Safety Related Auxiliary Power System 8a. 480 Vac Safety Related Auxilhry Power System
- 9. Isolation Condenser System (including Makeup Water)
- 10. Offsite Power System i 1. Emergency Core Cooling System (ECCS) Initiation Logic
- 12. High Pressure Coolant Injection (HPCI) System B-1 Venr<auca Screen ag of Key Pwancert itswpop 0 =
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i 4 Summary of Results by Systtm 1
i The following provides a brief summary of the results of the key parameter screening review for the twelve j most important systems from a risk perspective.
SafetyRelated125 VdcSystem i
l The safety related 125 Vdc system is designed to provide a reliable source of de motive and control power I to various station systems and components. Calculations exist that support the key parameters for the j system. The screening review for this system identified six potential discrepancies, with none considered
] major. Some duplication of calculation purpose or scope was noted in the calculations reviewed. These calculations do not have appropriate cross references to the related calculations. Other discrepancies require revision of the calculation or clarification of the rebaseline updated final safety analysis report (UFSAR). Discrepancies for this system will be resolved and tracked through NTS.
t i Safety Related250 Yde System The safety related 250 Vdc system is designed to provide a reliable source of de motive and control power j to various station systems and components. Calculations exist that support the key parameters for the
] system. The screening review for this system identified two potential discrepancies, with none considered
. major. The other discrepancy was identified that requires a clarification in the technical specifications. A l final discrepancy is the lack of a calculation to address the battery requirements for station blackout.
j Discrepancies for this system requiring future action will be resolved and tracked through NTS.
i l Low Pressure Coolant injection System (LPCI)
The LPCI is designed to remove reactor core decay heat, as well as sensible heat, from the reactor and containment in the event of a loss of coolant accident (LOCA). The screening review of this system identified twelve potential discrepancies. The most significant of these potential discrepancies are related to inconsistencies between calculations and system performance requirements or design basis document requirements. However, system operability is ensured through system testing and supporting evaluations.
A number of other discrepancies were associated with differing values for key parameters between various design documents or the unavailability ofcalculations. Other existing calculations require revision to assure key parameters are consistently evaluated and used. New calculations need to be prepared to address some of the key parameters. Discrepancies for this system will be resolved and tracked through NTS.
Containment Cooling Service IVater (CCSID System The CCSW provides cooling water to the tube side of the LPCI/CCSW heat exchangers to remove heat from the primary containment and limit the temperature of the suppression pool. The screening review of this system identified five potential discrepancies. Three of the discrepancies were associated with the lack of formal calculations. However, testing of these systems ensures their operability. The other discrepancies are associated with inconsistencies between various documents. Discrepancies for this system will be resolved and tracked through NTS.
B-2 Veriricanoe kresung o(Key Parmaetes, Resepea 0
Feedwater. Condensate System The feedwater/ condensate system provides condensate from the main condenser to the reactor at a sufficient ra;e to support full power operation. The screening review for this system identified one potential discrepancy. This discrepancy is associated with the lack of specific system calculations. The basic conclusion reached for this non-safety-related system is that the plant is operating and providing the required flows at the correct conditions to support power operation. The discrepancy for this system will be resolved and tracked through NTS.
Turbine Building Closed Cooling Water (TBCCID The TBCCW is a closed loop system that provides cooling water to the systems and components located in the turi >ine building and crib house. He screening review for this system identified two potential 1 discrepancies. Rese were: (1) calculations to support the flowrate through the TBCCW system could not j be found; and (2) calculations to establish the basis for the heat transfer rate of the TBCCW heat l
exchangers could not be found. Data for the selected key parameters were available for the most part, but i not in total. The basic conclusion reached for this non-safety-related system is that the plant is operating l and therefore, the TBCCW is operating and providing the required flows at the correct conditions to the various systems it services. Discrepancies for this system will be resolved and tracked through NTS. ,
l Main Steam Safety and Relief Valves ne main steam safety and relief valves protect the reactor from overpressurization and provide the capability for rapidly depressurizing the reactor to allow the functioning of the LPCI and core spray systems. The screening review for these components identified two potential discrepancies. These discrepancies are associated with inconsistencies between documentation and the lack of specific i calculations. The basic conclusion reached for these components is that they are capable of fulfilling their l required functions. Discrepancies for this system will be resolved and tracked through NTS.
Service Water (SW) System l
The service water system is an open loop system that provides strained cooling water to various plant equipment in plant as well as to the plant cooling systems for heat removal. The screening of this system did not identify any discrepancies. A calculation for the system hydraulic model that supports the service water systems major key parameter, f.owrate :o the TBCCW heat exchanger, was found. The hydraulic model system was compared to the system design basis values using a line-up representing normal operating conditions of the system. The model appears to reasonably represent the system. Normal plant operation ensures that this system is performing as intended.
Automatic Depressuri:ation System (ADS)
The ADS initiation circuitry provides the capability for the reactor pressure vessel under conditions where vessel level cannot be maintained by the normal operating or high pressure makeup systems. Calculations supporting all key parameters were found. The nreening review for this system identified one potential discrepancies. This discrepancy is associated witn an inconsistency between calculations and design basis documents. No system operabiliy concerns were identified. However, documents require updating to assure consistency. The discrepancy for this system will be resolved and tracked through NTS.
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4 k VSafety Related Auxiliary Power System The 4 kV safety related auxiliary power system provides power to both safety related and non-safety related electrical equipment that are supplied by the system buses. The system voltage calculations were l found to have adequate system margins but, in many cases were limited to the bus voltage only. Critical 4 kV calculations are primarily in support of downstream 480 Vac loads which are the limiting case for the auxiliary power system. The reasoning being, that if the 4 kV voltage is maintained at an adequate level which will support the operation of the limiting case, the system is sufGcient. The screening review for this system identified two potential discrepancies. Various clarifications and corrections of several documents
- need to be performed to ensure design adequacy and provide consistency between design documents, with l none considered major. These documents include existing calculations, the UFSAR, and various procedures related to this system. The terminal voltage available when large motors are starting / running are not explicitly calculated in most cases. Discrepancies for this system will be resolved and tracked through NTS.
480 Vac Safety Related Auxiliary Power System l
l He 480 Vac safety related auxiliary power system provides power to both safety related and non-safety l
related electrical equipment that are supplied by the system buses. The system key parameters are supported by the ac auxiliary power system calculations. The screening review for this system did not identify any discrepancies.
- Isolation Condenser System (including Atakeup Water) l The primary function of the isolation condenser system is to provide core decay heat removal in the event the reactor becomes isolated from the main condenser. The screening review of this system identified twelve potential discrepancies. The most significant of these potential discrepancies are related to the lack of calculations to support key parameters established by the system design basis, particularly related to the performance of the isolation condenser and makeup water systems. However, system testing, plant i surveillances, or diversity of water sources ensures continued system operability. Otner discrepancies were associated with key parameters values differing between various design documents or the unavailability of l calculations. Some existing calculations require revision to assure key parameters are consistently !
evaluated and used. New calculations need to be prepared to address some of the key parameters. ;
Discrepancies for this system will be resolved and tracked through NTS.
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Ofsite Power System 1
The offsite power system is designed to provide a reliable source of electrical power to the Dresden 138 kV and 345 kV switchyards and to the plant auxiliary power system. In general, the offsite power system was found to be in good shape for Dresden. The assessment of the offsite power system was somewhat limited since the operation of the Comed bulk power transmission system is not controlled from the station. As such, documentation to support the operation of the offsite power system was not nearly as detailed as that of a calculation to support a nuclear application. However, adequate documentation was retrievable to I support offsite power key parametars (voltage). De screening review for this system identified a potential l discrepancy. Some documentation which provides a guideline for operation of this system requires l revision to ensure the system ineing operated and maintained within appropriate values. The other discrepancies address minor inconsistmies. The discrepancy for this system will be resolved and tracked through NTS.
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l Emergency Core Cooling System (ECCS) Initiation Logic i
The ECCS initiation logic consists of the instrumentation designed to initiate the various systems that comprise the ECCS. Calculations supporting all key parameters were found. The screening review for tnis system identified two potential discrepancies. These discrepancies are associated with inconsistencies between calculations and design basis documents. No system operability concems were identified.
However, documents require updating to assure consistency. Discrepancies for this system will be resolved and tracked through NTS.
High Pressure Coolant injection (HPCI) System The HPCI system is designed to maintain reactor water coolant inventory to provide for adequate core cooling in the event of a postulated LOCA which does not result in rapid depressurization of the reactor pressure vessel. The HPCI system can also provide a source of water for high pressure makeup. The screening review of this system identified eight potential discrepancies. These potential discrepancies are .
related to the lack of calculations to support key parameters estab!!shed I,y the system design basis or inconsistencies between design basis documents. However, system testirg ensures continued system operability. Some existing documents require revision to assure key g>arameters are consistently evaluated and used. New calculations need to be prepared to address some of t' e key parameters. Discrepancies for this system will be resolved and tracked through NTS.
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i Conclusions
} Based on the screening review of the key parameters for the twelve risk significant systems, it can be
.; concluded that sutlicient test and/or calculational basis exists to support safe operation at Dresden.
Although fifty six (56) Performance Improvement Forms (PIF) or Nuclear Tracking System (NTS) items were initiated to document the noted discrepancies during the review, only one resulted in a system being declared in-operable.
Retrievability of the calculations has shown significant improvement during the screening review of the key parameters for the twelve risk significant system when compared with the NRC ISI audit. This was due to a more complete and readily available index of the calculations from EWCS, and that more of the l
- calculations had been transferred from Downers Grove and AEs to the site. !
The discrepancies identified during the screening review of the key parameters for the twelve systems were similar to those identified during the Dresden self assessment and the NRC ISI audit at Dresden. We 4
believe that the activities in initiated at Dresden to improve the performance of the engineering
, organization as described in JSPLTR #96-0125 are sufficient to resolve the finding of this review. The
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i specific actions being taken at Dresden include-
- 1. Formation of an Engineering Assurance Group to ensure the soundness of the current engineering l
activities.
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- 2. Revision of Nuclear Engineering Procedures to provide clearer guidance on calculation control and specific direction for the treatment of potential design basis discrepancies. 1 1
- 3. A continuing validation or reconstitution of the design basis or calculations as modifications are made l for the twelve most risk significant systems.
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- 4. The conduct of a two year reconstitution and revalidation program of calculations for the twelve most risk significant systems.
- 5. A review of the updated final safety analysis report (UFSAR) requirements against the design basis documents.
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METHODOLOGY SsigmJng Verification Methodology This section identifies the twelve systems selected for the verification screening of the key parameters, the method use i to identify the key parameters for the systems selected, and the NRC notification process.
Also desc ibed are calculation review activities which were performed to verify that calculations support the key pometers. Calculation revalidation and reconstitution activities which will be included in the longer term Design Basis and calculation verification and'or Reconstitution effort are also listed for completeness.
Selected Systems The twelve systems selected for the key parameter screening to verify that existing calculations support the key parameters are based on the twelve most risk significant systems as determined by the system Risk j Achievement Worth (RAW) determined as part of the Dresden Individual Plant Evaluation (IPE). The j Dresden IPE is being reviewed by the NRC
- The RAW numbers reprernt the C tor increase in risk resulting from a system failing (or being removed fron service) while at power win all other systems l available. A RAW of 2 means that the risk is doubled.
Tables C 1 through C-4 provide a listing of the " top nodes" utilized in the Dresden 2 and Dresden 3 modified PSA model for the IPE. Each " node" has a RAW value and is grouped into related groups of 1 systems / components / functions. Table C-1 identifies the top twelve (12) Dresden 2 system / components /
functions. Table C-2 lists the RAW for the remaining systems / components / functions for Dresden 2. !
Tables C-3 and C-4 list the corresoonding information for Dresden 3.
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The difference between the Dresden 2 and 3 models and the top twelve listing is driven by the implementation of the modification to eliminate the Feedwater Pump trips following the loss of 125V DC on Dresden 2. The upcoming modification to Dresden 3 will cause its PSA model to duplicate the Dresden 2 model. To determine the important Dresden 3 systems / components / functions, the equivalents of the identified Dresden 2 systems were utilized.
The importance of any given system / component / function is based upon how often it is required to respond to the set of initiating events and the reliability / availability of the system itself and its support equipment.
Examples are:
- Each individual ECCS injection path is oflittle importance because there are so many paths available.
- Each Diesel Generator is oflittle importance because of the availability of off-site power, the bus cross-ties and the other Diesel Generators.
. 'Ihe 125V DC buses are important due to the number of systems' trains that would be disabled by its loss.
The 125V DC batteries and chargers are not as important.
The twelve systems selected for the verification screening of key parameters are:
- 1. Safety-Related 125 Vdc System la. Safety-Related 250 Vdc System
- 2. Low Pressure Coolant Injection (LPCI) System 2a. Containment Cooling Service Water (CCSW) System
- 3. Feedwater/ Condensate System (3) Dresden Station Units 2 and 3, Response to NRC Review ofIndividual Plant Examination Submittal-Intemal Events, June 28,1996 C-I w cm,. so.~sa x.y % am ,. o
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4 2 4. Turbine Building Closed Cooling Water (TBCCW) System
- 5. Nfain Steam Safety and Relief Valves
- 6. Service Water (SW) System
- 3. 4 kV Safety Related Auxiliary Power System 8a. 480 Vac Safety Related Auxiliary Power System
- 9. Isolation Condenser System (including hiakeup Water)
- 10. Offsite Power System l 11. Emergency Core Cooling System (ECCS) Initiation Logic i
- 12. High Pressure Coolant bjection (HPCI) System The following information ularifies the extent of the top twelve Dresden systems:
I. The buses dominate the importance for the 125 and 250V DC system. The batteries and chargers are of lesser importance, but should be included within the system.
- 2. The CCSW and the portion of LPCI that is required for containment heat removal is the second most imponant " system (s)". The initial list
- showed LPCI as separate system, however it is for the heat removal function - the injection function is ranked 19 in importance.
- 3. The feedwater and Condensate have multiple parallel paths and components, each is oflittle irnportance as long as there is a source of high pressure water for RPV injection.
- 5. The Main Steam relief and safety valves are required to prevent over-pressurization of the primary system, which may lead to a LOCA.
- 6. Service Water is suppo-t for the TBCCW and thus the Feedwater/ Condensate system.
- 7. The Automatic Depressurization System is obvious, but the valves individually are not very important.
Only one or two valves are required to depressurize the RPV.
- 8. The AC buses are important to supply the ECCS equipment and the containment heat removal function.
The individual feeds to the buses are not as important as demonstrated by the importance of the Diesel.
Generators.
- 9. The Isolation Condenser and its make up sources are important. Again, any given make up source is not significant, because of the many back-up sources. The diesel driven make-up pumps are more significant due to their independence of off. site power.
- 10. Off-site power was initiallyW included within the AC power system, but is separated for better clarity, and for comparison with the Diesel-Generators.
- 11. Common Actuation is a " quasi" system established for the PSA model. The instruments and relays (such a low RPV level, high DW pressure, etc.) that are within a given ECC System that provide actuation signals for other systems are included within this " quasi" system.
- 12. HPCI is rated here because of its availability / reliability in relationship to the Feedwater, Isolation Condenser and the ADS systems / components / functions.
(4) Information Presented to the NRC, December 19,1996, at the Public hfecting held at Dresden Station C2 Ver skaison $sroenung et Key Parameesrs, Rewmece 0
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Table C 2 lists the remaining systems in their decreasing order ofimpenance. There are a few other nodes identified in the PSA model that are not specifically systems that affect core damage, these are operator i actions, containment venting hardware and other miscellaneous nodes.
Kev Parameter Selection and Screenine Verification The methodology used for screening the key parameters for the twelve systems most important from a risk perspective is consistant with the station commitments stated in the " Actions to Ensure Current Status of Key Safety Parameters", which is contained in Dresden letter JSPLTR #96-0125 from Mr. J. Stephen Perry to Mr. A. Bill Beach, dated November 8,1996.
l The purpose of the screening process was to identify and quantify the key parameters for the subject twelve systems. In addition, calculations supporting the quantification of the key parameters were identified. The identified calculations were reviewed to ensure that calculations supported the values of the key parameters. The system boundaries are those defined under the probabilistic risk assessment used ta establish the twelve most important systems from a risk perspective. Work performed relative to i identifying and quantifying the key parameters of the twelve s; stems were independently reviewed. The l screening process can be summarized as follows:
l identify the system components that are important for system operation and performing the system function ide.,tify the key system op: rating modes, which includes their initiation points and the system function for that mode '
. identify the key oper6g parameters for the system / components e determine a numerical value for the key parameter e
identify the design reference which provides this numerical value for the key parameter e identify available mechanical, electrical, instmment and control, and structural calculations which provides the basis of the numerical value for the key parameter a
for those calculations used as a basis reference, establish adequacy (completeness) of the design calculation scope, inputs, and analysis performed
. address differences between Unit 2 and Unit 3 systems, if any
. identify missing or incomplete calculations for key parameter values e initiate Performance Improvement Forms (PIFs) for missing calculations e initiate PlFs for incomplete calculations e
initiate PIFs for variance between the Rebaseline Updated Final Safety Analysis Report (UFSAR) and the Technical Specifications / Technical Specification Upgrade Program (TSUP) e perform Operability Evaluations, when required 4
The activities that are not part of the screening verification process but will be included in the two year Design Basis and calculation verification and/or Reconstitution effort are listed in Table C-5.
The documents used to perform the semning review were as follows:
Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR)
Technical Specification Upgrade Program (TSt P) with Dresden Technical Specifications TSUP/ Technical Specification Bases Dresden Administrative Technical Requirements (DATR)
Dresden Calculation Database Index System Design Basis Document (DBD)
. Referenced Calculations in the DBD Environmental Qualification (EQ) Binders C3 Venficarma $creening of Key Parammers flemce 0
l Seismic Qualification Utilities Group (SQUG) Evaluation Systematic Evatuaiion Program (SEP) Evaluation input Parameters for Safety Analysis (NFS)
System Notebooks (IPE/PSA) i System Notebooks (System Managers' Engineers)
General Electric Design Basis Data (Specifications) l The process for identifying the key parameters is shown in Figure C 1. As shown in Figure C 1, there are three sets of parameters that comprise the key system parameters. These are (1) critical parameters, (2)
TSUP/ Techaical Sr ecification parameters, and (3) DATR parameters.
The critical pmmeters are derived from the current plant safety analysis. These system performance parameters are assun.ed in the analysis of the events that are considered a part of the overall plant safety analysis. These parameters must be maintained or the plant would be considered to be operating in an l unanalyzed condition. For the purpose of this report, the current safety analysis consists of:
- 1. The event analysis contained in the Rebaseline Updated Final Safety Analysis Report (UFSAR)
Chapter 15
- 2. Containment analysis described in UFSAR section 6.2
- 3. Emergency core cooling system performance analysis described in UFSAR section 6.3
- 4. Evaluation of anticipa:ed transients without scram (ATWS)
- 5. Evaluation of station blackout
- 6. Analysis of the safe shutdown systems capability for fire protection
- 7. The standby liquid control system capability analysis
- 8. Overpressure protection analysis
- 9. Stability analysis and evaluation A review of the UFSAR and supporting documentation was performed to establish the critical parameters associated with the current safety analysis for the subject twelve systems. The critical parameters are the system input to the current safety analysis. These inputs consist of the following:
- 1. The system performance requirements
- 2. Analytical instrument limits
- 3. Monitored parameters assumed for operator actions (Emergency Operating Procedures)
- 4. Constraints on initial conditions for events
- 5. Constraints on event analysis assumptions Other supporting information that was used to establish the critical parameters includes the following:
- 1. Current reload analysis documentation
- 2. Documentation of Systematic Evaluation Program (SEP)
I
- 3. Specific analysis documentation associated with specific events (i.e. ATWS)
In developing the critical parameters, the screening review process considered both the specific events that require the system operation and the operating mode of the system for that event. The critical parameters for each of the twelve systems were determined for both the event and the operating mode. Where bounding values have been used to cover a spectrum of events, the bounding critical parameters associated with the limiting event were used. This approach enables the critical parameters to be characterized consistent with the safety analysis requirements.
The second group of parameters are associated with the station Technical Specification Upgrade Program (TSUPy Technical Specifications. As required by 10CFR50.36, the TSUP/ Tech Specs are derived from the analyses and evaluations in the safety analysis report and amendmet ts thereto and are issued by the C-4
%.= s- . .r m., e., ,s st o
r i
)
i NRC. In addition, the NRC may include additional technical specifications as it finds appropriate. There j are five categories of technical specifications:
1
- 1. Safety limits and limiting safety system settings
- 2. Limiting conditions for operation
- 3. Surveillance requirements
- 4. . Design features
- 5. Administrative controls Only the first four categories of technical specifications potentially contain parameters that may be considered key parameters. Administrative controls contain provisions relating to organization and management, procedures, record keeping, review and audit, and reporting, which are not associated with the work being performed during the screening review process. A summary statement of the bases for the TSUP/ Tech Specs is provided with them. The bases are not a part of the technical specifications but are an important source of information relative to the rationale for the selection of the specific technical specification requirements.
A review the of station TSUP/ Technical Specifications was performed to establish the associated parameters for the twelve risk significant systems. The technical specification bases were also identified and associated with the applicable parameters.
The DATR parameters are associated with station commitments. There are two primary sources of information that are included in the DATR. These are:
- 1. Items that have been relocated from previous station. Technical Specifications 2 Items that are in the standard technical specifications but not in the station Technical Specifications The screening review process reviewed the DATR to establish the associated DATR parameters for the twelve risk significant systems. Where applicable, the DATR parameters were associated with the critical parameters identified above.
Examples of key parameters are the following:
Mechanical System Flow, Temperature and Pressure j System Performance (Head vs. Flow) i Heat Exchanger Heat Transfer, Flow and inlet temperature .
Volume of Tanks !
Valve Opening and Closing Times !
Pump NPSH Requirements -
Maximum Area Temperature l Motor Brake HP l
Motor Voltage Requirements !
Analytical or Process Limits for instrument setpoints Nuclear Safety AnalysisInputs Accident Analysis inputs Anticipated Accidents Without Scram (ATWS) Analysis Assumptions Containment Analysis inputs Station Blackout Analysis Assumptions Safe Shutdown Analysis for Fire Protection Analysis Assumptions j l
Cs v.nne.so .tx, % a-o j i
l Electrical and Instrumentation and Control j System Voltage Limits
- Battey Sizing Requirements l Electrical Protection and Coordination j Short Circuit Current Levels at Supply Buses j Equipment AmbientTemperatures l Initiation Permissives I initiation Timing Functions j Channel Actuation Logic 1
l For each of the key parameters identified by the process described above, existing system calculations were i reviewed to verify that the calculations support the values of the parameters. This was accomplished through the use of the system calculation tog specifically developed for each of the twelve risk significant systems. In this process, the calculation log was used to identify those calculations that most likely contain j
supporting information relative to quantification of the key parameters. He selected calculations were
! retrieved and reviewed to obtain the relevant information. Any supporting calculations found through this j process were associated with the specific key parameter. If no calculation support was found for a specific l parameter, it was so noted for future reconstitution as part of the Design Basis and calculation verification j and'or Reconstitution effort for these systems.
i
} The key parameters are a subset of the overall system design basis parameters. The key parameters are defined to satisfy safety analysis requirements or license or plant specific regulatory commitments. The l remaining system design basis parameters are selected to support the key parameters or assure acceptable j system performance, considering the overall relationship of the plant systems. The overall relationship of j the plant systems include providing the required system supporting functions. He relationship of the key parameters to the design basis parameters is shown on Figure C 2. The sources ofinformation supporting
, the identification and quantification of the remaining design basis parameters include:
1 4
- 1. De UFSAR
- 2. System Design Basis Documents (DBD)
- 3. Other design documents The identification and quantification of the other system design basis parameters is beyond the scope of this project. However, it is considered a part of the commitment in JSPLTR 196-0125 Design Basis and l calculation verification and/or Reconstitution effort which is being tracked as NTS Items 237-121 96 1 01607 and 237 121-96-01608. I Figure C 3 shows a simplified process for the development of the design basis parameters. It is provided to demonstrate the distinction between the key parameters and the system design basis parameters. The process for the development of both the key parameters and the design basis parameters is dependent on I the highly interactive system design and analysis and safety analysis processes. j i
The station is designed consistent with the set station design, safety, and licensing requirements, which take the form of the applicable industry codes and standards, regulatory requirements and guidance, and station specific licensing commitments. This set of requirements is the same for both the systems design and analysis process and the safety analysis process.
l In the design process, the design and configuration of the plant systems is based on a set of system design criteria and requirements. Based on the system design, a system performance analysis is performed to I establish the system capabilities and conformance to system design requirements and criteria, it is this process that establishes the majority of the system design basis parameters. I 1
C-6 vaw so .t x.y :. . ;m o
The safety analysis process is performed to demonstrate that the station does not pose an undue risk to the public health and safety. The safety arialysis is based on a set of safety analysis inputs. The majority of the safety analysis inputs are derived from die systems design and configuration and performance analyses.
Other inputs are derived consistent with the station design, safety, and licensing requirements. The primary interaction between the safety analysis and system design and analysis process is at the level of the safety analysis results. If the results of the safety analysis meet the applicable event analysis limits, the safety analysis is considered acceptable. If the results of the safety analysis do not meet the applicable event analysis limits, the system design and configuration or system performance analysis is modified and a new safety analysis is performed. This is continued until an acceptable set of safety analysis results has been 4
obtained. The key parameters are derived from the safety analysis inputs that are used to obtain an acceptabic safety analysis. The key parameters form the remainder of the design basis parameters.
NRC Notification The NRC enforcement letter (and Comed internal commitments) required that the NRC be immediately informed if critical parameters on any of the 12 system selected for screening are discovered to be outside of normal acceptance values or when the calculations verifying the critical parameters cannot be located.
The process used for informing the NRC is shown in Figure C-4. As shown in the figure, a P!F was initiated w hen a deficiency (missing or incomplete calculation to support a key operating parameter) was confirmed. The PlF was reviewed by the Team Lead. The designate Comed lead noted the immediate ;
corrective action and forwarded the PIF to the shiit supervisor to screen the PIF for reportability using the Comed Reportability Manual In addition, for any missing or 'outside of normal acceptance values
- calculations which are determined to lead to ' degraded
- operability or 'in operable' condition for the plant, the Lead Design Engineer informed the Regulatory Assurance Lead on the condition. The Regulatory Assurance Lead informed the NRC 'immediately' of critical parameters that r.re di: covered to be outside of j nonnat acceptance values or when the calculation or other basis supporting the critical parameters cannot be located. All discrepancies were discussed with the NRC at the scheduled monthly meetings between the l
Dresden Station and the NRC.
C-7 Verdemana Smaieng of Kev Pam Re,w ese 0
TABLE C-1: RISK ACIIIEVEMENT WORTil(RAW) OF TWELVE RISK SIGNIFICANT SYSTESIS (DRESDEN UNIT 2)
SYS NODE j RAW DESCRIPTION j i 125V/250V DC l l 2h11 8010.00 UNIT 2 STAIN 125VDC BUS l 1 2RI 299.00 UNIT 2 RESERVE 125VDC BUS 1 3hil 58.00 UNIT 3 htAIN 125VDC BUS I 3RI 1.00 UNIT 3 RESERVE 125VDC BUS la 2P2 609.00 UNIT 2 REACTOR BUILDING 250VDC BUS 2,2a LPCl/CCSW COh1PONENTS FOR CONTAINMENT HEAT REMOVAL (principally SPC) 2 LP 245.00 LPCI PUh1PS 2 SPC 238.00 HARDWARE FOR SUP POOL CLG
-~
2 LPC 3.47 HARDWARE TO ALIGN CCSW TO LPCI HX 3 FEEDWATER 3 FW 21.60 FEEDWATER (WITH CONDENSATE) 4 TURBINE BUILDING CLOSED COOLING WATER 4 2TB 21.20 UNIT 2 TBCCW 5 MAIN STEAM RV/SRVs 5 RVO 20.70 RV/SRVS OPEN 5 RVC 1.47 RV/SRVS CLOSE 6 SERVICE WATER 6 SW 19.90 SERVICE WATER (SHARED BY BOTH UNITS) 7 AUTOMATIC DEPRESSURIZATION 7 ADS 19.80 AUTOMATIC DEPRESSURIZATION SYSTEM 8 4KV/480 V AC 8 231 15.70 UNIT 2 BUS 231 8 241 13.00 UNIT 2 BUS 241 8 24 5.02 UNIT 2 BUS 24 8 23 3.66 UNIT 2 BUS 23 8 34 1.02 UNIT 3 BUS 34 8 33 1.00 UNIT 3 BUS 33 8 331 1.00 UNIT 3 BUS 33-1 8 341 1.00 UNIT 3 BUS 341 8a 28 14.00 UNIT 2 BUS 28 8a 29 4.75 UNIT 2 BUS 29 8a 26 1.39 UNIT 2 BUS 26 8a 27 1.20 UNIT 2 BUS 27 8a 38 1.02 UNIT 3 BUS 38 8a 25 1.01 UNIT 2 BUS 25 8a 39 1.00 UNIT 3 BUS 39 C-8 ven6 casus Smaunt of Key Panmeert Re=ssion 0
- - .-_ . . - -. _ . . - - - - ~ _ . _ . - . . = . - . . . _ - - _ = _ _ - . _ - -- . - _ . . . . . .. .-
l SYS NODE RAW DESCRIPTION l
9 ISOLATION CONDENSER (including makeup) l 9 (CHI 12.70 PROPER OPERATION OF IC HARDWARE (EARLY) 9 AIC 10.50 AUTO INITIATION OF ISOLATION CONDENSER 9 (CH2 10.50 PROPER OPERATION OF IC HARDWARE (LATE) 9 MUP 5.04 ADDITION Of SHELL SIDE MAKEUP 10 OFFSITE POWER 10 ROPl 10.40 RECOVERY OF OFFSITE POWER (EARLY) l 10 ROP 2 1.15 RECOVERY OF OFFSITE POWER (LATE) i 10 ROPIA 1.00 RECOVERY OF OFFSITE POWER AFTER ROPI (SBO4 ONLY) l 11 COMMON ACTUATION (including parts oILPCI, CS, HPCI that actuate other systems) 1I 2CA 9.50 UNIT 2 COMMON ACTUATION 12 HIGH PRESSURE COOLANT INJECTION 12 HPl 9.03 AUTO INITIATION! OPERATION OF HPCI (SINGLE START) 12 HP2 4.72 AUTO INITIATION / OPERATION OF HPCI(MULTIPLE STARTS)
C-9 Va.fkama $asemag of Key tw Reumoe 0
_ -- - - - . - _ - . - - - ~ _ . - . - - . . _
TAllLE C 2: RISK AClllEVEMENT WORTil(RAW)OF REMAINING SYSTEMS AND PSA MODEL NODES (DRESDEN UNIT 2) 4 SYS NODE l RAW l DESCRIPTION 13 STANDBY LIQUID CONT ROL 13 SLC 7.15 S'.C SYSTEM HARDWARE 14 ATWS SYSTEM 14 AT 7 Ci AUTOMATIC ATWS SYSTEM INITIATION
! 14 RPTl 6.71 AUTOMATIC RECIRC PUMP TRIP 14 RPT2 1.35 MANUAL RECIRC PUMP TRIP 14 ARI l.00 AUXILIARY ROD INSERTION i I i 15 DIESEL GENERATOR 1/2 15 DGB 3.26 DO 2/3 STARTS AND RUNS l
- 16 CORESPRAY l 16 CS 2.55 CORE SPRAY l ,
l 16 CS2 1.01 CORE SPRAY SUCCEEDS FOR 1 HR(ISLOCA) l? DIESEL G ENERATOR 3 J DG3 2.46 DG 3 STARTS AND RUNS 18 DIESEL GENERATOR 2 18 DG2 2.33 DG 2 STARTS AND RUNS 4 19 LPCI INJECTION 19 LV 2.31 LPCI INJECTION VALVES 20 CONTROL ROD DRIVE - high pressure injection
, 20 CRD 1.04 CRD CROSSTIE TO OTHER UNIT
? STANDBY COOLANT SUPPLY 21 SBCS 1.03 HARDWARE FOR CONTAINMENT FLOODING 22 CONTAINMENT SPRAY 22 CNTS l.00 HARDWARE FOR CONTAINMENT SPRAYS 4
l 23 FIRE PROTECTION - low pressure injection i
23 FP 1.00 DIESEL FIRE PUMP STARTS AND RUNS FOLLOWING ARE NODES IN PSA MODEL BUT ARE NOT CONSIDERED SYSTEMS OR ARE NOT RELEVANT TO CORE DAMAGE (CONT VENT)
- OSPC 459.00 OPERATOR ACTION TO ALIGN FOR SUP POOL CLG l OAD 35.50 OPERATOR ACTION TO INITIATE ADS OMUP 34.50 OPERATOR ACTION TO PROVIDE MAKEUP TO IC OIADS 23.30 OPERATOR ACTION TO INHIBIT ADS (ATWS) l OSLI 1.80 OPERATOR ACTION TO INITIATE SLC (t/2 PUMP)
C 10 Ventkmeon Saesmos o(Key Pwm Rew 0 4
SYS NODE RAW DESCRIITION OIC 1.60 OPERATOR INITIATES IC MC i 1.53 AVAILABILITY OF MAIN CONDENSER AFTER ATWS FWA 1.46 FEED /COND FOLLOWING ATWS (% OF IEs THAT ARE LOFW)
RCFM l.44 FAILURE MODE (E/M) OF RPS ORP 1.43 OPERATOR ACTION TO INITIATE RPT OHX 1.31 OPERATOR ALIGNS CCSW TO LPCI HX OCST 1.27 OPERATOR ACTION TO ALIGN ECCS SUCT TO CST OIC2 1.23 OPERATOR ACTION TO PREVENT LODC FLR OF IC OSL2 1.20 OPERATOR ACTION TO INITIATE SLC (2/2 PUMPS)
WW-DW l.18 LOCATION OF CONTAINMENT FAILURE 4
OAT 1.17 OPERATOR ACTION TO INITIATE ARI SYSTEM OVNT 1.09 OPERATOR ACTION TO VENT CONTAINMENT SVW l.09 HARDWARE FOR 2 IN. WETWELL VENT OCRD 1.04 OPERATOR ACTION TO CROSSTIE CRD OSBCS 1.03 OPERATOR INITIATES STANDBY COOLANT SUPPLY OAL 1.02 OPERATOR ACTIONS TO CONTROL RV LEVEL AFTER ATWS
~
OFW l.01 OPERATOR ACTION TO RESTORE FW INJ OSS 1.01 OPERATOR ACTION TO RECOVER SW/TBCCW
- RC 1.00 REACTIVITY CONTROL LVD 1.00 HARDWARE FOR 10 IN. DRYWELL VENT LVW l.00 HARDWARE FOR 10 IN. WETWELL VENT OAVR I.00 OPERATOR ACTION TO VENT RX VESSEL FOR CONT FLDG
! OCNTS 1.00 OPERATOR ACTION TO INITIATE CONTAINMENT SPRAYS OFP 1.00 OPERATOR ACTION TO CONNECT FP TO FW I O!S 1.00 OPERATOR ACTION TO ISOLATE ISLOCA (VLVS, PMPS, FLO) i l
SVD 1.00 HARDWARE FOR 2 IN. DRYWELL VENT i
i i
1 C-11 Ventsmann $ menses o(Key Paramvers. Revumn 0
TABLE C-3: RISK ACHIEVEMENT WORTil(RAW) OF TWELVE RISK SIGNIFICANT SYSTEMS (DRESDEN UNIT 3)
L l
SYS NODE RAW DESCRIPTION (D3 equivalent to D2 PSA Model Node)
I 2MI 5420.00 UNIT 3 MAIN 125VDC BUS I 2R1 508.00 UNIT 3 RESERVE 12aVDC BUS l 1 3M1 39.70 UNIT 2 MAIN 125VDC BUS 1 3R1 1.00 UNIT 2 RESERVE 125VDC BUS la 2R2 417.00 UNIT 3 REACTOR BUILDING 250VDC BUS ;
2 241 313.00 UNIT 3 BUS 'l4-1 2 24 115.00 UNIT 3 BUS 34 2 231 11.00 UNIT 3 BUS 331 2 2's 2.81 UNIT 3 BUS 33 2 34 1.01 UNIT 2 BUS 24
_2 33 1.00 UNIT 2 BUS 23 2 331 1.00 UNIT 2 BUS 23-1 2 341 1.00 UNIT 2 BUS 24-1 2a 29 307.00 UNIT 3 BUS 39 2a 28 9.79 UNIT 3 BUS 38 2a 26 1.57 UNIT 3 BUS 36 2a 27 1.43 UNIT 3 BUS 37 2a 25 1.01 UNIT 3 BUS 35
~
2a 38 1.01 UNIT 2 BUS 28 2a 39 1.00 UNIT 2 BUS 29 3 LP 171.00 LPCI PUMPS 3 SPC 162.00 HARDWARE FOR SUP POOL CLG 3 LPC 2.67. HARDWARE TO ALIGN CCSW TO LPCI HX 4 ADS 39.80 AUTOMATIC DEPRESSURIZATION SYSTEM 5 FW 15.00 FEEDWATER(WITH CONDENSATE) 6 2TB 14.80 UNIT 3 TBCCW 7 RVO 14.40 RV/SRVS OPEN 7 RVC 1.32 RV/SRVS CLOSE 8 SW 14.10 SERVICE WATER (SHARED BY BOTH UNITS) 9 2CA 10.90 UNIT 3 COMMON ACTUATION l 10 HPl 10.40 AUTO INITIATION / OPERATION OF HPCI (SINGLE START) 10 HP2 3.88 AUTO INITIATION / OPERATION OF HPCI(MULTIPLE STARTS) l C-12 Venfr son Screenmg of Key Parm llevmon 0
. - ~ . . . . . .. . . . . . . . . - . .. . . - - . - . . . _.
i i
i a
SYS NODE RAW DESCRIPTION (D3 equivalent to D2 PSA Model Node)
II ICHI 8.92 PROPER OPERATION OF IC HARDWARE (EARLY) 1I AIC 7.42 AUTO INITIATION OF ISOLATION CONDENSER i
II ICH2 7.42 PROPER OPERATION OF IC HARDWARE (LATE) li MUP 3.74 ADDITION OF SHELL SIDE MAKEUP 4 I
- 12 ROPl 7.36 RECOVERY OF OFFSITE POWER (EARLY) 12 ROP 2 1,10 RECOVERY OF OFFSITE POWER (LATE) 12 ROPIA 1.00 RECOVERY OF OFFSITE POWER AFTER ROPl (SBO4 ONLY) i 4
a 1-1
)
i 4
i l.
I l
d 4
I 4
4 4
C-13 v ic - sc .rx.y %. w- o 4
. . - - - - - - . . . . . . ~ . ...
TABLE C-4: RISK ACHIEVEMENT WORTH (RAW) OF REMAINING SYSTEMS AND PSA MODEL NODES (DRESDEN UNIT 3)
't SYS NODE RAW DESCRIPTION (D3 equivalent to D2 PSA Model Node) 13 LV 6.36 LPCI INJECTION VALVES I4 SLC 5.17 SLC SYSTEM HARDWARE 15 AT 5.11 AUTOMATIC ATWS SYSTEM INITIO TON 15 RPTl 4.87 AUTOMATIC PECIRC PUMP TLP I
15 ARI 1.00 AUXILIARY ROD INSERTION
- 16 CS 3.76 CORE SPRAY 16 CS2 1.00 CORE SPRAY SUCCEEDS FOR I HR(ISLOCA)
~
17 DGB 2.53 DG 2/3 STARTS AND RUNS 18 DG3 1.99 DG 2 STARTS AND RUNS
? 19 DG2 1.93 DG 3 STARTS AND RUNS i
20 CRD 1.03 CRD CROSSTIE TO OTHER UNIT 4 21 CNTS 1.02 HARDWARE FOR CONTAINMENT SPRAYS 22 SBCS 1.02 HARDWARE FOR CONTAINMENT FLOODING 23 FP 1.00 DIESEL FIRE PUMP STARTS AND RUNS OSPC 615.00 OPERATOR ACTION TO ALIGN FOR SUP POOL CLG OAD 50.50 OPERATOR ACTION TO INITIATE ADS OMUP 23.70 OPERATOR ACTION TO PROVIDE MAKEUP TO IC i OIADS 16.10 OPERATOR ACTION TO INHIBIT ADS ( ATWS)
OCST 4.49 OPERATOR ACTION TO ALIGN ECCS SUCT TO CST OSLI 1.54 OPERATOR ACTION TO INITIATE SLC (1/2 PUMP)
OIC 1.40 OPERATOR INITIATES IC MC 1.36 AVAILABILITY OF MAIN CONDENSER AFTER ATWS FWA 1.31 FEED /COND FOLLOWING ATWS (% OF IEs THAT ARE LOFW)
RCFM 1.30 FAILURE MODE (E/M) OF RPS ORP 1.29 OPERATOR ACTION TO INITIATE RPT RPT2 1.24 MANUAL RECIRC PUMP TRIP OHX l.21 OPERATOR ALIGNS CCSW TO LPCI HX
- SVW l.21 HARDWARE FOR 2 IN. WETWELL VENT OVNT 1.20 OPERATOR ACTION TO VENT CONTAINMENT OIC2 1.15 OPERATOR ACTION TO PREVENT LODC FLR OF IC OSL2 1.14 OPERATOR ACTION TO INITIATE SLC(2/2 PUMPS)
.; C 14 Venficane Screening of Key Parm Rewmon 0
l l
SYS NODE RAW DESCRIPTION (D3 equivalent to D2 PSA Model Node) j OAT 1.12 OPERATOR ACTION TO INITIATE ARI SYSTEM j WW DW l.12 LOCATION OF CONTAINMENT FAILURE OCRD 1.03 OPERATOR ACTION TO CROSSTIE CRD
~
l OCNTS 1.02 OPERATOR ACTION TO INITIATE CONTAINMENT SPRAYS OSBCS 1.02 OPERATOR INITIATES STANDBY COOLANT SUPPLY OAL 1.01 OPERATOR ACTIONS TO CONTROL RV LEVEL AFTER ATWS OFW l.01 OPERATOR ACTION TO RESTORE FW INJ I OSS 1.01 OPERATOR ACTION TO RECOVER SW/TBCCW I
! LVD l.00 HARDWARE FOR 10 IN. DRYWELL VENT l
- LVW ! .00 HARDWARE FOR 10 IN. WETWELL VENT l OAVR 1.00 OPERATOR ACTION TO VENT RX VESSEL FOR CONT FLDG OFP 1.00 OPERATOR ACTION TO CONNECT FP TO FW
) OIS 1.00 OPERATOR ACTION TO ISOLATE ISLOCA (VLVS, PMPS, FLO)
RC 1.00 REACTIVITY CONTROL RDG 1.00 REALIGN DG 2/3 TO OTHER UNIT SBO. 1.00 QUERY EP STATUS FOR STATION BLACKOUT
! SVD 1.00 HARDWARE FOR 2 IN. DRYWELL VENT l i I l
l
,i l
3 7
4 i
l 1
4 I
i i
i 1
C-15 Venfeson $creenmg of Key Parm Reveen 0
_ _ _ . - - - . . _ - . . . . _ . _ _ _ . _ . . _ . _ _ . . . _ _ _ . . _ __ _ . _ . _ . _ _m ._ . __
a 1
i i
TABLE C-5: Activities Related to Design Basis and Calculation Verification and/or Reconstitution Effort
- for the calculations used as the basis for the numerical value of a system key parameter, verification of the numerical accuracy of the input used in the calculation I e
for the calculations used as a basis reference, evaluation of the adequacy of calculation methodology e for the calculations used as a basis reference, verification of the numerical accuracy of the calculation
. Quality Assurance (QA) verification of vendor or third party design input for calculations used as basis i
references I
e performance of new calculations
{
. revisions of existing calculations e contact with vendor, A/E, and NSSS for locating missing calculations if currently does not exist in Dresden database e identification of similar calculations or vendor input from Quad Cities Station e verification of the accuracy of numerical values used as design references for system key parameters obtained from the Rebaseline Updated Final Safety Analysis Report (UFSAR) or Technical l
Specifications / Technical Specification Upgrade Program (TSUP) e verification of the adequacy ofISI/IST procedures or parameters contained in these parameters e identification of Operating Procedures that control and/or reflect the numeric parameters e identification ofIST procedures that verify system performance to the numeric parameters
)
These activities are not within the scope of the screening verification of the key parameters. l l
l l
C-16 verificmaan Screemas of Key Pm Remoe 0
l KEY PARAMETERS Safety Analys.is q Inputs + ...................v............., i i :
i These Parameters Must Be Reload Analysis m m Current - m Critical -
Inputs Safety Analysis : Parameters . > Mainta.med or the Plant is m .
- i an Unanalyzed Condition t
Systematic j j Evaluation Program + i !
Commitments j j ,
U 5 V 5 i i Technical Specification TSUP/ Tecn.ical TSUP/ Techn.ical -
i NRC Impcsed m : m Parameters are Derived l Spec.ficat.
i ion -: > from Safety Analysis in >
Requirements SPecifications ; .
Parameters i : Accorde::e with 10CFR50.36 V i. V i i
- #'#'8 ""
Plant Dresden Admin. DATR m
- .> > Adm? inistrat
."""ively Controlled Commitments Tech. Requirements : Parameters .
- : by the Plant i i Figure C-1: Process for Identification of Key Parameters verification screening of Key Parameters. Revision 0
{
DESIGN BASIS PARAMETERS i Key i i Parameters i
- ~.
! U ! .
Final Safety : -
. ' : Primary System i System Des.ign Requ.irements An lysis Report -
!> Design Basis > Necessary to Support Key i Parameters i Parameters Design Basis y y ! j Documents ! V i
- : System Des.ign Requ.irements !
j Support.ing System m Design Basis i > Necessary to Support Key , Other Des.ign Parameters and Primary System y j Parameters Design Basis Parameters Documents ! i Figure C-2: Relationship of Key Parameters to Design Basis Parameters Verification screening of C-18 .
I . Station Design, Safety, and j Licensing Requirements : i Regulatory Station Specific l i
- Industry Codes . :
- Requirements and L.icensing :
. and Standards :
- i Gu.i dance Commitments :
- Systems Design :: i
- :: Safety Analysis Process -
. and Analysis Process .: .
i System Design :i . :
- - :: m Safety Analysis :
Criteria and -
- Inputs :
Requirements ! ! :.
!. V i i.
V i. m j System Design 3 Safety i i and Configuration p : Analysis j i i! i i V !! V i
- . System -
m . :: Key : Perfom1ance - .-
. .: Parameters -
! Analyses !! !
I....................... ........... ..! .......... ... .......................! V U DESIGN BASIS PARAMETERS Figure C-3: Development of Design basis Parameters Verirication screening or C-19 **"*""""'"
- - . . - - - - _ . - - - - . - - - - - - . . - _ - . . - - - . ~ _ - - -
I f i Figure C-4: NRC Notification Process i If ; Discrepancy Revise ! No + PIF Screem.ng Identified y , Valid ? Evaluations ! KPST KPST ! j lI jf Yes Initiate Screen PIF for Inform PIF Reportability Operability N NRC Monthly Ki>ST SM Evaluation n "- I equire KPST- Lead lf Review Perform lf PIF Significance Perform Screen "E Operability Degraded o No KPST- Lead m SQV CAG Evaluation in-Operable ? 1f DE - Lead PIF KEY Yes > Disposition KPST : Key Parameter Screening Team II Inform DE - Lead DE : Design Engineering NRC I"I " SM : Shift Manager - Immediately eg Assmance - RA : Regulatory Assurance SQV CAG : SQV Corrective Action Group DE - Lead RA - Lead Verification screening or C-20
a
)
I l
SUMMARY
OF KEY PARAMETER SCREENING ; l Summary Overview 1 This section presents the summary of the results for the screening verification of key parameters to I determine that calculations support the key parameters. The summary results are presented in Tables D 1 through D-12 for the twelve systems as follows: ( l Risk Rank System Page Number i Safety-Related 125 Vdc System D-3 la Safety-Related 250 Vdc System D-10 l 2 Low Pressure Coolant Injection (LPCI) System D-15 2a Containment Cooling Service Water (CCSW) System D-29 3 Feedwater/ Condensate System D-37 4 Turbine Building Closed Cooling Water (TBCCW) System D-45 5 Main Steam Safety and Relief Valves D-48 6 Service Water (SW) System D-55 7 Automatic Depressurization System (ADS) D-59 8 4 kV Safety Related Auxiliary Power System D-63 l 8a 480 Vac Safety Related Auxiliary Power System D-69 1 9 Isolation Condenser System (including Makeup Water) D-72 10 Offsite Power System D-87 . I1 Emergency Core Cooling System (ECCS)lnitiation Logic D-91 12 High Pressure Coolant injection (HPCI) System D-98 He following is a template of the format used for the above tables. The following also gives specific reference to items that were not addressed for the subject twelve systems.
System Description
This section of Tables D-1 through D 12 gives a short, high level description of the system. It will describe that main functions of the system, and the type of components that make up the system. Kev System Components Comoonent EPN Descriotion This section of Tables D 1 through D-12 list those components that are important to system operation. These are major pieces of equipment, or equipment specifically needed for the system to perform its intended function.
- Re only valves that are listed here are those valves that are active (need to change state) in order for the system function to be accomplished.
D-1 Venfamme $meneng o(Key W Revisene 0
. - - - . - - - . - - . . ~ = - _ . --_ -. _ . _ -. -- _ . . _ . _ .
I d 4 i l I s ; ! Onerational Modes a i l This section of Tables D-1 through D 12 describes the system's significant modes that are utilized to i accomplish the system's design function. The initiation point (if applicable) is provided for each mode, as ) well as the function of the system for each mode. ' 3 , e Containment isolation is not considered a mode for any of the twelve system. Containment Isolation is i considered a separate system that was not chosen among the twelve most significant from a risk i perspective. Therefore, the containment isolation function is not addressed within the subject twelve , t systems. I Kev Parameters Parameter Value Parameter Reference Calculation Reference This section of Tables D-1 through D-12 defines the system's key operating parameters and ti:e numeric value for that parameter. The parameter reference is the document which the numerical value was obtained from. Parameter references are typically the UFSAR, TSUP/ Tech Specs, or DATRs. The calculation reference is the document that provides the acceptable basis for th> numerical value obtained from the parameter reference. These are typically formal calculations. Potential Discrepancies This section of Tables D-1 through D-12 lists inconsistencies that were identified during the screening review process. References 1 This section of Tables 0-1 through D-12 lists the references that were utilized during the screening review i process. l The first five references for each system are the following:
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision Ol A.
- 2. Dresden Station Unit 2 and 3 Technical Specifications Upgrade (TSUP), January 13,1997.
- 3. Dresden Station Unit 2 and 3 Technical Specification Upgrade Basis (TSUP Basis), January 13,1997.
- 4. Dresden Administrative Technical Requirements (DATR), September,1996.
- 5. Dresden Technical Specifications, December,1996.
All other references are specific to the system. D-2 Venrum Screemns of Key Param Revem 0
l SAFETY RELATED 125 VDC System Descrintion l The 125 Vdc System provides a reliable source of de motive and control power for various systems l and components throughout the station. i The 125 Vdc system consists of batteries, battery chargers, cabling, distribution buses and panels that are used to supply power to control circuits, diesel generator auxiliaries & excitation system switchgear, l turbine system, etc. The 125 Vdc system is designed so that each battery is sized to start and carry the l normal de loads plus all de loads required for safe shutdown of one unit and operations required to limit the consequences of a design basis event on the other unit for a period of 4 hours following the loss of ac sources. Kev System Compcnents Comoonent EPN Descriotion 2-8300-2(3-8300-3) 125 Vdc Battery Charger 2(3) 2 8300-2A(3-8300-3A) 125 Vdc Battery Charger 2A(3 A) 2-8300-2B(3-8300-3 B) 125 Vdc Alternate Battery Charger
- 2(3)-8300-BC 125 Vdc Main Battery 2(3)-8300 125A 125 Vdc Alternate Battery 3-8303A 125 Vdc Main Bus 3A 2-8302Al(3-8303Al) 125 Vdc Main Bus 2A-l(3 A 1) 2 8302A2(3 8303A2) 125 Vdc Main Bus 2A-2(3 A-2) 2-8302 125 Vdc Reserve Bus 2 2-8302B(3-8303B) 125 Vdc Reserve Bus 2B(3B) 2-83028-l(3 8303B-1) 125 Vdc Reserve Bus 2B l(3B-1) 2 83028-2(3 83038 2) 125 Vdc Reserve Bus 2B-2(3B-2) '
2(3)-83125 125 Vdc Reactor Building Distribution Panel 2(3) 2-83125-2(3 83125-3) 125 Vdc Battery Bus 2(3) Operational Modes l Operating Mode Initiation Not applicable; the de system is normally operable regardless of plant status. Function
-The 125 Vdc system provides de electrical power for normal de loads during plant operation and the required safe shutdown loads when necessary.
Station Blackout Mode l During a station blackout (SBO) on loss of all ac power in the plant, the batteries suppiv de power i for safe shut down of the plant for I hour SBO coping until ac power is avai!able from the 91temate (AAC) diesel generator. l l I D-3 v.nc- sm .,sxn % n ->.o
SAFETY RELATED 125 VDC Appendit R Mode The Dresden Appendix R Fire Protection Program, under the topic of Supporting Associated Circuit Analysis - Coordinated Fault Protection Analysis, mentions that to prevent potential faults in branch circuits from affecting the 125 Vdc system, all loads that are not essential for safe shutdown will be l tripped after the essential loads are started. Thus, the Dresden Appendix R Analysis primarily relies i on the procedural tripping of a particular list of associated circuits for a fire in a particular fire zone, i and it takes no credit for coordination of electrical protection. 1 Key Parameters l l Parameter Value Parameter Reference Calculation Reference l Battery Voltages - Normal (Float): 128.8 130.5 V Ref. 2, Sec. 3/4.9.C Mate: The battery ( for 58 cell U2 & U3 & DES 8300-07. float voltage is Main & U3 Alt. Battery); based on the battery i 133.2 135.0 V manufacturer's ! (for 60 cell U2 Alt. Battery) recommendation. - Minimum Float 125.9 V Ref. 2, Sec. 4.9.C. I.b See note above. V@:;;c ( for U2 & U3 Main and U3 Alt. Battery) 130.2V for U2 Alt. battery. - Minimum Battery Voltage: 105 V Ref.1, Sec. 8.3.2.2 Calc.7056-00-19-5, (minimum value applies at the end of design load profile) Rev. 31(12/20/96), U2/3,
" Load Estimation of 125 Vdc Buses";
Calc.8684-41-19-1, Rev 0 (6/19/90)," U2, 125 Vdc Alternate Battery Addition"; Calc.8706-41 19-1, Rev.1 (12/21/92),
" Unit-3 125 Vdc Alt.
Battery Addition". Ente: All four 125 Vdc batteries have a design minimum 105 Volts at the battery terminals. Since the U2/3 Main and U3 Alt. Batteries have 58 cells they discharge to 1.81 volts per cell (VPC) and the 60 cell U2 A' ternate Battery discharges to 1.75 VPC. D-4 Venrwauan $asenmg of Key Paramaten. Remum 6
SAFETY RELATED 125 VDC l Parameter Ve'as Parameter Refergnss Calculation Reference Load Profile / Duty Cycle 4 hour load profile Ref.1, Sections Cales. 7056-00-19-5, per Dresden ELMS- 8.3.2.2 & 8.3.3 Rev. 31, for U2 & U3 DC data base / load Ref. 2, Sec. 4.9.C.4 Main Batteries; l files. 8684-41 19 1, Rev.0, for U2 Alt. Battery;
& 8706-41 191,Rev.I for U3 Alt Battery.
Nats: Revision 31 of Calc. 7056-00-19-5 pertains to ISI Question 3164-000 on battery loads and sizing and it incorporates Comed response arid commmit-ments for this question. The Calc. Revision Summary (pages 291 A through 291 C) describes resolution ofNRC comments. The revised runs of ELMS DC incorporate the required I loading changes. These runs are for LOOP + LOCA scenario. Runs l for SBO loading were not I in the scope of this l revision (Revision 30 includes the SBO loading). Rev.3 I Conclusion (page l 310) shows that the units 2 and 3 main batteries have adequate capacity to supply their respective design basis duty cycle loads. The remaining capacity margins are i1.4% and and 18.3% for units 2 and 3, respectively, at 65 F, 1.25 aging factor, and 1.81 volts per cell end of l duty cycle discharge voltage. l l D-5 Verificance Screeneas of Key Paramm Revueen 0 l l
. _.m ._ __ __ .__ _ . . _ . - ., _ _ _ _. _... .._ . . _ - . . _ . . _ . _ _ _ _ . . _ _ .
4 i 4 1 ! I i } SAFETY RELATED 125 VDC i Parameter Value Parameter Reference Calcu'ation Reference Some inaccuracies / j discrepancies were noted during the 1 i cale. review and these are listed under" Potential j Discrepancies". j Calculations for the U2 i
& U3 alt. batteries have
- not been revised for the load changes since 1990 l for U2 (1992 for U3).
Since these batteries q have same size NCX 21 1 (U3) or larger NCX 27 j (U2) cells, they are } therefore covered for the l battery capacity by the bounding case of the
- I main batteries having 4
NCX-21 cells, which 4 have 11.4% & 18.3% remaining capacity j margins. l ! The above referenced f cales. for the alt. batteries are multi-scope and some i items of their scope / l purpose are now better i covered, in intent, by j other calculations . utilizing current load data and
- better methodology.
3 l Battery Minimum Capacity 2 80% of Ref. 2, Sec.4.9.C. 5 & 6 IEEE Std.485 l manufacturer's rating. Ref.7(IEEE 450) l; (up to 85% of battery service life) 4 j Cell Electrolyte Specific 11.200 (pilot cell) Ref. 2, Table 4.9.C 1 Nats: Typically these j Gravity & Temperature values are based on i > 1.195(connected cells) Ref. 2, Table 4.9.C 1 the battery 4 manufacturer's I > 60 F (average for all Ref. 2, Sec. 4.9.C.2.c recommendations. 3 connected cells) (See Ref.1, sec,8.3.2.4) d 1 D-6 v-__ .. - -. 4
i l i i SAFETY RELATED 125 VDC . Parameter Value Parameter Reference Calculation Reference I Battery Electrolpe - l Temperature 1 Minimum: 65 F Ref.1, Sec.8.3.2.1 Cale.7056-00-19-5 4 for Main Batteries ; j 8684-41-191 for U2 j Altemate Bat:ery; & j . 8706-41-191 for U3 Alternate Battery. 4 Ents: The design basis ! minimum electrolyte
- temperature is 65 F in the above calculations, i
Charger ac Supply Voltage: 2: 432 V Ref.8, Nameplate Data Cales.9198-1819 l& j
-3, Rev.l(12/31/96),
l "Dresden Safety 4 Related Continuous j Load Running / } Starting Voltages". Charger Rated de Supply 200 A for Ref. 2, Sec. 4.9.C.3.d Ref. 8, Nameplate data ] Current at least for 4 hours. ( It is based on the j equipment manufacturer's ! design calculations.) {
- DC System Ground <l25 kohm (Level 1) Ref.1 Sec.8.3.2.2 Ref.10 i Resistance Alarm > 20 kohm & < 125 kohm(Level II) & Table 8.3 8 l Set Points
- 120 kohm (Level III) Eqis: The de system i ground resistances
{ are monitored per l Ref. 9. l i System Electrical Full coordination Engineering Report SL-4500, j Protection does not exist for all Design. Vol. 3," Overcurrent l Coordination. 125 Vdc related circuits. Protective Device
- Coordination Study, i Dresden Station- Units j 2 and 3", 03/24/1989.
! Calc.5569 31 19 2, i Rev.2(9/12/89),
"125 Vdc System Breaker & Fuse
- Coordination".
\
1
?
1 1 } D7
% arta:ance $4rsemang o(Key Parmneters, Revision 0 W
g+ , , -c,- , _
SAFETY RELATED 125 VDC Parameter Yalug Parameter Reference Calculation. Reference Natt: The above study and calculation identify the 125 Vdc circuits which do not have acceptable coordination. Potential Discrepancies
- 1. Reference I lists 1495 AH rating for NCX-21 lead calcium battery cells for the 125 V batteries which does not apply to them since their design basis "End of Discharge Voltage"is 1.81 volts instead of 1.75 volts per cell. Listing the battery manufacturer's nominal 8 hour,77 F, rating of 1495 AH for NCX 21 cells in the RUSFAR and the system DBD is ambigous and potentially misleading when the cells are not discharged to 1.75 Volts. For a higher (1.81V) end of discharge voltage the cell capacity / usable rating will be less than 1495 AH.
- 2. The following potential discrepancies exist in Calc. 7056-00 19-5, Rev.31:
2.1 The Assumption listed on page 2B (Rev.30)is no longer applicable and should be deleted. 3 2.2 On page 9, Para 2, reference to 4 hour Station Blackout should be changed to one hour and the referenced page number should be 269(Rev.8 Summary) instead of 222. 2.3 Page 32(Rev.30) needs to be revis:d to list the new SF6 breaker types and their current data for 4 kV switchgear buses 23,24,33, and 34 to make it consistent with the data on page 40c. Also the revision number of this page (40c) should be 31 instead of 24. 2.4 The current time data shown on pagM S for U2 ATWS panel feed (Circuit 13, Bus 2A-1) following Modification M12-2-94-002 also applie. to Circuit 10, Bus 2B-1, listed on page 104(Rev.31). This should be made clear by adding a note on pge 45. 2.5 The load cycle period ( end time) should also w shown on page 222. 2.6 The description of MOD. M12-2-94 002 (Item 8, page alC, Rev.31) is not correct for the U3 ATWS feed circuits as this mod. is not yet implemented on U3. The load profile of 3.5 A from t= 0 to 240 minutes applies to only U2 circuits. For the U3 ATWS circuits / feeds the stepped load profile (10.39/3.5 A) shown on page 134A, Rev.31, is conservative and covers both the pre-mod and post mod loading. However, the U2 reference and mod. number on page 134A needs to be corrected to show that it applies to U3 feeds.
- 3. The following potential discrepancies exist in Calc. 8684-41 1o i, Rev.0 (U2 Alt.) .
3.1 Part A of this calculation concerning the old FPS 23 battery ce!ls is nc. longer relevant and it should be superseded in the next revision of this calculation. 3.2 Parts B, C, and D of this calculation are not current since they do not utilize the present load data utilized in Rev. 31 of cale. 7056-00-19-5. Part C may be superseded by cross referencing it to the updated 125 Vdc battery sizing calculation 7056-00-19-5, Rev.31 which validates adequacy of NCX 21 size cell. (The NCX-27 cells being about 30% larger than the NCX 21 cells
- 13 versus 10 positive plates - are enveloped by the adequacy of smaller, similar type and make NCX 21 cells). Parts B, D, F, and I of this calc. concerning the cable voltage drop and voltage profile may be incorporated in the current 125 Vdc Bus Voltage calculation 8982-66-19 l.
( Present Rev. 5 of 8982-66-19-1 does not include feeds from the alternate batteries.) D-8 ventw.saan Screening of Key Parameiert Asvmee 0
l l SAFETY RELATED 125 VDC l l 4 Parts B, C, and D of Calc. 8706-41 19 1, Rev.l( U3 - Alt.) are not current since they do not utilize the present load data utilized in Rev.31 of calc. 7056-00-19-5. Part C concerning sizing of the battery cells may be superseded by cross referencing it to the updated 125 Vdc battery sizing l calculation 7056-00-19-5, Rev.31 which verifies adequacy of NCX 21 size cell. Parts B and D l of this calc. concerning the cable voltage drop and voltage profile may be incorporated in the current ; 125 Vdc Bus Voltage calculation 8982-66-19-1, i ( Present Rev. 5 of 8982 66191 does not include feeds from the alternate batteries.) 1
- 5. Revision 31 of Calc. 7056-00-19 5 does not address or verify adequacy of the 125 V batteries for the SBO condition with the current load data utilized in this revision.
- 6. Calc. 5569-31-19-1, Rev.2, determines the maximum fault current level at specific locations / buses in the 125 Vdc system when fed from the main batteries only. It would be more appropriate to include the alternate batteries also within its scope to make it more representative of the entire 125 Vdc system.
References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Units 2 & 3 Techn 71 Specifications Upgrade (TSUP), January 13,1997.
- 3. Dresden Station Units 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis), January 13, 1997.
- 4. Dresden Administrative Technical Requirements (DATR), September 1996.
- 5. Dresden Technical Specifications, December 1996.
- 6. Dresden Design Basis Doc'! ment (DBD), DBD-DR 006, Revision A,"125 V and 250 V DC System",
dated 12/18/92.
- 7. IEEE Standards 450 and 485.
- 8. PCP Battery Charger Model 3 SD-130-200 Instruction Manual.
- 9. Dresden Procedures DOP 6900-06 and DOP 6900-07 for Units 2 and 3 respectively.
- 10. DC System Grounds Task Force Report, Rev.A, dated 5/25/1989, and related calculation 8256-14-19 1, Rev.0.
D-9 %entwatson Screemns of Key Param Ravenson 4
- . .- - - - .. ..- - - - - . ~ . - - - .- - .- - - - - -
l SAFETY RELATED 250 VDC System Descripti:m The 250 Vdc System provMes a re!iable source orde motive and control power for various plant systems and components in the station. The 250 Vdc System consists of batteries, battery chargers, cabling and distribution equipment that are l used to supply de to various systems / components throughout the station. The 250 Vdc system is designed so that each battery is sized to start and carry the normal de loads plus all de loads required for safe ! shutdown of one unit and operations required to limit the consequences of a design basis event on the other l unit for a period of 4 hours following the loss of ac power. j i l Kev System Components Comoonent EPN Descriotion 2(3)-83250 250 Vdc Battery 2/3-83250-2/3 250 Vdc Swing Battery Charger ' 2-83250-2(3-83250-3) 250 Vdc Battery Charger 2(3) 2-8302(3-8303) 250 Vdc Turbine Building Motor Control Center 2(3) 2-8302A(3-8303A) 250 Vdc Reactor Building Motor Control Center 2A(3A) 2-8302B(3 8303B) 250 Vdc Reactor Building Motor Control Center 2B(3B) Onerational Modes i l l Operating &tode l Initiation Not applicable; the de system is normally operable regardless of plant status. ! Function The 250 Vdc system provides de electrical power for normal de loads during plant operation and the required safe shutdown loads when necessary. Station Blackout Afode During a station blackout (SBO) on loss of all ac power in the plant, the batteries supply de power for safe shut down of the plant for I hour SBO coping until ac power is available from the alternate (AAC) diesel generator. Appendix R hiode l The Dresden Appendix R Fire Protection Program, under the topic of Supporting Associated Circuit Analysis - Coordinated Fault Protection Analysis, mentions that to prevent potential faults in branch circuits from affecting the 250 Vdc system, allloads that are not essential for safe shutdown will be tripped after the essentialloads are started. Thus, the Dresden App iix R Analysis primarily relies on the procedural tripping of a particular list of associated circuit:, for a fire in a particular fire zone, and it takes no credit for coordination of electrical protection. l \ D 10 Ventwm,oa Susaning o(Key Pumneters, Revu on 0
~. ,_.m_. . - ._ _.- _ . - _.._ _ _ . _ . _ _ _ . _ _ _. ._ _. _ _ . _ ._ . . . . . _ _ _ . _ . _
i I l l l l SAFETY RELATED 250 VDC l l- Key Parameters i Parameter Value Parameter Referente Calculation Reference i i Battery Voltages
)
- Normal (Float): 262.8 265.2 V DFL 95029 for Natt: The battery '
(2.19 - 2.21 V/ cell) Reference I float / equalize i Sectio : 8.3.2.1.1 voltages are based on the battery manufacturer's j recommendations. '
- Maximum (Equalize): 270.0 - 271.2 V Same as above. Same as above.
(2.25 - 2.26 V/ce11)
- Minimem: 210 V (1.75 V per cell) Ref.1, Sec. 8.3.2.1.1 Calc. PMED 8982-(minimum value applies at end of design load profile), 30-01, Rev ii (12/13/1996), U2/3
" Development of a Duty Cycle Based on a More Conservative l
Application of Coincident Starting Currents for the 250 Vdc System" N 3: Tne batteries have 60 cells each and they are sized for I.75 V per cell at the end of their design duty cycle. l l Load Profile / Duty Cycle 4 hour load profile Reference I, Sections PMED-8982-30-01, I per Dresden ELMS- 8.3.2.1.1 & 8.3.3 Rev. I1. DC data base / load files. Notc; Revision 11 of Calc. PMED-8982- ) 30-01 pertains to ISI question 3386-001 and it incorporates Comed response and commitments for this que.<G n. The calc. Revision l Summary (pages xi, I thru xiii) describes the incorporation of various inrush currents and their rationale. D-11 Venrm Sceaning of Key Pm llevman 0
1 l ' 4 4 i SAFETY RELATED 250 VDC I' Pararneter Value Parameter Reference Calculation Reference $l . 1 The 472 A first step inrush current 4. 4 for the HPCI Aux. Oil Pump (AOP) seems reasonable ! for a resistance start i de motor but its i basis is not clear. i (The AOP start test l- data showed that the first step inrush l current went beyond [. the 316 A scale of 1 the test current I recorder.) The statement l regarding opening of-Valve 23013 (with which valve 2301-14 is interlocked)in f more than 0.2 seconds to its inter-mediate open could I be referenced to the valve data sheet. 4 0.2 seconds is realistic; the stroke time in the valve data sheet will simpty justify it. The conclusion shows that the battery capacity margins improve to 11.9% for U2 and 11.6% for U3
- from 2.3% - with a 5 second time delay in closing the HPCI valve 2301-15 thru a modification.
Battery Minimum Capacity 2 80% of Ref. 2, Sec. 4. 9.C.5 & 6. IEEE Std.485 manufacturer's rating Ref. 7 (IEEE 450) ( up to 85% battery service life) D-12
% errmance $agenes of Key Parameters, Rewmoe 0
SAFETY RELATED 250 VDC Parameter Value Parameter Referenca Calculation Reference Cell Electrolyte Specific 21.200 (pilot cell) Ref. 2, Table 4.9.C-1 Noic: Typically Gravity & 21.195(connected cells) Ref. 2, Table 4.9.C-1 these values are Temperature > 60 F(average for all Ref. 2, Sec. 4 9.C.2 .c based on the connected cells) battery manufact-Cell Terminal Connection 5 150 micro-ohms , or Ref. 2, Sec. 4. 9.C. 2.b urer's recommen-Resistance 5 20 % above the cell dations. baseline connection resistance. (See Ref.1, Sec.8.3.2.4) Battery Electrolyte Temperature
- Minimum: 65 F(U2); 70 F(U3) Reference 1, Sec.9.4.4.5(U3) PMED 8982 30-01 Ngig: The design basis minimum electrolyte temperature is 65 F for both U2 and U3 batteries.
Charger ac Supply Voltage : 2432 V Ref. 8, Nameplate data Cales. 9198-18-19-1
& -3,Rev.1, (12/31/96),
" Dresden Safety Related Continuous Load Running /
Starting Voltages" Charger Rated de Supply 200 A for Ref. 2, Sec. 4.9. C.3.d Ref. 8, Nameplate Current at least 4 hours. data. (Itis based on the equipment manufacturer's design calculations.) DC System Ground < 125 kohm (LevelI ground) Reference i Sec.8.3.2.1.1 Ref.10 Resistance Alarm > 40 kohm & < 125 kohm (Level 11) & Table 8.3-8 Set Points: 5 40 kohm(LevelIII) Noic: The de system ground resistances are monitored per Re f. 9. System Electrical Full coordination Engineering Report SL45v0, Protection does not exist for all 250 Vdc Design. Vol. 4, "Overcurrent Coordination safety related circuits. Protective Device Coordination Study, Dresden Station - Units 2 and 3", 07/31/1992. D-13 Verifiance $acining o(Key Pm Revruce 0
SAFETY RELATED 250 VDC Potential Discrepancies 1. Reference 1, Section 8.3.2.1.1 (pg. 8.3-20), states that the 8 hour battery rating of 1495 ampere-hours is based on the lowest (minimum) expected electrolyte temperature of 65 F but TSUP Section 4.9.C.2.c lists a minimum of 60 F average for all the connected cells in the battery. This implies that I some cells may have less than 60 F minimum electrolyte temperature. The battery capacity / capability is based on a minimum 65 F cell electrolyte temperature since the battery cells are sized for 65 F minimum electrolyte temperature.
- 2. The 250 V battery sizing calculation PMED 8982-30-01 does not address the station blackout (SBO) requirements. An earlier 250 V battary sizing calculation (No.7056-00-19-4, Rev.7, dated 5 91) addresses the 4 hr. SBO in its Rev. 3 and I hr SBO in Rev. 5. This calculation is not current and it does not reflect the current, updated load data utilized in Rev.ll of Calc. PMED 8982 30-01. Also ^
there is duplication of Scope / Purpose concerning sizing of the 250 V battery cells between these two calculations.
- 3. See notes I and 2 below.
References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Units 2 & 3 Technical Specifications Upgrade (TSUP), January 13,1997.
- 3. Dresden Station Units 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis), January 13, 1997.
- 4. Dresden Administrative Technical Requirements (DATR), September 1996.
- 5. Dresden Technical Specifications, December 1996.
- 6. Dresden Design Basis Document (DBD), DBD-DR-006, Revision A,"125 V and 250 V DC System",
dated 12/18/92.
- 7. IEEE Standards 450 and 485.
- 8. PCP Battery Charger Model 3 SD-260-200 Instruction Manual.
- 9. Dresden Procedures DOP 6900-04 and DOP 6900-05 for Units 2 and 3, respectively.
- 10. DC System Grounds Task Force Report, Rev.A, dated 5/25/1989, and related calculation 8256-14-19 1, Rev.0.
HQIES
- l. Reference 6 ( Table 9-7, pg 9 18) lists the number of cells in the 250 V batteries as 116 instead of 120 ( this is being corrected by the DBD Change Request No. 96-56 ).
- 2. Presently there is a discrepancy between the battery float and equalizing voltages listed in the Reference 1, Section 8.3.2.1, and the actual values in use at Dresden listed in the battery surveillance procedures ( DES 8300-07, -15, -16, -17 and -20 ). This is being addressed by DFL 95029 ( to revise the UFSAR to resolve this discrepancy).
D-14 % erifw:saan 5 :reen.ag of Key Parameters, flew, son 0 l l
i < l l I J LOW PRESSURE COOLANT INJECTION (LPCI) System Descriotion
- The LPCI system is designed to remove reactor core decay heat, as well as sensible heat, from the reactor and containment in the event of a Loss of Coolant Accident (LOCA). i i
The LPCI system consists of pumps, drywell and suppression chamber spray headers, valves, and l associated piping and controls. The system is required to inject sufficient makeup water for postulated loss ! of coolant accidents to reflood the vessel to the appropriate core height to provide adequate core cooling and is later required to maintain the level at two-thirds core height. The LPCI system may also be used for l containment cooling by circulating suppression pool water through the containment cooling heat exchanger l ? and returning it to the torus. Cooling water is provided to the heat exchanger by the CCSW system. In l addition, the LPCI system may be used for drywell and suppression chamber spray to cool hot I noncondensable gases and condense water vapor in the containment. Key System Components l Comoonent EPN Descrintion 2(3)-1502 A LPCI A Pump 2(3)-1502-B LPCI B Pump 2(3)-1502-C LPCI C Pump 2(3)-1502 D LPCI D Pump 2(3)-1501-1 I A LPCI Heat Exchanger A Bypass Valve 2(3) 1501 1IB r Exchanger B Bypass Valve LPCI'leat 2(3)-1501-13A LPCI Pump A and B Minimum Flow Bypass Valve 2(3)-1501 138 LPCI Pump C and D Minimum Flow Bypass Valve 2(3) 1501-18A LPCI Loop A Torus Spray Outboard Isolation Valve 2(3) 1501 18B LPCI Loop B Torus Spray Outboard Isolation Valve 2(3)-1501-19A LPCI Loop A Torus Spray Inboard Isolation Valve 2(3)-1501 19B LPCI Loop B Torus Spray Inboard Isolation Valve 2(3Fl501-20A LPCI Loop A Inboard Full Flow Bypass Valve 2(3)-1501 208 LPCI Loop B Inboard Full Flow Bypass Valve 2(3)-150121 A LPCI Loop A Outboard Injection Valve 2(3)-1501-218 LPCI Loop B Outboard Injection Valve 2(3)-150122A LPCI Loop A Inboard Injection Valve 2(3)-150122B LPCI Loop B Inboard Injection Valve 2(3)-1501-25A LPCI Loop A Inboard injection Check Valve 2(3F1501258 LPCI Loop B Inboard Injection Check Valve 2(3)-150127A LPCI Loop A Drywell Spray Outboard Isolation Valve 2(3F150127B LPCI Loop B Drywell Spray Outboard Isolation Valve 2(3Fl50128A LPCI Loop A Drywell Spray Inboard Isolation Valve 2(3Fl50128B LPCI Loop B Drywell Spray Inboard Isolation Valve 2(3Fl50138A LPCI Loop A Outboard Full Flow Bypass Valve 2(3F150138B LPCI Loop B Outboard Full Flow Bypass Valve 2(3)-0202-05A RECIRC Loop A Discharge Valve 2(3F0202-05B RECIRC Loop B Discharge Valve 2(3)-150192A Flow Indicating Switch (Bypass Valve Closure) 2(3F1501-928 Flow Indicating Switch (Bypass Valve Closure) 2(3F3 5700 30A LPC1/CS Pump Room Cooler Fans 2(3F3-5700 30B LPCl/CS Pump Room Cooler Fans D-15 v.nn so orm., e.r aii o
i i l LOW PRESSURL COOLANT INJECTION (LPCI) j Comoonent EPN Descriotion 2(3)-1503 A LPCl/CCSW A Heat Exchanger l 2(3)-1503B LPCI/CCSW B Heat Exchanger 4 2(3) 1501-62A Contain Spray Interlock-High Pressure 2(3F1501-62B Contain Spray Interlock High Pressure l 2(3)-1501-62C Contain Spray Interlock-High Pressure [" 2(3F1501-62D Contain Spray Interlock High Pressure 2(3)-0263149 A Containment Spray Interlock 2/3 Core Height 2(3F0263-149 B Containment Spray Interlock 2/3 Core Height 2(3)-0263-111 A Loop Break Detection Logic Pressure Switch 2(3)-026311I B Loop Break Detection Logic Pressure Switch 2(3)-0263 111 C Loop Break Detection Logic Pressure Switch 2(3)-0263111 D Loop Break Detection Logic Pressure Switch 2(3)-026134 A Loop Break Detection Logic Dif. Pressure Switch 2(3)-026134 B Loop Break Detection Logic Dif. Pressure Switch 2(3)-0261-34 C Loop Break Detection Logic Dif. Pressure Switch 2(3)-026134 D Loop Break Detection Logic Dif. Pressure Switch 2(3)-026135 A Loop Break Detection Logic Dif. Pressure Switch 2(3)-0261-35 B Loop Break Detection Logic Dif. Pressure Switch 2(3)-0261-35 C Loop Break Detection Logic Dif. Pressure Switch 2(3)-026135 D Loop Break Detection Logic Dif. Pressure Switch 2(3)-026135 E Loop Break Detection Logic Dif. Pressure Switch 2(3)-026135 F Loop Break Detection Logic Dif. Pressure Switch 2(3)-026135 G Loop Break Detection Logic Dif. Pressure Switch , 2(3)-026135 H Loop Break Detection Logic Dif. Pressure Switch ! 2(3)-1530-182 Two Second Timer l Operational Modes low Pressure Coolant injection Mode Initiation (automatic) $ low-low reactor water level and reactor low pressure
- high drywell pressure low low reactor water level for 8.5 minutes (from ADS) ,
- initiated manually I Function l
The LPCI mode of the LPCI system restores and maintains the reactor core coolant inventory in 4 the reactor vessel following reactor depressurization after a LOCA for all design basis break sizes. ; I Containment Spray Mode l Initiation ! initiated manually when drywell pressure is greater than the permissive pressure Function The Containment Spray mode of the LPCI system cools hot noncondensable gases in the drywell and pressure suppression chamber, as well as condenses steam in the drywell, to reduce pressure ; and ternperature in these areas. This function is accomplished in conjunction with the , Containment Cooling Service Water (CCSW) system. ! 4 l
')
D-16 l Venficanon scronung of Key Parm lleviewe 0 I I J
s i LOW PRESSURE COOLANT INJECTION (LPCI) , s i Suppression PoolCooling Afode j Initiation ; initiated manually when drywell pressure is greater than the permissive pressure j Function 1 The Suppression Pool Cooling mode of the LPCI system removes reactor core decay heat and { sensible heat from the suppression pool during a LOCA and is one method of reaching hot shut l and /or cold shut down in the Appendix R response. This function is accomplished in conjunction I with the Containment Cooling Service Water (CCSW) system. ! i Station Blackout blode ; No additionalrequirements l l Appendit R Afode l In addition to the modes listed above the LPCI system, in conjunction with the Containment Cooling I Service Water (CCSW) system, provides a reactor make up method when pressure is below 350 psig and a reactor cool down method when the reactor is at low pressure and low decay power. ] i i ColdShutdown using Safety Grade Systems Afode The CCSW system and pressure relief system are used in conjunction with the LPCI system to provide a means of achieving cold shut down using safety grade systems. See UFSAR Section 6.3.1.2. i Kev Parameters { 4 Parameter Value Parameter Calculation Reference Reference f' Maximum Permissible 160* F 26 NFS-BS A-97-002 1/9/1997 "Dresden 2/3 Suppression Pool Temp. Containment Performance Under Reduced Suppression Pool and Service Water Temperature DRE 96-0214," Minimum Available CCSW Flow To Maintain 20 psi Differential Between LPCI and CCSW in the LPCI Heat Exchanger." Note: A licensing amendment was submitted to the NRC on Feb 17,1997 which will raise the maximum temperature to 176* F. Initial Suppression Pool 75* F 26 No calculation is necessary. This Maximum Tempera.ure is a boundary condition on the system. Note: A licensing amendment was submitted to the NRC on Feb 17,1997 which will raise the maximum temperature to 95* F. D-!7 v.a - s - .n.,c., n,a. .
_ __ _ . _. _. _ - _ - .__ _ - _ . . _ _ . . . - _ . _ _.._m._ . . _ ._ ___. . _ _._ . _ _ _ ._. - i j-1 LOW PRESSURE COOLANT INJECTION (LPCI) ] Low Pressure Coolant Inlection Mode q- Parameter Value Parameter Calculation Reference Egfstsact BlIng ! (2(3)-1502 A,B,C,D) l j Start of Flow 275 psid* 12 DRE 96-0211 Rev i 12/17/96 Unit 2 l "LPCI system derivation of system resistance 4 (*psid is the pressure difference curves, pump curves, and comparison to LOCA between the reactor vessel and analysis." I the containment) t i Low Flow Bypass Sufficient 21 Note: Reference 21 explains that based on a j technical evaluation it was decided that a specific j low flow requirement would not be adopted and l instead a periodic observation strategy would be used to determine if the pumps are deteriorating in use with the existing low flow bypass of l approximately 500 gpm. as given in Ref. 21. j Low Flow Bypass ! Opening Setpoint 21000gpm 2 NED I-EIC 124 Rev.0 9/21/95 Units 2/3 l 2(3)FIS 1501-92 A,B (Table "LPCI Pump Discharge Header Flow Setpoint 3.2.B-1) Loop Accuracy Calculation"- Note: See Potential Discrepancy item 1. ! Flow (Two Pumps) 9,000 gpm 12 DRE 96 -0211 Rev i 12/17/96 Unit 2 at 20 psid* "LPCI system derivation of system resistance curves 4 pump curves, and comparison to LOCA analysis." j (* psid is the preswre difference j between the reactcr vessel and j the containment) Note: See Potential Discrepancy items 2 and 3. Net Positive Suction 36 feet @ 11 DRE-97-0021/3/97 Rev. 01/9/97 Units 2/3 l Head Required $600 "Dresden LPCI/ Core Spray NPSH Analysis Post-l gpm/ pump DBA LOCA: GE SIL 151 Case Short Term" l ) Note: This calculation shows that, with an initial
- pool temperature of 75* F and the containment at atmospheric pressure, cavitation will occur but the
- resulting reduction in flow is acceptably small. A licensing amendment was submitted to the
! NRC on Feb 17,1997 which will raise the $ maximium temperature to 95* F. 1 j Note: See Potential Discrepancy item 4. a 4 i a ,' D-!8 Venficanon $avahat of Key Paramees Revees e J
- . - . __ . . _ . . - -_ .___.._ _ .. - .. _ _..___.- ___ __.__ _____._ _ _m __.m__ _ . . . _ . _
i ? i i LOW PRESSURE COOLANT INJECTION (LPCI) Eatsmeter Value Parameter Calculation Reference Reference l LPCI Inboard Inicetion Valve (2(3)-150122A,B) ] Opening Duration 27 seconds 12 MOV DATA SHEET #-3-150122A 6n/95 Min Operator Terminal 24 004-MN-334, Rev.4,8D/93 Voltage " Thrust Windows Calculation for Functional Group , Unit 2 Valve A 381 Volts LPCI 9," i Unit 2 Valve B 382 Volts 1 Unit 3 Valve A 402 Volts 4 Unit 3 Valve B 394 Volts Recire Dischares Valve 1 (2(3)-0202-05A,B) , Closure Duration 40 seconds 12 MOV DATA SHEET #-3-0202 5A I/6/97 I i
; Min. Operator Terminal 004 MN-309 Rev. 7,4/14/93 Voltage "7hrust Window s for Dresden Functional Group i
- Unit 2 Valve A 347 Volts 23 Recirc 2" i Unit 2 Valve B 356 Volts 23 Note
- There is a voltage mismatch for Unit 2 Valve A.
See Potential Discrepancy item 12. j Unit 3 Valve A 384 Volts 16 Unit 3 Valve B 390 Volts 16 ! Autoinitiation signals ] (See ECCS Initiation Key Parameters ! Table D-ll) i l Line Break Detection i ES < 900 psig 22, 1 ' NED-I EIC-0114 "LPCI Recire Loop Line Break (2(3)-0263-111 A,B,C,D) Sec. Detection" 1 7.3.1.2.2 Note: See Potential Discrepancy item 10 DflS 1 psid 22, 1 NED-I-EIC-Oll5 "LPCI Recirculation Loop Break (2(3)-026134 A,B,C,D) Sec. Detection" 7.3.1.2.2 J DEIS 2 psid 22,1 NED-I EIC-0118 "LPCI Recirculation Pump A&B (2(3)-0261 35A,B,C,D E,F,G,H) Sec. Diff. Pressure" 7.3.1.2.2
- Two Second Timer (2(3) 1530-182) 2sec. I DIS 1500 - 20 Rev. " LPCI Loop Selection
. Circuitry Figure Testing" 1.8 see to 2.2 sec. 1 7.35 l l r 5 D 19 ve s, a x.y e.,- st. .
_ . _ - - . ~ - . - . . - . -. . - _ ~ - .. .. - . .. - - . - .. . - - . - . . . ~ - . ~ . . - . . . - . J l j \ 4 LOW PRESSURE COOLANT INJECTION (LPCI) Containment Snrav Mogg - Parameter Value Parameter Calculation Reference Reference j Flow (One Pump) 5,000 gpm 13 DRE-97-0031/3/97 Rev.01/8/97 Units 2/3
- "Dresden LPC1/ Core Spray NPSH Analysis Post l DBA LOCA
- Reduced Torus Temperature Long -
1 Term." ( Note: This calculation shows that except for the i
- i. LPCIand 2 CCSW pu np case the NPSH !
l requirement of the pumps can be met if they are throttled back. The 1 LPCI and 2 CCSW results in ! cavitation and throttling is not an acceptable t solution because if the flow is reduced the j suppression pool temperature limit of 170* F will be exceeded, By reducing the maximum service water temperature to 75* ttom 95* F the maximum pool i temperature is reduced from 170* F sufficiently to provide adequate NPSH. See Potential Discrepancy item 5. f ' Flow to Suppression 0 to 250 gpm 13,15 No calculations were found. Chamber ', (Remainder of 5000 gpm to Drywell See Potential Discrepancy item 5 Spray) Heat Exchanger } ~ (2(3)-1503 A,B) (See CCSW Key Parameters l Table D 2a) i ! Pressure Permissive Switch 0.5spsl.5 2 NED-I EIC-0089 " Containment Spray Interlock- ! PS (2(3)-1501-62 A,B,C,D) psig Table High Pressure" l (Set Point) 3.2.11 2/3 Core Height Sorav 2-48 2 NED-I-EIC-0103 "2/3 Core Height Containment L Interlock Inches Table Spray Interlock" LIS(2(3)0263149 A,B,C,D) 3.2.11 j Sunnression Pool Cooline Mode } Flow (Two Pump) 10,700 gpm 13 DRE-97-0031/3/97 Rev.01/8/97 linits 2/3
, "Dresden LPCI/ Core Spray NPSH Analysis Post i DBA LOCA: Reduced Torus Temperature Long -
1 Term" Y
' D-20 Verifkanon Samunens of Key W Rewaon 0 a
, - +r ,---s, me = ,e, w- , - , r--- n v e --
1 ,--mv w
J l a : 1 l 4 LOW PRESSURE COOLANT INJECTION (LPCI)
- Parameter . Value Parameter Calculttion j Reference Reference i
Note: This calculation shows that this flow cannot 4 be met for NPSH reasons. The achievable flow is ] 10000 gpm after throttling from an NFSH stand
- point.
See Potential Discrepancy item 5 1 i Heat Exchanger
- (2(3)-1501-11 A,B)
(See CCSW Key Parameters Table D-2a ) f LPCI Pumo Motor i l Rathig 800 HP 11 Note: Reference (12) shows that the pump was -
- tested with an 800 HP motor driving it. l l Terminal Voltage i 10 % 19 Calculations have not been found. See Potential
[ Limits ofRated Discrepancy item 6. i
- Maximum Pumo Room s !W F 1 RSA-D-92-07 Dec.141992
- Temocrature (Fig. "LPCI Room Temperature Response due to Loss of j 3.111) Room Coolers at Dresden Station" Note
- It is shown that the rcom temperature will not
- exceed 185' F.
i [ Valves LPCI Heat Exchanger Bvoass 2(3)-1501-11 A,B - Min Operator Terminal 24 004 MN-338,Rev 1,11/11/93 . Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 405 Volts LPCI 3," l Unit 2 Valve B 405 Volts Unit 3 Valve A 417 Volts Unit 3 Valve B 416 Volts a f' LPCI Min. Flow Bvnan 2(3)-1501-13 A,B Min Operator Terminal 24 004-MN 320, Rev 5,2/4/94 Voltage " Thrust Windows Calculation for Functional Group ~ Unit 2 Valve A 421 Volts LPCI 4," i Unit 2 Valve B 420 Volts Note There is a voltage mismatch for Unit 2 valve Unit 3 Valve A 430 Volts B. See Potential Discrepancy item 7.
- Unit 3 Valve B 431 Volts e
4 I i l D-21 Vanfkauon $essenes of Key Paramesert Rewmeon 0
. ,m,- - , -v -- #-- ..v ,, -%--,c- - - . _ , - , - , , r y- + - - - ..
l i l LOW PRESSURE COOLANT INJECTION (LPCI) 1 Parameter Yalg Parameter Calculation Reference Reference l LPCI Torus Sorav Outboard 2(3F1501 18A,B Min Operator Terminal 24 004-MN-337. Rev 1,5/17/92 Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 419 Volts LPCI5" Unit 2 Valve B 417 Volts Unit 3 Va've A 427 Volts Unit 3 Valve B 430 Volts l l LPCITorus Sprav Outbo rd 2(3F1501 19A,B Min Operator Terminal 24 004-MN-323, Rev 3,2/11/94 ; Voltage " Thrust Windows Calculation for Functional Group ' Unit 2 Valve A . 409 Volts LPCI 6," l Unit 2 Valve B 407 Volts Note: There is a voltage mismatch for unit 2 valve l Unit 3 Valse A 418 Volts B. See Potential Discrepancy item 7. l Unit 3 Valve B 421 Volts LPCI Inboard Full Flow Byoass 2(3F1501-20A,B Min Operator Terminal 24 004-MN-3 t2, Rev 5,4/14/94 Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 3' - olts LPC! 7," Unit 2 Valve B 3; Volts Unit 3 Valve A 388 Volts Unit 3 Valvo B 394 Volts LPCI Outboard Iniection 2(3)15012 I A,B 4 Min Operator Termino! 004-MN 333, Rev 1,3/25/92 l Voltage " Thrust Windows Calculation for Functional Group l Unit 2 Valve A 416 Volts 24 LPCI 8," Unit 2 Valve B 417 Volts 24 Unit 3 Valve A 424 Volts 25 Note: there may be a voltage mis match, on Unit 3 Unit 3 Valve B 425 Volts 25 valves A &B LPCI Drvwell Sprav Outboard 2(3F1501-27A,B Min Operteor Terminal 24 004-MN-331, Rev 5,6/21/94 Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 410 Volts LPCI 10," Unit 2 Valve B 413 Volts Note: There is a voltage mismatch for unit 3 valve Unit 3 Valve A 407 Volts B. See Potential Discrepancy item 7. Unit 3 Valve B 393 Volts D-22 l v.nt- was or n., :. ,n m o
]
LOW PRESSURE COOLANT INJECTION (LPCI) Parameter Value Parameter Calculation Reference Reference LPCI Drvwell Sorav Inboard 2(3)-150128A,B Min Operator Terminal 24 004-MN-331 Rev 5,6/21/94 ' Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 409 Volts LPCI 10," Unit 2 Valve B 391 Volts Unit 3 Valve A 400 Volts Unit 3 Valve B 390 Volts Thrust Adequacy LPCI Full Flow Bvnass Outboard 2(3)-150138A,B Min Operator Terminal 24 004-MN 336, Rev 1,3/9/92 Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 392 Volts LPCI 12," Unit 2 Valve B 392 Volts Unit 3 Valve A 404 Volts Unit 3 Valve B 412 Volts FANS LPCI/CS Pumn Room Cooler Fans 2 5746A Minimum Terminal Voltage 9198-18-19-1. Rev. I i
" Running /Staiting Voltages" l Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable.
2 5746B Minimum Terminal Voltage 9198 18 19-2, Rev. I
" Running / Starting Voltages" Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable.
3 5746A Minimum Tenninal Voltage 919818-19 3, Rev. I
" Running / Starting Voltages" Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable.
D-23 Verwasami Smassag of Key Pmmers, Rew,snm 0
LOW PRESSURE COOLANT INJECTION (LPCI) Parameter Value Parameter Calculation Reference Reference 3-5746B Minimum Terminal Voltage 919818-19-4, Rev. I
" Running / Starting Ve!tages" Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable.
Potential Discrenancies
- 1. Calculation NED-I-EIC-0124 performed a setpoint error analysis for the LPCI minimum flow bypass flow switch for the valve opening requirement of > I,000 gpm. This calculation should be revised to determine the flow ranges where the flow switch would close the minimum flow bypass valve. This information would be used for calculation DRE96-0211, which calculated LPCI flow to the reactor vs.
reactor pressure and compared that to what is required under the LOCA analysis. Calculation DRE96-0211 did not include the minimum iow r bypass in the calculation. Calculation DRE96-0211 should be revised to include the minimum flow bypas; for the flow ranges specified in calculation NED-I-EIC-0124.
- 2. Dresden calculation DRE96-021I was performed to compare LPCI system resistance and pump curves te the LOCA analysis for Unit 2. At'achment P, Figure I shows the predicted 2 pump flow into the reactor vessel as a function of reactor vessel pressure and compares it to the flow used as input in the LOCA analysis. There is an error in this curve in that the minimum flow bypass was not accounted for in this curve.
- 3. Dresden calculation DRE96-0211 demonstrates that the LPCI system can deliver flow to the reactor vesselin excess of the flow required by the Appendix K LOCA analysis. However, this calculation was only prepared for Unit 2. A similar calculation needs to be completed for Unit 3.
- 4. The required Net Positive Suction Head (NPSH) for the LPCI pumps is identified in the LPCI DBD DBD-DR-172A as 31 feet (for I pump operation - 5,000 gpm) and 37.5 feet (2 pump operation 5,600 gpm). However, the LPCI pump curves ideo.ify the required NPSH to the pumps as 30 feet and 36 feet for 5,000 gpm and 5,600 gpm, respectively.
- 5. An evaluation performed by Nuclear Fuel Services (NFS), NFS:BSA: 96-140, specifies that 5,000 gpm flow is required by the LPCI system through the containment cooling heat exchanger t.ad to the containment to ensure adequate containment cooling. However, no formal hydraulic calculations exist that demonstrate the LPCI system can provide this flow to ensure adequate cooling in the suppression pool cooling or containment spray mode.
- 6. No calculations could be fod that demonstrate there is sufficient motor terminal voltage for the LFCI pumps so the pumps can perform adequately.
D-24 vw-o. so m or w-- n. n~ o
LOW PRESSURE COOLANT INJECTION (LPCI)
- 7. The available motor terminal voltage for 1501 series motor operated valves (MOVs) was calculated l under calculation NED EIC MOV DR-0001. For three valves, the thrust calculations used a motor terminal voltage value greater than the motor terminal voltage available under calculation NED-EIC-MOV-DR-0001. These valves are 2 1501 13B (thrust calculation 004-MN-320),2 1501 19B {for proposed modification scenario} (thrust calculation 004 MN-323), and 3 150127B (thrust calculation 004-MN 331).
- 8. Section 6.2.1.3.3 of the UFMR addresses the containment response after a LOCA, and analyzes four 1 containment cooling cases. Howcar, no calculations exist that provide the basis for these scenarios.
This section of the UFSAR should b' revised to address those cases which are important from a licensing perspective (i.e., I LPCI pump' 2 CCSW pumps) and perform the necessary calculations to provide the basis for these scenarios.
- 9. The LPCI system flow requirement for injection to the reactor vessel that is used in the LOCA analysis is 9,000 gpm at 20 psid. However, section 4.5 of the TSUP and Dresden Operating Surveillance (DOS) 1500-05 test for LPCI flow of 14,500 gpm. The TSUP and DOS 1500-05 should be evaluated for revision to test LPCI flow of 9,000 gpm as required by the LOCA analysis.
- 10. The calculation, NED-I EIC Oll4 " Reactor Vessel Pressure Switch ( LPCI Recirculation Loop Line Break Detection Logic) Setpoint Error Analysis." specifies a switch closure at 944 psig. The UFSAR in section 7.3.1.2.2 specifies the set point to be less than or equal to 900 psig.
- 11. There are several portions of section 6 of the UFSAR which are out of date. Specifically:
The available NPSH values given in Tables 6.3-5, Table 6.3 17, and Table 6.3-18 are not consistent l with those given in DRE-97-0021/3/97 Rev. 01/9/97 Units 2/3 "Dresden LPCI/ Core Spray NPSH Analysis Post DBA LOCA: GE SIL 151 Case Short-Term," This calculation is part of the basis of the recent license amendment. The LPCI system flow requirement for injection to the reactor vessel that is used in the LOCA analysis is 9,000 gpm at 20 psid. See: EMF 89-065 Rev. 3 July 1995 "Dresden Units 2 and 3 Principal LOCA Analysis Parameters." However, Table 6.3-1 and Section 6.3.3.1.2 of the USFAR give other values such as 14500 gpm at 20 psid for three pumps. Figure 6.3 1 of the UFSAR is out of date. It describes the operation of the ECCS over a range ofline breaks and infers that the LPCI and Core Spray are independent of each other for large breaks. This is not correct for the current Appendix K methodology.
- 12. The thrust calculations (004-MN-309 Rev. 7,4/14/93," Thrust Windows for Dresden Functional Group Recirc 2") for the Recire Discharge Valve (2-0202-005 A) used a terminal voltage which is greater than the voltage calculated in DRE 96-0010, Rev. O,1/16/96, Motor Terminal Voltage Cat:. for Dresden Unit 2.
13 In procedures DIS 150013 Rev. 9, and DIS 150014 Rev. 6, the note at the bottom of the Data Sheets for the master trip unit pages incorrectly lists the setpoint milliamp value for i 155 gpm as 9.097 mA.
'!he correct value should be 9.907 mA as shown in the trip column of the data sheet, and as shown in calculation NED-1 EIC-0124, Rev.1. This is for low-flow bypass, which is part of LPCI.
1 D-25 Wnrwauus Saeoning of Key Pa,% Remon 0
~._. - _ . _ _ . . _ . . _ _ . _ . _ _ _ . .. . _ _ . . -_ . . LOW PRESSURE COOLANT INJ ECTION (LPCI) Parameter References i
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR) i
- 2. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP)
- 3. Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis) January 13,1997
- 4. Dresden Administrative Technical Requirements (DATR) September
- 5. Dresden Technical Specifications December,1996
- 6. Memo.: J. W. Dingler to R. J. Goebbert 02 19-93,"20 psid Differential Pressure Serpoint Between l CCSW and LPCI Across LPCI HX " Ref 06.028 to Design Basis Document 172.
- 7. VV-11 Rev 0 Feb/8/92," Determine CCSW cooler cooling coil's new capacity using test data." l
- 8. D. H. Legler to C. W. Schroeder Nov 9,1992 "Dresden - Unit 2 Finalized Operability Determination of the CCSW Pumps" Ref 06.055 to Design Basis Document 172:
- 9. "CCSW pump curve - NPSH" Ref 03.014 to Design Basis Document 172:
- 10. K. W. Hess to D. P. Galle Aug. 13,1974 " Containment Cooling Water Pumps (CCSW) Technical Specification Change" Ref 06.004 to Design Basis Document 172:
- 11. LPCI Pump Curve by Bingham Pump Co. Jan 10,1968
- 12. EMF-89-065 Rev. 3 July 1995 "Dresden Units 2 and 3 Principal LOCA Analysis Parameters"
- 13. 729E583 Rev i 1968 " Process Diagram LPCI Containment Cooling System"
- 14. 257HA654 Rev. 3 April 15,1969 " Auxiliary System Data Book"
- 15. Memo to DBD DRF by W. G. Myers et al March 4,1992 "BWR 2,3,4, & 5 RHR Containment Spray Cooling (CSC) Requirements" Ref 06.100 to Design Basis Document 172"
- 16. NED EIC-MOV DR-0003 Rev. O Sept. I,94 "MOV Terminal Voltage Catculation"
- 17. "MOV Terminal Voltage Calculation"C. N. Mathewson to R. J. Ascheral Feb. 25,1971
" Containment Cooling Service Water Pumps Motor Ratings Ref 06.008 to Design Basis Docuecent 172:
- 18. 21 AS450 Rev 1, Nov. 5,1965,"Soccification for Containment Cooling Motor"
- 19. 2 l A5580 Rev 4, Oct. 31 1972," Motor General Requirements"
- 20. Containment Cooling Heat Exchanger Specification Sheet. March 291967 By Berlin Chapman Inc.
- 21. CHRON # 0306316 K. Simmons to R.L.Bax et al March 24,1995 " Orifice In Min Flow Lines /
LPCI and Core Spray Systems" D-26 4 andcuuou $creameg 6f Key Parameters, item.oe 0
LOW PRESSURE COOLANT INJECTION (LPCI)
- 22. 257HA350AN1, Rev.10, i 1/9/71 " Nuclear Boiler System - Data Sheets. l
- 23. DRE96-0010 Rev. 0,1/16/96,"blotor Terminal Voltage Calc. for Dresden Unit 2 MOV's 2-0202 5 A/B" 24 NED EIC MOV-DR-0001, Rev 0 8/2/94," Valve Actuator Motor Voltage Calculation for Dresden 1501 System Units 2&3" l
- 25. DRE96-0127, Rev 0, 7/8/96," Motor Terminal Voltage Calculation for Dresden ;
Unit 3 MOVs 3-1501-21 A/B"
- 26. Amendment No.152 to Facility Operating License No. DPR-19 and Amendment No.147 to Facility Operating License No. DPR-25. Jan 1997 1
- 27. S. Mintz to S. L. Eldridge / B. M. Viehl, "Dresden LPCl/ Containment Cooling System - Comparison of Heat Exchanger Heat Transfer Rates." Dec. 28,1992 falculation References
- 1. NFS-BSA-96-140, Nov. 4,1996, ' Unit 2/3An Evaluation of Reduced LPCI Heat Exchanger Performance at Dresden 2/3, Reduced CCSW Flow"
- 2. DRE 96 -0211 Rev i 12/17/96," Unit 2 LPCI system derivation of system resistance curves, pump !
curves, and comparison to LOCA analysis."
- 3. NED-I EIC-124, Rev. O,9/21/95," Units 2/3 LPCI Pump Discharge Header Flow Setpoint Loop Accuracy Calculation"
- 4. DRE 97-002,1/3/97, Rev. O,1/9/97," Units 2/3 Dresden LPCl/ Core Spray NPSH Analysis Post-DBA LOCA: GE SIL 151 Case Short-Term"
- 5. MOV DATA SHEET #-3 150122A 6/7/95
- 6. 004 MN-309,Rev.7,4/14/93 " Thrust Windows Calculation for Functional Group RECIRC-2." i
- 7. MOV DATA SHEET #-3-0202-5A l/6/97
- 8. DRE96-0010 Rev.01/16/96 " Motor Terminal Voltage Calc. for Dresden Unit 2 MOV's 2-0202-SA/B"
- 9. DRE-97-0031/3/97 Rev.01/8/97 Units 2/3 "Dresden LPCl/ Core Spray NPSH Analysis Post DBA LOCA: Reduced Torus Temperature Long Term."
- 10. RSA D-92-07, Dec.141992,"LPCI Room Temperature Response due to Loss of Room Coolers at Dresden Station"
- 11. 9198 18-19 1, Rev.1, " Running / Starting Voltages"
- 12. NED-I EIC-Oll4 "LPCI Recire Loop Line Break Detection"
- 13. NED-I EIC-Oll5 "LPCI Recirculation Loop Break Detection" D-27 Ventkassesi Saesaiapf Key Pm Reisen 0
l LOW PRESSURE COOLANT INJECTION (LPCI)
- 14. NED-I EIC-Oll8"LPCI Recirculation Pump A&B Diff. Pressure"
- 15. NED I EIC-0089 " Containment Spray Interlock High Pressure" l i
- 16. NED I E1C-0103 "2/3 Core Heignt Containment Spray Interlock" 1
t i i 1 D-28 v.iiwoo. ser r .t m., e. s a. o
1 e i i CONTAINMENT COOLING SERVICE WATER (CCSW) i j System Descrintion i The CCSW system provides cooling water to tube side efine LPCl/CCSW Heat Exchangers to remove i heat from the primary containment. The function of cooling the water in the suppression pool, thereby limiting the temperature of the suppression pool water, assures the following: i
- cooling of the Containment.
no radioactive release through the containment heat exchanger i The CCSW system is an open loop system consisting of pumps, heat exchangers, valves, and associated piping and controls. The pumps take suction from the crib house bay and circulate water through the containment heat exchanger to cool suppression pool water. The water is then discharged to the Service Water main header. The CCSW system circulates water through room coolers to maintain the temperature l inside a watertight vault that contains two of the four CCSW pumps for flood protection.The CCSW } system also provides a safety related source of cooling water to the control room air conditioning i condensers. Kev System Comnonents - Comoonent EPN Descriotion 2(3Fl501-44A CCSW A Pump i 2(3F1501-44B CCSW B Pump 2(3Fl501-44C CCSW C Pump i 2(3Fl501-44D CCSW D Pump j 2(3FI503A LPCl/CCSW A Heat Exchanger j 2(3Fl503B LPCl/CCSW B H at Exchanger 2(3F5700-30A CCSW Vault Room A Coil 2(3F5700 308 CCSW Vault Room B Coil 2(3F5700-30C CCSW Vault Room C Coil i 2(3F5700 30D CCSW Vault Room D Coil 2(3F1543 A Pressure Differential Sensor 2(3F1543 B Pressure Differential Sensor , 2(3F15013 A CCSW Flow Control Valve 2(3F15013B CCSW Flow Control Valve Onerational Modes Containment Spray Mode Initiation Manually initiated j Function , CCSW provides cooling water to the tube side of heat exchangers 2(3F1503A,B to cool the
- water in the suppression pool. Suppression pool water is circulated through the shell side of the heat exchangers by the Low Pressure Coolant injection (LPCI) system to spray nozzles in the i containment. Minimum number of pumps in operation
- 2 CCSW and I LPCI.
buppression Pool Cooling Mode No unique requirements except; pumps in operation are: 2 CCSW and 2 LPCI pumps and LPCI flow is retumed to the suppression pool. I i D-29 Venounce Saeemag a(Key Parm Reve.on 0
l a J f CONTAINMENT COOLING SERVICE WATER (CCSW) , i i ' Station Blackout Afode No additional requirements l Appendix R olode No additional requirements { i Dam Failure Afode Similar to modes above except USFAR section 9.2.5.3.2 implies only 1 CCSW pump is required. This is contrary to Section 9.2.1.3 which calls for a minimum of 2 CCSW pumps. See Potential l Discrepancy item 5. l ColdShutdown using Safety Grade Systems Atode The CCSW system and pressure relief system are used in conjunction with the LPCI system to provide a means of achieving cold shut down using safety grade systems. See UFSAR Section 6.3.1.2. Kev Parameters Parameter Value Parameter Calculation Reference Esference Maximum service inlet water temperature 75'F 26 No calc. is required. This is a boundary condition on the system. Note: A licensing amendment was submitted to the NRC on Feb 17,1997 which will raise the maximium temperature to 95* F. 4 Minimum Water Source 500 FT Mean 2 No calc. is required. This is a boundary condition on . Elevation Sea Level Sec. the system (3.8 C) i Note: See Potential Discrepancy item 1. l Containment Sorav Mode l 1 i Pump j 2(3)-150144 A,B,C,D j l Flow (Two Pumps) 5600 gpm 26 DRE96-0214Rev.011/12/96 Unit 2/3 j
" Minimum Available CCSW flow !
to maintain a 20 psi differential between I LPCI and CCSW in CCSW Heat Exchanger" Note: A licensing amendment was submitted to the { NRC on Feb 17,1997 which will lower the flow rate to 5000 gpm. 1 Net Positive Suction 18 t1 9 No calculation was found. I Head Required @ 3,500 gpm/ pump Note: See Potential Discrepancy item 2. 1 l 1 1 D-30 Venftsanos Sewa,ag o(Key Parameeses Lewmoe 0
CONTAINMENT COOLING SERVICE WATER (CCSW) Parameter yajug Parameter Calculation Reference Reference CCSW Pump Motor Rating 500 HP 17 No calculation was found. Note See Potential Discrepancy item 3. Terminal Voltage 4000 Volts 18 No calculation was found. Note: See Potential Discrepancy item 3. Sunnression Pool Coolino Mode Flow (Two Pumps) 5600 gpm 26 Same as Containment Spray Mode Flow shown above. The hydraulic conditions are the same. Heat Exchanner i 2(3)-1501-44 A,B,C,D l l Heat Transfer Duty 98.6 MBtu/hr 27 S. Mintz to S. L. Eldridge et al" Dresden l
@ Tube Side Intet 95*F 27 LPCl/ Containment Cooling System - Comparison of l
@ Tube Side Flow 7,000 gpm 27 Heat Exchanger Heat Transfer Rates." 12/28/92 1
@Shell Side Temp 165* F 27
@Shell Side Inlet 10,700 gpm 27 Note: These parameters and the physical data characterize the heat exchanger perfonnance and are used to determine the heat transferred under actual operating conditions. For the l design basis calculations the LPCI flow is
$000 gpm and the CCSW flow is $600 gpm.
Vault Room Maximum Temperature 120* F See VV 13 Rev 0 5/7/93 Units 2/3 Note "CCSW vault cooler performance and effectiveness" Note: The calculation VV 13 shows that the max. vault temperature will be well below 120* F. This then became the de facto temperature limit. Vault Room Coil 2(3)-5700-30A,B,C,D Coils / Cooler 2 i No cale. is required. This is a statement of fact. Sec.9.2 (Fig. 9.2. I, 9.2-2) D-31 VeVmaene $crosew$ of Key Parameters Reision 0
l l ! CONTAINMENT COOLING SERVICE WATER (CCSW) Parameter Value Parameter Calculation ! Reference Reference t ) Total Flow For Two 110 gpm 7 ATD 0253 Rev.0 Units 2&3 htar 5,1993 j Coils " Determination of Flow Restricting Orifices for l CCSW Pump Room Coolers and CR Refrigeration j . Condenser in the CCSW System" i ) i Note: Flow does not pass through flow meter }
- Eans J
{ 2 5700 30 A,B hiin. Terminal Voltage 91981819-1 Rev. I
" Starting and Running Voltages" Note: This calculation determines the tenninal voltages under starting and run conditions and shows that these voltages are acceptable. l l
1 l 2 5700-30 C,D hiin. Terminal Voltage 919818-19 2 Rev. I
" Starting and Running Voltages" Note: This calculation determiaes the terminal voltages under starting and run conditions and shows that these voltages are acceptable. l 3 5700 30 A,B hiin.Tenninal Voltage 91981819 3 Rev. I l
" Starting and Running Voltages" Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable.
3 57M-30 C,D D-32 v.ne s, aac. , a- o
._ _ . . . . ~ _ _ _ _ _ . - . _ _ _ _ . _ _ _ _ _ . _ _ . _ - _ . _ _ _ _ _ . _ _ _ _ _ _ .
i l ' ) CONTAINMENT COOLING SERVICE WATER (CCSW) I } Parameter Value Parameter Calculation I i Refereneq Reference
- l Min. Terminal Voltage 91981819-4 Rev. t
" Starting and Running Voltages
)
a j Note: This calculation determines the terminal voltages under starting and run conditions and shows that these voltages are acceptable. Except as noted below.
- Note
- The running terminal voltage was found to be j 89 % instead of 90 %. To reconcile this difference 1
- calculation 9198 18 19 6 Rev. O was prepared to j justify accepting the lower voltage i
! Control Room HVAC - 102 gpm 8 ATD-0253 Rev.0 Units 2&3 Mar 5,1993 i Refrieeration Condensing (Unit 2 only) " Determination of Flow Restricting Orifices for , Unit (RCU) Flow CCSW Pump Room Coolers and CR Refrigeration
- . Condenser in the CCSW System" l
4 Note: This calculation shows that the room cooler and CR Refrigeration orifices were sized for 55 gpm , per cooler and 102 gpm respectively.
- Note
- Flow does not go through CCSW flow meter.
- it can be measured by a meter in the CR. HVAC System. However, during surveillance testing there j is no HVAC condenser flow. Therefore, to have a j valid test, the flow required must include the actual HVAC condenser flow. This is reported to be 121 gpm. l l
3 CCSW/LPCI Differential 20 psi 6 DRE96-0214Rev.011/12/97 Unit 2/3 l Pressure Sensor i " Minimum Available CCSW flow i 2(3) 1543 A,B (Sec. 6.2.2) to maintain a 20 psi differential between LPCI and CCSW in CCSW Heat Exchanger" [, Note: Calculation DRE 96 214 demonstrates that for j the i LPCI /2 CCSW case the CCSW pressure differential can be maintained 20 psi above LPCI for a range of containment pressures. These calculations should be extended to cover the 2 LPCI
- / 2 CCSW case it can be shown that the 20 psi can be maintained for these cases but quantitatively the flow rates are not known, and hence, the effectiveness of the containment heat exchanger is
; not known. All of which means that the long term
- temperature and pressure of the containment cannot
, be predicted. The analyzed case cos ers the limiting j containment cooling case from a licensing
; standpoint. However, the other cases are shown in D-33
- s.nna so . .r m., P.,- n o 4
l CONTAINMENT COOLING SERVICE WATER (CCSW) l Parameter Value Parameter Calculation Reference Reference section 6.2 of the USFAR as illustrations of capability and should be corrected. ; See Potential Discrepancy item 8 of LPCI Key Parameters-(Table 2a). The reason why a 20 psi differential is adequate can not be found in a calculation. A calculation has been performed by C. B. Johnson and F. J. Mollerus that shows the 20 psi is adequate. This cale should be reviewed and adopted by Comed. See Potential Discrepancy item 4. CCSW Flow Control Valve 2(3)-1501 3 A,B Min. Operator Terminal 24 004-MN 344, Rev 8,5/6/93 Voltage " Thrust Windows Calculation for Functional Group Unit 2 Valve A 425 Volts LPCI 1," Unit 2 Valve B 424 Volts Unit 3 Valve A 431 Volts Unit 3 Valve B 434 Volts Potential Discrepancin
- 1. Section 3.19.2 of the Dresden Technical Administrative Requirements (DATR) identifies the minimum water level in cribhouse CCSW suction bay of 499' - 0". However, Section 3.8.C.I of the TSUP identifies the minimum water level as 500' 0". Furthermore, the UFS AR references drawing M 10 which states the normal operating low water level in the suction bay is 501' - 0", and this value was used in calculation .
DRE96-0214. l
- 2. No formal calculations exist that demonstrate adequate Net Positive Suction Head (NPSH) is available to the CCSW pumps. l
- 3. No e,'culations could be found that demonstrate there is sufficient motor terminal voltage for the CCSW pumps so the pumps cari perform adequately, in addition, no calculations have been found that show the CCSW pump motu,3 are matched for power and speed requirements to the CCSW pumps.
- 4. There is a requirement that the CCSW side of the containment cooling heat exchanger be maintained at 20 psi above the LPCI side of the heat exchanger to prec!ude leakage from LPCI to CCSW. A report was generated by F. J. Mollerus and C. B. Johnson that shows why maintaining this pressure differential is adequate. This report should be formalized into a calculation.
- 5. Section 9.2.1.3 of the UFSAR specifies the minimum requirements for containment cooling include 2 CCSW pumps. However, section 9.2.5.3.2 indicate only 1 CCSW pump will be available during a postulated dam failure. Currently there are no containment cooling or heat exchanger differential pressure calculations for a I CCSW pump scenario.
D-34 Venrasee laeoning at Key Parm Ra,wce 0
I I CONTAINMENT COOLING SERVICE WATER (CCSW) Parameter References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR)
- 2. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP)
- 3. Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis) January 13,1997
- 4. Dresden Administrative Technical Requirements (DATR) September
- 5. Dresden Technical Specifications December,1996 i
- 6. Memo.: 1. W. Dingler to R. J. Goebbert,02 19-93,"20 psid Differential Pressure Setpoint Between l CCSW and LPCI Across LPCI HX, Ref 06.028 to Design Basis Document 172: l
- 7. VV 11 Rev. O, Feb/8/92," Determine CCSW cooler cooling coil's new capacity using test data."
- 8. D. H. Lagler to C. W. Schroeder, Nov 9,1992,"Dresden - Unit 2 Finalized Operability Determination of the CCSW Pumps," Ref. 06.055 to Design Basis Document 172:
- 9. "CCSW pump curve - NPSH," Ref. 03.014 to Design Basis Document 172:
- 10. K. W. Hess to D. P. Galle Aug. 13,1974," Containment Cooling Water Pumps,(CCSW) Technical Specification Change" Ref 06 '4 to Design Basis Document 172:
- 11. LPCI Pump Curve by Bingham b..sp Co., Jan 10,1968,
- 12. EMF 89-065, Rev. 3, July 1995,"Dresden Units 2 and 3 Principal LOCA Analysis Parameters."
13, 729E583, Rev.1,1968," Process Diagram LPCI Containment Cooling System."
- 14. 257HA654, Rev. 3, April 15,1969," Auxiliary System Data Book."
- 15. Memo to DBD DRF by W. G. Myers et al, March 4,1992,"BWR 2,3,4, & 5 RHR-Containment Spray Cooling (CSC) Requirements" Ref 06.100 to Design Basis Document 172."
- 16. NED-EIC-MOV DR 0003, Rev. O, Sept. I,94,"MOV Terminal Voltage Calculation."
- 17. "MOV Terminal Voltage Calculation "C. N. Mathewson to R. J. Ascheral, Feb. 25,1971
" Containment Cooling Service Water Pumps Motor Ratings Ref 06.008 to Design Basis Document 172; 1 - 18. Memo of Data Transmittal GE to S &L,4/8/68, S&L P.O. NO. 3447 134.
- 19. 21 A5580, Rev. 4, Oct. 31,1972," Motor General Requirements."
- 20. Containment Cooling Heat Exchanger Specification Sheet. March 29,1967, By Berlin Chapman Inc.
- 21. CHRON # 0306316 K. Simmons to R.L.Bax et al March 24,1995," Orifice in Min Flow Lines /
LPCI and Core Spray Systems." D 35 Vereceaan $cros***6 a(Key Parenweers As. mon 0
- _, . - . . . . - - . .._ . ~ - _ - _ - . . ~ . _ _ . .- - ,
1 1 CONTAINMENT COOLING SERVICE WATER (CCSW) 22 257HA350AM, Rev.10, i1/9/71," Nuclear Boiler System Data Sheets.
- 23. DRE96-0010, Rev. O,1/16/96," Motor Terminal Voltage Calc. for Dresden Unit 2 MOV's 2-0202 5A/B."
24 NED-EIC MOV-DR 0001, Rev 0 8/2/94," Valve Actuator Motor Voltage Calculation for Dresden 1501 System Units 2&3".
- 25. DRE96-0127, Rev. O, 7/8/96," Motor Terminal Voltage Calculation for Dresden l Unit 3 MOVs 3 15012 t A/B". I
- 26. Amendment No.152 to Facility Operating License Na. DPR 19 and Amendment No.147 to Facility :i Operating License No. DPR 25, Jan 1997 27 S. Mintz to S. L. Eldridge / B. M. Viehl, "Dresden LPCl/ Containment Cooling System - Comparison of Haet Exchanger Heat Transfer Rates," Dec. 28,1992. !
Calcubisc3 References
- l. UFSAR Log # DFL-96140 t/13/97
- 2. DRE96-0214, Rev. 0, i1/12/97," Unit 2/3 Minimum Available CCSW flow to maintain a 20 psi differential between LPCI and CCSW in CCSW Heat Exchanger".
- 3. NED M MSD-45, Rev. O,12/31/92," Units 2/3 Dresden Unit 2 LPCI Heat Exchanger Mode C Heat Exchanger Duty Calculation" 1
- 4. VV 13, Rev. O,5/7/93," Units 2/3 CCSW vault cooler performance and effectiveness". ]
- 5. ATD-0253, Rev. O, Mar 5,1993," Units 2&3 Determination of Flow Restricting Orifices for CCSW j Pump Room Coolers and CR Refrigeration Condenser in the CCSW System",
- 6. 9198 18-19 1, Rev.1, " Starting and Running Voltages". l l
l i l l l D-36 Vanficaea $creen**S e(Key Parww.sn, Iten,,a,0
l i FEEDWATER/ CONDENSATE i j System Description j j The Feedwater/ Condensate System (FWC) provides demineralized vater from the condenser to the reactor
- equivalent to the rate at which water is being generated to steam in the reactor pressure vessel. To
{' accomplish this function, the condensate in the main condenser hotwell must be pumped through various FWC system components before being returned to the reactor vessel. During power opecations, these components provide the condensate with necessary pressure boost, preheating and impurity removal. At
- this stage, the condensate is referred to as feedwater. The feedwater must be regulated (controlled) to l match the reactor steam generation rate and maintain sufficient water level in the reactor.
~ During normal power operations the FWC system operates in automatic control maintaining feedwater i flow equivalent to reactor steam generation rate and, thereby maintaining reactor water level. During non-l power or reactor trip operations, the same controls are adjusted to provide feedw ater flow control to match j decay heat steam generation and msintain reactor water level. 1 Key System Comnonents The following system components are important to the operation of the feedwater/ condensate system with respect to the eflicient generation of electricity. Since the overall process is monitored and adjusted by the plant computer system during normal operation, system performance verification by manual calculations is not required. If a component cannot perform at the purchase specification or expected level, the power generated will be reduced as necessary to balance the heat cycle. Generation of electricity at full rated capacity verities that the system has ban des: ped correctly. Comoonent EPN Descriotion l 2(3)-0220-58-A Reactor Feedwater injection inboard A Check CIV " l 2(3F0220-58.B Reactor Feedwater Injection Inboard B Check CIV " l 2(3)-0220 62 A Reactor Feedwater injection Outboard A Check CIV " 2(3F0220-62 B Reactor Feedwater injection Outboard B Check CIV " 2(3)-0642-A Feedwater A Main Flow Control Regulator 2(3F0642 B Feedwater B Main Flow Control Regulator 2(3F0643 Feedwater Low Flow Control Regulator 2(3F3101 Al Low Pressure Feedwater Heater A1 2(3)-3101 A2 Low Pressure Feedwater Hester A2 2(3)3101 A3 Low Pressure Feedwater Heater A3 2(3F3102.Al Low Pressure Feedwater Heater Al Drain Cooler 2(3)-3102 A2 Low Pressure Feedwater Heater A2 Drain Cooler 2(3h3102 A3 Low Pressure Feedwater Heater A3 Drain Cooler 2(3) 3103 B1 Low Pressure Feedwater Heater Bt 2(3)-3103 B2 Low Pressure Feedwater Heater B2 2(3) 3103 B3 Low Pressure Feedwater Heater B3 2(3)-3104 Cl Low Pressure Feedwater Heater C1 2(3) 3104-C2 Low Pressure Feedwater Heater C2 2(3) 3104 C3 Low Pressure Feedwater Heater C3 2(3F3105 DI High Pressure Feedwater Heater DI 2(3)-3105 D2 High Pressure Feedwater Heater D2 2(3F3105 D3 High Pressure Feedwater Heater D3 2(3)-3201-A Reactor Feedwater A Pump 2(3)-3201 B Reactor Feedwater B Pump 2(3F3201 C Reactor Feedwater C Pump D-37 r.a,=i.so s.n.,e m . m a,,,,,,o
1 i i j FEEDWATERI CCNDENSATE a Comnonent EPN Descriotion j 2(3F3202 A Reactor Feedwater A Pu np Oil Cooler 4 2(3F3202 B Reactor Feedwater B Punp Oil Cooler j 2(3)-3202 C Reactor Feedwater C Pu.np Oil Cooler l 2(3F3202 Al Auxiliary Reactor Feedw ster Oil A Pump 2(3F3202 Bl Auxiliary Reactor Feedwater Oil B Pump 2(3F3202-Cl Auxiliary Reactor Feed. vater Oil C Pump I 2(3) 3302.A Condensate A Pump ! 2(3F3302 B Condensate B Pump 2(3)-3302-C Condensate C Pump i 2(3) 3302-D Condensate D Pump i 2(3F3401 A Condensate Booster A Pump ! 2(3) 3401 B Condensate Booster B Pump i 2(3)-3401 C Condensate Booster C .Nmp 2(3)-3401 D Condensate Booster D Pump I LITS 2(3F0263 59A Reactor High Water Lev si Reactor Feed Pump Trip Switch l. LITS 2(3F0263 59B Reactor High Water Lev:1 Reactor Feed Pump Trip Switch , " Containment Isolation Valves (CIV) 2(3)-0220-58 A/B and 2(3)-0220-62 A/B are in line check valves 3 that do not require auxiliary systems to function. i
- Onerational Modes
! ) Operating Afode Initiation j - operable during normal plant operation l l Function l l The Condensate /Feedwater system takes condensate from the main condenser and provides j feedwater to the reactor pressure vessel at the rate in which the water is converted to steam. The i Condensate /Feedwater system preheats the water through a series oflow pressure and high j pressure heaters to increase overall plant operating efficiency. Station Blackout Afode The Feedwater/ Condensate system will not be operable during a station blackout. l Appendix R Afode Depending on the location and magnitude of a fire, portions of the Feedwater/ Condensate system may be operab!e. However, no credit for Feedwater/ Condensate is taken in the Appendix R analysis. ) Transient Analysis Afodes ! 1. Decrease in Feedwater Temperature (UFSAR 15.1.1) l A decrease in feedwater temperature due to loss of feedwater heating would result in a core power i increase due to the increase in core intet subcooling and the reactivity effects of the corresponding increase in moderator density. { The coolant void fraction decrease and the negative void reactivity coeflicient would result in a gradual initial increase in reactor power. If power exceeded the normal 100% flow control line, the operator would be expected to insert control rods to return power and flow to their normal range. If i this were not done, the neutron flux could exceed the scram setpoint . However, a scram is assumed to
; not occur for this transient analysis.
i i D-38 l' Ventkance " ; of Key Parasamers Remaan a n
... - - . , _ - y.-. ....-..,,.m__. -
,m, _, ., , ., , ,_m_y.-,__. - , , , _ , _ ,..,, , -_,
l l FEEDWATER/ CONDENSATE
- 2. Increase in Feedwater Flow (UFSAR 15.1.2)
The increase in feedwater Gow event is postulated on the basis of a single failure of a control device, speciScally one which can directly cause an increase in coolant inventory by increasing feedwater Dow. The most severe applicable event is a feedwater controller failure during maximum flow demand. The feedwater controller is forced to its upper limit at the beginning of the event. In addition, the final feedwater temperature is redaced below the rated feedwater temperature. This excess flow results in an increase in core subcooling which results in a rise in core power and an increase in the reactor water level. The rise in the reactor vessel water level eventually leads to a high water level turbine trip and a feedwater pump trip. The reactor scram is initiated by closure of the turbine stop valves.
- 3. Loss of Normal Feedwater Flow (UFSAR 15.2.7 & 15.8.3)
A loss of feedwater transient is postulated to occur due to a feedwater controller malfvnetion demanding closure of the feedwater control valves. The unit response to tripping of:all feedwater pumps would be very similar to the transient evaluated. The reactor water level would decrease due to the mismatch between the steam flow out of the vessel and the shut off feedwater flow. A reactor low-low water level trip would initiate ECCS. LCCA Analysis Mode Since the feedwater/ condensate system is non-safety related, the system operability during a LOCA event cannot be guaranteed. Ilowever, an ( perator may use any system that is operable to inject water into the reactor during a LOCA. Availabilhy of the feedwater/ condensate system greatly increases the overall plant safety margins. Postulating that the feedwater/ condensate is operable following a LOCA event, the only key system parameter would be the ability to provide flow into the reactor without an explicit flow magnitude requirement. The availability of the Feedwater/ Condensate system for ECCS postulates no loss of auxiliary power (Reference 1, Table 6.3 2). Therefore, when available, the Feedwater Condensate system is assumed to function normally during a LOCA event. Reactor Feed Pump Trip on Reactor High Water Level The Nuclear Boiler Instrumentation provides a reactor high water level trip signal to the reactor feed pumps. This prevents reactor water from reaching the main steam lines which could result in equipment damage. Failure of the Feedwater control system re:ulting in increased How to the reactor vessel will result in a high vessel level trip of the main turbine and reactor feed pumps. Kev Parameters Normal Operation at Maximum Pown Parameter Value Parameter Referegg Calculation Reference Feedwater Flowrate 9,725,000 lbm/hr Ref 1, Sect 10.4.7.1 Reference 8 Feedwater Temperature 340.I'F Reference 8 Reference 8 D-39 Ver%%uon kreeaang of Key Partnem Remion 0
_ >_ _ _ . _ _ . _ . . _ _ _ _ _ _ _ _ . - _ . _ _ _ .. .m_._ . ... - . - . _ . . _ . . - - . _ _ _ _ _ - . l f i i FEEDWATER/ CONDENSATE
- Parameter Yalut Parameter Reference Calculation Reference j Flowrate Uncertainty 1
- 16 % Ref 1, Table 4.4 1 Calculation 19
- (Nominal) See Note I j Temperature Uncertainty 0.76 % Ref 1, Table 4.4 1 Calculation 20 l (Nominal) See Note 2 i
) Transient Analysis Innut
- Initial Conditions at Rated Power Level I
Parameter ,Value Parameter Reference Calculation Reference Feedwater Flowrate 9,800,000 lbm/hr Ref 14, Sect 2.4 See Discrepancy 1 , Feedwater Enthalpy 312.9 BTUilbm Ref 14, Sect 2.9 See Discrepancy I j i l , Steam Flow Versus FeedwaterTemocrature ' Parameter Blue Parameter Reference Calculation Reference l
- Steam Fraction vs Table of Ref 14, Sect 3.13 See Discrepancy I Feedwater Temperature Values
-{ Steam Fraction with ) Respect to Rated -I j Steam Flow I a 4 Parameter Measurement Uncertainties 4 ) Parameter Value Parameter Reference Calculation Reference Feedwater Flowrate 2.0 % Ref 14, Sect 5.1 Calculation 19 See Discrepancy I & Note i Feedwater Temperature 0.91 % (340*F) Ref 14, Sect 5.2 Calculation 20 Feedwater Temperature 4.47 % (340*F) Ref 14, Sect 5.2 See Discrepancy I & - w/ Stratification Error Note 2 Reactor Pressure 2.3% at 1000 psig Ref 14, Sect 5.3 Calculation 21 See Discrepancy I & Note 3 Transient: Maximum Feedwater Temnerature Loss Due to Single Failure Parameter Yah:e Parameter Reference Calculation Reference Temperature Loss 200*F Ref 1, Sect 15.1.1.2 See Discrepancy i Ref 14 Sect 8.4.1 D-40 Venficause smenwig 4 Key Pumess As= won 0
. - - - . . . ~ . . , ., ,. -
1 1 FEEDWATER/ CONDENSATE Parameter Value Parameter Reference Calculation Reference l Time Step for Loss Not Used Ref 14 Sect 8.4.1 See Discrepancy 1 Transient: Reactor Feed Pumn Runout Flow 1 Parameter Value Parameter Reference Calculation Reference Max runout flow 3311 lbm/sec Ref 14 Sect 8.3.3 See Discrepancy 1 Temperature Loss 100*F Ref 1, Sect 15.1.2.2.E See Discrepancy 1 Reg Valve Stroke Time 12 see Ref 14 Sect 8.3.4 Calculation 8 Ruetor Feed Pumn Trin l Parameter Value Parameter Reference Calculation Reference 1 LITS 263 59A, B $ 201 inches Ref 2 Table 3.2.1 1 Calculation 22 Trip Setpoint (s $8" RWL) l Potential Discrepancies I. A formal calculation providing a basis for feedwater parameters used in the transient analyses could l I not be located. This discrepancy is not considered an operability issue because the data used is reasonable. During normal operation, the plant computer monitors feedwater flow and temperature in I order to balance the heat cycle for eflicient generation of electricity. I For the Loss of Feedwater Heating Transient (LFHT), UFSAR section 15.1.1.2 states Seimens assumed feedwater temperature dropped over a relatively long period. UFSAR Section 15.1.1.3 indicates the LFHT is longer than 100 seconds. However, Seimens analysis, Reference 14 Section < 8.4.2, indicates that a time step was not used. Reference 14 Comment 8 stated that LFHT is evaluated I with the generic LFHT methodology ANF-1358(P)(A). l Input to the transient analysis should be reviewed and if needed, a basis documented as part of the Adequacy and Retrievability of Design Basis Calculations project. Notes l
- 1. Per Calculation 19, measured feedwater flowrate uncertainty at full flow is approximately 2.5%. .
Data indicated in the UFSAR represents typical historical uncertainties not current analysis l uncertainties. No action is required.
- 2. Per Calculation 20, measured feedwater temperature uncertainty at 340*F is 3.I'F (t 0.91%) without temperature stratification and 8.0*F ( 2.35%) with temperature stratification. Data indicated in the UFSAR represents typical historical uncertainties not current analysis uncertainties. No action is required.
- 3. Per Calculation 21, measured reactor pressure uncertainty at 1200 psig is approximately t 18 psig (t 1.50%). This is consenative with respect to the transient analysis. No action is required.
D-4 I % e>ruma Sawasag of Key Parinnviers Remon 0
.-.-.-. _ - - - - . - . - - - - - ~ _
i 4 FEEDWATER/ CONDENSATE { 4. Documentation discrepancies between Reference i Section 10.4.7.2.1, Reference 15 Drawing M 14, and Reference 17 Drawing 12E 2417 Sheet I need to be reconciled and the documentation updated to I reflect the current fictd condition. Drawing M 14 indicates Feedwater Regulating Valve 2-642B has 4 been changed to have an air operated actuator not a hydraulic activated actuator as shown on drawing ! 12E 2417 Sheet I and as indicated in UFSAR Section 10.4.7.2,1. This discrepancy is a configuration management issue not an operability concern. l ! Unit 2 vs. Unit 3 Differences ' 1 , Per UFSAR Section 10.4.7.2.1 (Reference 1), Feedwater Regulating Valves (FWRV) 2-642B,3-642A and j 3-642B have hydraulic operated actuators while valve 2-642A has an air operated actuator. Low flow i regulating valve 2 643 is rated at 7% of system full flow capacity while valve 3-643 is rated at 22% of ! system full flow capacity, t l Reference 15 indicates valve 2-6428 has an air operated actuator Reference 17 indicates FWRV 2-6428 l has a hydraulic operated actuator. ! The differences between units have no effect on key system parameters. References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision Ol A.
- 2. Dresden Station Units 2 & 3 Technical Specification Upgrade (TSUP), January 13,1997.
- 3. Dresden Station Units 2 & 3 Technical Specification Upgrade Basis (TSUP Basis), January 13,1997.
- 4. Dresden Administrative Technical Requirements (DATR), September 1995
- 5. Dresden Technical Specitications, December 1996.
- 7. Dresden NPS Units 2 and 3, System Notebook for Feedwater and Condensate System, Revision 0, February 1993.
- 8. GF. Heat Balance No.1083, Preliminary Plant Heat Balance, Dresden 11 Uprated N.B. Stretch, Rev 0, Ul7/66.
- 9. GE Specification 257HA461 Rev 1, Feedwater Control System Design Specification.
- 10. S&L Specification K 215512/11/65, Condensate Pumps and Condensate Booster Pumps.
- 11. S&L Specification K 2156 08/02/65, Reactor Feed Pumps.
- 12. S&L Specification K 2157 08/27/65, Closed Feedwater Heaters.
- 13. Dresden Design Basis Document (DBD), DBD-DR-086 Revision A,"Feedwater/ Condensate System", dated 03/07/96.
D-42 v.*u.,= so .< x., en mu e
FEEDWATER/ CONDENSATE Parameter Value brameter Reference Calculation Reference
- 14. Siemens EMF-95-043 dated June 1996, Dresden Unit 3 Cycle 15 Principal Transient Analysis Parameters.
- 15. Drawing M 14 Revision LA, Diagram of Reactor Feed Piping, Dresden Unit 2.
- 16. Drawing M 347 Revision AY, Diagram of Reactor Feed Piping, Dresden Unit 3.
- 17. Drawing 12E 2417 Sheet 1 Revision W. Block Diagram Feedwater Control, Dresden Unit 2.
- 18. Drawing 12E-3417 Sheet i Revision Y, Block Diagram Feedwater Control, Dresden Unit 3.
1 Desien Calculations Identined i l
- 1. MAD 89-004 Rev 0, RFP Operability Assessment.
]
- 2. MAD 87-0816 Rev 0, Feedwater System Analysis
- 3. DE-398 C-001 Rev 0, Feedwater Regulating Valve Control Panel Mod 4
- 4. DR-292-C-002 Rev 0, Misalignment of Condensate / Condensate Booster Pump
- 5. DR-347-M-001 Rev 0, Condensate Load Calc for Replacement 2-230123 Valve
- 6. DR 353 M-002 Rev 1, Feedwater Reg Valves Backup Bottle Sizing
- 7. DR-353-M-00lRev 0, Feedwater Reg Fire Protection
- 8. DR-3531-001100 Rev 0, E/J Feedwater Valve Setpoints PS-2 656A, PS 2-656B & PS-2-662
- 9. DR-353 P-001 Rev 0, Feedwater Piping Reanalysis with Valves 642A & B
- 10. DR-353 P-002 Rev 0, Service Air to Feedwater Reg Valve
- 11. 033951(EMD) Rev 1, Certified Design Report Feedwater Condensate Piping System
- 12. 049993(EMD) Rev 0, Feedwater Condensate Piping System
- 13. 062956(EMD) Rev 0, Certified Design Report Feedwater Condensate Piping System
- 14. 067191(EMD) Rev 0, Feedwater Piping Design Pressure
- 15. PMED-8982-62-02 Rev 0, Condensate Transfer Pump Suction Piping Pressure Loss and NPSH
- 16. DRE96-0150 Rev 0, Flow Characterization Calculation for Dresden Unit 2 Feedwater Regulating Valves
- 17. NSLD-LIC-91-5 Rev 0, LPCI and Feedwater Systems Free Volumes and Water Masses D-43 Venfar.ance $memag of Key Pwaniews Ww o
FEEDWATER/ CONDENSATE
- 18. NSLD LIC 3 Rev 0, Reactor Feedwater & Core Spray Free Volumes & Water Masses i
- 19. NED-I EIC-0300 Revision 0, Feedwater Flow Error in the Process Computer.
- 20. NED-I EIC 0301 Revision 0, Feedwater Temperature Error Indice' ion at the Process Computer.
l
- 21. NED-I EIC 0295 Revision I, Reactor Pressure Error Indication at Process Computer.
)
- 22. NED-I EIC-0268 Revision 3, Main Turbine Trip, Reactor Feedwater Pump Trip and Feedwater Pump Runout Reset on High Reactor I.evel Setpoint Error Analysis.
l l 1 l l s.naao %% a x., tv- w= o
l TURBINE BUILDING CLOSED COOLING WATER (TBCCW) l System Description l The TBCCW system is a closed loop system that provides cooling water to various systems and components that are located in the Turbine Building and Crib House. The TBCCW system consists of pumps, heat exchangers, expansion tank, and chemical feeder that is separate and independent per unit. Pumps and heat exchangers circulate the cooling water throughout the i unit. The expansion tank ensures there is adequate NPSH for the pumps, and provides control of the ' amount of demineralized water in the system due to water expansion with overtlow considerations The chemical feeder allows for induction of a corrosion inhibitor at the pump suction, Key System Components Comeonert EPN Dgscription 2(3)-3801 A TBCCW A Pump 2(3)-3801-B TBCCW B Pump 2(3)-3802-A TBCCW A Heat Exchanger 2(3)-3802-B TBCCW B Heat Exchanger 2(3)-3803 TBCCW Expansion Tank 2(3)-3804 TBCCW Chemical Addition Pot 2(3))-LS 384122B TBCCW Tank LevelControl I Operational Modes Plant Operating Afode Initiation operable during normal plant operation ** Function ; The TBCCW system provides clean demineralized water for cooling purposes of the following system / components:
- Instrument AirCompressors
- Service AirCompressors Radwaste Sparging Air Compressors (Unit 2 only)
Resin Transfer Air Compressor
- Electro-Hydraulic Control (EHC) Oil Cooler Circulating Water Pump Thrust Bearing Oil Cooler (Normally Unit 2 only)
- Condensate and Booster Pump Seal Coolers
- Control Rod Drive Pump Seal Coolers Reactor Feed Pump Lube Oil and Mechanical Seal Coolers Bus Duct Coolers Main Generator Exciter Air Cooler Sample Coolers
- Decontamination Facility Air Conditioner (Unit 2 only)
- Considered to be TBCCW support functions for risk significant systems.
Station Blackout Afode Not available during Station Blackout D-45 Venflasm seseneng of Key Pwm Re ace 0
l i 1
\
\ \
- TURBINE BUILDING CLOSED COOLING WATER (TBCC%)
Appendix R Afode No additional requirements i Kev Parameters } Parameter Value Parameter Reference Calculation Reference 4 1 TBCCW Pump Motor b , Minimum Voltage TBD i TBCCW Heat Exchanger { 1
- Number (per unit) 2 Reference 6,7 1
l Heat Transfer per Heat 12.5x10' BTU /hr Reference 6,7 j Exchanger 4 . Flowrate per Heat Shell- 1000 gpm Reference 6,7 i j Exchanger Tube - 2500 gpm Reference 6,7 1
, Operating Inlet Shel! - 130*F Reference 6,7 j Temperature Tube - 95'F Reference 6,7 Operating Outlet Shell 105'F Reference 6,7 Temperature Tube - 105*F Reference 6,7 Flowrate Instrument Air Adequate Compressor Condensate and Booster 4@l0 gpm Reference 7 Pump Seal Coolers Control Rod Drive Adequate Pump Seal Coolers Bus Duct Coolers 2@ 96gpm Reference 7 Reactor Feed Pump
' Lube Oil and Mechanical Seal Coolert 3@ $9 gpm Reference 7 Potential Discrenancies No formal calculations could be found that provide a basis for the flowrate through the TBCCW system.
No formal calculations could be found that provide a basis for heat transfer rate o' the TBCCW heat exchangers. D 46
% ennsuJe '- ; ef Key Parsneters. Ammoe 0
- . - . . - . .-. . . . . . ~ . -
l I i TURBINE BUILDING CLOSED COOLING WATER (TBCCW) Refersscal
- 1. Dresden Rebaseline Updated Einal Safety Analysis Report (UFSAR), Revision Ol A.
- 2. Dresden Station Unit 2 & 3 Technical Specification Upgrade (TSUP), January 13,1997
- 3. Dresden Station Unit 2 & 3 Technical Specification Upgrade Basis (TSUP Basis), Januaryl3,1997 1
- 4. Dresden Administrative Technical Requirements (DATR), September 1996 '
1 l S. Dresden Technical Specifications, December 1996 l
- 6. TBCCW System Notebook (IPE) l I
- 7. Equipment Manual, Cha:;;:: iv, Closed Cooling Water Systems, GEK-786, June 1972 l l
I l l i l I l I l l l l l l I i-I D-47 Venficasan $creeneng of Key Pm Reveme 0
. - . - . . - . -. - _- . - .. - - - _ ~ . - . _ ~ .
1 o MAIN STEAM SAFETY AND RELIEF VALVES l I l System Descrintion ' The Main Steam Safety and Relief Valves protect the reactor from overpressurization, and rapidly depressurize the reactor to allow the Low Pressure Coolant injection (LPCI) and Core Spray systemt to function. He nuclear system pressure relicf system includes eight safety valves, four relief valves, and one safety-reliefvalve. All of these valves are located on the main steam lines within the drywell. l l The safety valves are self-actuating, spring loaded valves that discharge directly into the drywell. I The four electromatic relief valves are opened by energizing their respective 125 VDC solenoids. One i Target Rock relief valve operates as a safety-relicf valve. Each of the five relief valves can be individually } opened manually by operation of keylock switches located in the main control room, or automatically as part of the Automatic Depressurization System (ADS). l Key System Comnonents Comoonent EPN Descriotion 2(3F0203-3A Target Rock Safety / Relief Valve 2(3)-0203 3B Electromatic Relief Velve (ERV) B l 2(3)-0203-3C Electromatic Relief Valve (ERV)C l 2(3F0203 3D Electromatic Relief Valve (ERV) D 2(3h0203-3E Electromatic Relief Valve (ERV) E 2(3)-0203-4A Reactor Safety Valve A 2(3)-0203-4B Reactor Safety Valve B l 2(3F0203-4C Reactor Safety Valve C 2(3F0203-4D Reactor Safety Valve D j 2(3)-0203-4E Reactor Safety Valve E l 2(3) 0203-4F Reactor Safety Valve F l 2(3)-0203-4G Reactor Safety Valve G 2(3)-0203-4H Reactor Safety Valve H i 2(3)-0203 106A Rupture Disc on Reactor Safety Valve A j 2(3)-0203 106B Rupture Disc on Reactor Safety Valve B ! 2(3F0203106C Rupture Disc on Reactor Safety Valve C l 2(3)-0203 106D Rupture Disc on Reactor Safety Valve D 2(3)-0203106E Rupture Disc on Reactor Safety Valve E 2(3)-0203 106F Rupture Disc on Reactor Safety Valve F 2(3)-0203-106G Rupture Disc on Reactor Safety Valve G 2(3)-0203-106H Rupture Disc on Reactor Safety Valve H l PC-2(3F0203-3A Hl/LO Pressure Controller Pressure Switch for Relief Valve A PC-2(3F0203-3B Hl/LO Pressure Controller Pressure Switch for Relief Valve B PC-2(3)-0203-3C Hl/LO Pressure Controller Pressure Switch for Relief Valve C PC-2(3F0203-3D Hl/LO Pressure Controller Pressure Switch for Relicf Valve D PC-2(3>0203-3E Hl/LO Pressure Controller Pressure Switch for Relief Valve E 4 D-48 Venfkenen Screensag of Key Perunees, llevumm 0 ( 1-- - --
l l l l l l l MAIN STEAM SAFETY AND RELIEF VALVES I l Operational Modes l Operating Afode - Target Rock Sapty/ Relief Valve l Initiation l relief valve setting - see serpoint below l safety valve setting - see setpoint below l
- initiated manually Function The Target Rock Safety / Relief Valve depressurizes the reactor and discharges the steam to the suppression pool below the minimum torus water level.
Operating Afode - Electromatic Relief Valves Initiation two valves - see setpoint below two vah es - see setpoint below
- initiated manually Function The ERVs depressurize the retctor and discharge the steam to the suppression pool below the minimum torus water level. These valves also rapidly relieve reactor pressure to allow operation of the LPCI and Core Spray systems.
Operating 5fode - Rcactor Safety Valves initiation two Valves - see setpoint below
- two valves - see setpoint below
- four valves - see serpoint below Function
- The Reactor Safety Valves depressurize the reactor and discharge the steam directly to the drywell atmosphere.
Operating blode - Automatic Depressuri:ation System (ADS) Initiation - Valves 203 3 A through -3E Reactor low-low water level (-59 inches) and high dryw ell pressure (s; 2 psig) after a two-minute time delay and indication that a Core Spray or LPCI pump is in operation with 2100 psig discharge pressure
- Reactor low-low water level continuous for 8.5 minutes and confumation that a LPCI or Core Spray pump is running Function Depressurize the reactor when HPCI system cannot maintain reactor vessel water level.
ADS mode evaluated as a separate critical system Station Blackout Afode No additional requirements Appendh R Afode No additional requirements D-49 Ven(m Scramme of Kev Pwamm Revnon 0
, . - - - -. -.. . . . -- - . ... . - . - _ . ~ .- . . _ , MALN STEAM SAFETY AND RELIEF VALVF3 Key Parameters Transient Analysis Parameter Value Parameter Reference Calculation Reference Relief valve open/close set points (capacity) Ref 9 & 10 Calculation not required I valve 1150.1/1104.1 psia See Note 4 (166.1 lbm/see at 1080 psig + 3%) 2 valves 1126.9/1081.2 psia (155.0 lbnt/sec at 1120 psig) 2 valves 1150.1/1104.1 psia (155.0 lbm/see at 1120 psig) Safety valve open set point (capacity) Ref 9 & 10 Calculation not required I valve 1184.2 psia See Note 4 i (159.46 lbm/sec at 1080 psig + 3%) 2 valves 1292.3 psia (171.84 lbm/sec at 1240 psig + 3%) Valves -4 A thru -4H 2 valves 1302.6 psia tested per Ref 16 l (171.84 lbm/sec at 1240 psig + 3%) 4 valves 1312.9 psia (171.84 lbra/sec at 1240 psig + 3%) LOCA Analysis Mia Capacity 2.16E06 lbrn/hr Reference !I None Required. 4 Relief valves at i135 psid Miscellaneous Reauirements Required Minimum Voltage 203-3 A through 3E Not Identified No Design Reference See Discrepancy i Relicf Valve Time Delay UFSAR Section 5.2.2.4 203 3B through 3C 8 see min Ref 1, Sect 5.2.2.4 indicates 8 see min time delay compares to 8.5 sec min Ref13 a max 6.3 sec elevated Timer Setpoint water leg time duration. Calculation not located. See discrepancy 2 Background Information Total Capacity 2.7E06 lbm/hr Ref 1, Table 5.2-1 No Calculation needed. 5 Relief Valves I I f D-50
% entkanon screeness of Key Parammes, Itavissa 0 l
l
l l l MAIN STEAM SAFETY AND RELIEF VALVES ; I Parameter Value Parameter Reference Calculation Reference l Total Capacity 5.2E06 lbm/hr Ref 1, Table 5.2 1 No Calculation needed. 8 Safety Valves (not Target Rock) l Relief Valve Pressure Switch Setnoints ' l 203 3A i124 psig Ref. 6, Section 2.0 See Calculation 3 203 3B,3C 1101 psig Ref. 6, Section 2.0 203 3D,3E I124 psig Ref. 6, Section 2.0 See Note 2 Ref I Sect 5.2.2.4 203-3A si135 psig Ref. 2, Section 3.6.F See Calculation 3 ! 2 Valves s1112 psig 2 Valves $1135 psig See Note 2 l ! 203 3A I124 psig Ref. 5, Section 4.6.E See Calculation 3 203 3B,3C 1101 psig 203 3D,3E I124 psig Setpoint Uncert il% (Ref. 5, Section 4.6.E) See Note 2 l Safety Valve Pressure Setnoints t l
.. 203-3 A i135 psig Ref 1 Table 5.2-1 See Note 2 203 4A through 4H 1240 psig Ref. 6, Section 2.0 I (2 valves) Ref i Table 5.2 1 Valvas -4A through -4H
! 1250 psig Sect 5.2.2.2.3 set per Ref 17 l (2 valves) ) l 1260 psig - i l (4 valves) [ .203 3A 1135 psig11% Ref 2 Sect 3.6.E See Note 2 , 403-4A through 4H 1240 psig tl% ' l (2 valves) Valves -4A through -4H I 1250 psig il% set per Ref17 l (2 valves) 1260 psig11% (4 valves)
.. 203 3A 1135 psig Ref 5 Sect 4.6.E See Note 2 203-4A through 4H 1240 psig
> (2 valves) Valves -4A through -4H l 1250 psig set per Ref17 (2 valves) 1260 psig ! (4 valves) Setpoint Uncertil% l (Ref 5 Sect 4.6.E) D-51 Ventwaeon $creemag of Key Parama,rs Revision 0
1 l l l MAIN STEAM SAFETY AND RELIEF VALVES Parameter Value Earameter Reference Calcubtion Reference Flow Rate Capacity 203 3A 622.000 lbm/hr Ref. 6, Section 2.0 No Calculation needed (at 1125 psig) Ref.1, Table 5.2-1 Vendor Certified. 203 3B through 3E 540,000 lbm/hr Ref. 6, Section 2.0 No Calculation needed l (at 1101 psig) Vendor Certitled. 1 203-4A through 4H 644,501 lbm/hr Ref. 6, Section 2.0 Calculation 2 derates (at 1240 psig) Ref I Table 5.2 1 Vendor Certified 649,638 lbm/hr valaes (at 1250 psig) 654,774 lbm/hr Certified values per j (at 1260 psig) Reference 14 l l Reseat Pressure Below Initiation Setpoint (non-ADS) l l 203 3A 50 psig hef6 See Note 1 203-3B through 32 70 psig 203-4A through 4H 100 psig 203-3A 46.0 psig Ref. 9 & 10 See Note 1 203-3B through 3E 45.7 psig l l Max Relief Valve 40% of set Reference 15 See Calculation i Backpressure Pressure ! Rupture Discs l l l 203106A through 106H Burst pressure 13.5 psig at 72*F Reference 12 EJ . No calc required l I1.9 psig at 150*F Small rupture pressure I will not affect set pressure for blowdown ' Potential Discrenancies I. No calculation could be found that determines the minimum voltage at the solenoid terminals, or if this voltage is adequate to operate the valve. A voltage drop calculation for both Units 2 & 3 needs to be performed for the 125 VDC sysMra. '/endor information pertaining to the minimum required voltage needs to be obtained and documented at part of the voltage assessment. These calculations should be part of the 2 year program. 1 D-52 Venre Saeoning o(Key Parameren. flevision 0
I SIAIN STEASI SAFETY AND RELIEF VALVES I l
- 2. The SRV valves have a nominal 10 second timer to inhibit reactivation to allow the reflood level in the SRV discharge line to drop to normal level. UFSAR Section 5.2.2.4 (Reference 1) indicates that an 8 second minimum time delay compares to a maximum 6.3 second calculated elevated water leg time duration in the discharge lines. The backup design calculation demonstrathg that this 8 second value is conservative for all SRV discharge lines needs to be either recovered or reconstituted during the 2
]
year program. l Notes
- 1. There are valve close serpoint differences between the System Notebook (Reference 6) and the Transient Analyses. The vslues indicated in the System Notebook are nominal values. No action is required . I
- 2. Valve setpoints and tolerances are consistent between the TSUPs, Technical Specifications and the UFSAR. The setpoints used in the Transient Analyses include conservative setpoint tolerances permitted by the ASME code (Reference 7) which will give conservative analysis results. No action is required.
I 1
- 3. !
Calculation 3 has a typographical error on pages 9 and 10, Tech Spec Lower LCO and Tech Spec Upper LCO data should not be t. There is no impact on the calculation results. No action is required.
- 4. A formal calculation for backup to the transient analysis input parameters including tolerances was not i identified. However, the inputs values could be duplicated within acceptable margins using the l comments in Refccences 9 & 10 and Reference 7 as guidance. Since the input data could be readily l duplicated, no action is required.' !
References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision Ol A.
- 2. Dresden Station Units 2 & 3 Technical Specification Upgrade (TSUP), January 13,1997. l
- 3. Dresden Station Units 2 & 3 Technical Specification Upgrade Basis (TSUP Basis), January 13,1997.
l
- 4. Dresden Administrative Technical Requirements (DATR), September 1996.
- 5. Dresden Technical Specifications, December 1996.
- 6. Dresden NPS Units 2 and 3 System Notebook for Reactor Vessel Pressure Control and Depressurization, Revision 0, February 1993.
- 7. ASME Boiler and Pressure Vessel Code,1995 Edition, NB 7000.
- 8. TLP Calculation Note on " Determination of Failure Probability of Relief Valves Failing to Open Failing to Close," DR-CN-92-004.
- 9. EMF-94-049 May 1995,"Dresden Unit 2 Cycle 15 Principal Transient Analysis Parameters."
- 10. EMF-95-043 June 1996,"Dresden Unit 3 Cycle 15 Principal Transient Analysis Parameters."
D-53
%erinuuon Screeneng of Key Parameters Revu,ca 0
. - - _ - - . - . - - . . . - .. . . . .~ - - - - - .
l l l 1 l i MAIN STEAM SAFE ~lY AND RELIEF VALVES
- 11. EMF 93-176 Revision 2 September 1996, Updated Principal LOCA Analysis Parameters for Dresden l Units 2 and 3 " ~
1
- 12. Rupture Disc purchase order
- 13. DIS 0250-05 Revision 15, Automatic Depressurization System Auto-Actuation Test. I
- 14. GE Specification 21 A5743 Revision 1, Reactor Vessel Safety Valves,11/8/66.
l 15. GE Specification 257HA350AM Revision 10, Ntclear Boiler System - Data Sheets. ! R
- 16. DMP 0200-03 Revision 11, Main Steam Safety Valve Pre-Maintenance Test. l
- 17. DMP 0200-30 Revision 16, Reactor Main Steam Safety Valve Repair and Post Maintenance Testing.
l 4 Calculations Reviewed:
- 1. EMD-043057 Revision 0, SRV Discharge Steady State Back Pressure.
i l 2. NED-P-MSD l Revision 0, Reduction in Safety Relief Valve Capacity at Dresden and Quad Cities 1 Stations. i
- 3. NED-I EIC-00093 Revision 2, Electromatic Relief Valve /rarget Rock Valve Pressure Switch Setpoint _,
Error Analysis. l l l l 1 j 1 l I ! l I l 1 i I l l ! D-54 I v.ne so %.rnoe - now e I
l l l SERVICE WATER (SW) 1 1
System Description
1 The Service Water system is an open loop system that provides strained river water to various equipment in the station as well as plant cooling systems for heat removal. l The SW system is maintained at a higher pressure than any of the cooling loads. This feature minimizes the chance of contamination entering the service water system and being discharged into the river. The SW I system uns five pumps (2 to each unit, and I as backup) and three strainers (I to each unit, I as backup) to j supply cooling water to a common header that serves both units. Due to the arrangement of the pumps and 1 the common header, the SW is considered a completely shared system. The number and types of cooling l loads supplied are very similar in each unit. A hypochlorite system permits manual introduction of sodium hypochlorite to the water. The addition of the hypochlorite is intended to stop the growth ofliving organisms, such as slime producing bacteria, algae and clams, that can block flow or otherwise reduce heat transfer capability in heat exchangers. This system is no* considered part of the Service Water system, however, it is important to ensure proper operation of the Service Water system. 1 Key System Components ! Component EPN Descriotion 1 2/3-3901 Service Water Swing Pump 2(3F3901 A Service Water A Pump 2(3F3901-B Service Water B Pump 2/3-3902 Service Water Swing Strainer 2(3h3902 Service Water Pump Strainer 2(3F3901 Standby Coolant Supply Inboard Isolation Valve 2(3F3902 Standby Coolant Supply Outboard Isolation Valve 2-3906 Fire Protection Header Supply isolation Valve Operational Modes Plant Operating Mode Initiation operable during normal plant operation, acccrding to following pump operating schedule Operating Mode Summer Conilitions Winter Conditions Both units operating 4 pumps ru .ning 3 pumps running 1 pumn in standby 2 pumps in standby One unit operating, one 3 pumps running 2 pumps running unit in cold shutdown 2 pumps in standby 3 pumps in standby Two units in cold shutdown 2 pumps running 2 pumps running 3 pumps in standby 3 pumps in standby l J D-55 Venticanon Scessing of Key Psameers. Rmsoon 0
l SERVICE WATER (SW) l Function
- The Service Water system provides strained river water for cooling purposes of the following i system / components:
- Reactor Building Closed Cooling Water (RBCCW) Heat Exchangers Turbine Building Closed Cooling Water (TBCCW) Heat Exchangers
- Main Turbine Oil Coolers Reactor Recirculation Pump MG Set Oil Coolers
- Generator Hydrogen Coolers Generator Stator Water Coolers High Pressure Coolant Injection (HPCI) Room Coolers Low Pressure Coolant Injection (LPCI) Room Coolers X Area Room Coolers Main Control Room Air Conditioning Condensers
- Maximum Recycle Concentrator Condensers Maximum Recycle Condensate Holding Tank Coolers
- Off-Gas Glycol Chillers
- Auxiliary Electrical Equipment Room Air Conditioner Condenser Off Gas Filter Building Manifold Sample System Heat Exchanger Chlorinator Pumps Control Rod Drive Pump Coolers (alternate source of cooling)
Integrated Leak Rate Test Air Compressor (Unit 3 only) The Service Water system provides strained river water for operation of the following system / components:
- Containment Cooling Service Water (CCSW) System Keep Fill
- Traveling Screen Wash Spray
- Considered to be SW support functions for risk significant systems Standby Coolant Supply hiode initiation initiated manually Function
- The Service Water system provides strained river water to the standby coolant supply system to ensure an inexhaustible supply of water to the main condenser hotwell.
Fire Protection lleader Backup Supply Afode initiation irdtiated manually Function The Service Water system maintains a fire water header pressure during normal operation, and provides strained river water to the fire protection header when the fire water system is not in use. Station Blackout Afode No additional requirements l Appendit R &fode I No additional requirements I D-55 Venfusuon Sagems of Key Perwncm Rmenm 0
l l SERVICE WATER (SW) l Key Parameters Parameter Value Parameter Reference Calculation Reference SW Pump Minimum Water Level >500 ft (Mean Reference 2 - Section 3.8 C Sea Level) SW Pump Motor Minimum Voltage TBD SW Initial Temperature 175'F Reference 11 (UHS) TBCCW Flowrate 2500 gpm Reference 9 Note: This reference :stablishes that 2500 gpm will be delivered to the TBCCW HX. It does not provide the basis for the sufficiency of the 2500 gpm requirement. CCSW Keep Fill Line Flow 25 gpm Reference 10 Note: This reference establishes the size of orifice required to pass 25 gpm. It does not provide the basis for the sufficiency of the 25gpm requirement. Potential Discrepancies None Notes: No design basis calculations have been found for the service water system. However, a Service Water System Hydraulic Model calculation, ATD-0389, dated 10/7/94, was found. The hydraulic model system parameters were compared to the system design basis parameters utilizing a line-up representing normal operating conditions of the system. The model appears to be reasonably representative of the system. l However, it needs to be benchmarked against test data. i l ! References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Unit 2 & 3 Technical Specification Upgrade (TSUP), January 13,1997 i
D-57 Ven(kation Screen,ng o(Key Pwameters, Rewman 0
._ . - . . . . - - - - _ _ - - - . - ~ . . . . . . . ~ . . - - - . . - - - . - - - . - . - ~ . - - - . . . . -
4 e I i 4 SERVICE WATER (SW)
- 3. Dresden Station Unit 2 & 3 Technical Specification Upgrade Basis (TSUP Basis), Januaryl3,1997
{ i
- 4. Dresden Administrative Technical Requirements (DATR), September 1996 i
j 5. Dresden Technical Specifications, December 1996
- 6. Service Water System Notebook (IPE)
- 7. General Electric APED, Auxiliary System Data Book,257HA654 Rev. 3, April 15,1969.
! 8. General Electric Equipment Manual, GEK T!6, Chapter 4, June 1972. 1 i 9. ' General Electric Equipment Manual, GEK 786, Chapter 19, June 1972 1 { 10. ATD-0256, Flowrate and Orifice Sizing in the CSW Keep fill Line,7/16/93 4 ! . I1. Amendment No.152 to Facility Operating License No. DPR 19 and Amendment No.147 to Faci! sty j Operating License No. DPR-25, Jan,1997. i l I l l D-58 Vanficanoe $cremung o(Key Paresseeers, Revision 0
i . 1 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS) System Description I The Automatic Depressurization System (ADS) provides automatic blowdown of the reactor ves::1 under loss of Coolant Accident (LOCA) conditions for small pipe breaks upon failure of the High Pressure Coolant Injection (HPCI) system. The ADS consists of relays, timers, and associated instrumentation and controls used to provide proper actuation when the appropriate serpoints are reached. The ADS system is designed to actuate to prevent , against reactor pressure vessel overpressurization and to depressurize the reactor in order to allow l operation of the Low Pressure Coolant injection (LPCI) and Core Spray Systems in the event of HPCI system failure. Key System Components Comnonent EPN Descrictica 2(3)-PS-1628A ADS Initiation High Drywell Pressure Switch A 2(3 FPS 1628B ADS Initiation High Drywell Pressure Switch B 2(3 FPS 1629A ADS Initiation High Drywell Pressure Switch A 2(3 FPS-1629B ADS Initiation High Drywell Pressure Switch B 3 LIS 0263 72A ECCS Initiation Ret.etor Vessel Low-Low Level Switch A 3 LIS-0263-728 ECCS Initiation Reactor Vessel Low-Low Level Switch B 3 LIS-0263-72C ECCS Initiation Reactor Vessel Low Low Level Switch C 3-LIS-0263-72D ECCS Initiation Reactor Vessel Low-Low Level Switch D 2-LT-0263-25A2 ECCS/ ADS Initiation Reactor Vessel Low-Low Level Transmitter A2 2-LT-0263-25B2 ECCS/ ADS Initiation Reactor Vessel Low-Low Level Transmitter B2 2-LT-0263 25C2 ECCS/ ADS Initiation Reactor Vessel Low Low Level Transmitter C2 2-LT-C263-25 D2 ECCS/ ADS Initiation Reactor Vessel Low-Low Level Transmitter D2 2(3 FPS-1554A ADS Permissive LPCI Pump Discharge Pressure Switch A 2(3 FPS-1554B ADS Permissive LPCI Pump Discharge Pressure Switch B 2(3) PS-1554C ADS Permissive LPCI Pump Discharge Pressure Switch C 2(3)-PS-1554D ADS Permissive LPCI Pump Discharge Pressure Switch D 2(3)-PS-1554E ADS Permissive LPCI Pump Discharge Pressure Switch E 2(3 FPS 1554F ADS Permissive LPCI Pump Discharge Pressure Switch F 2(3)-PS-1554H ADS Permissive LPCI Pump Discharge Pressure Switch H 2(3 FPS-1554J ADS Permissive LPCI Pump Discharge Pressure Switch J 2(3 FPS 1466A ADS Permissive CS Pump Discharge Pressure Switch A 2(3 FPS-1466B ADS Permissive CS Pump Discharge Pressure Switch B 2(3 FPS-1466C ADS Permissive CS Pump Discharge Pressure Switch C 2(3) PS 1466D ADS Permissive CS Pump Discharge Pressure Switch D 2(3>0287-124A 8.5-minute timer, Initiate ADS on Low-Low Reactor Water Level A 2(3)-0287 1248 8.5-minute timer, initiate ADS on Low-Low Reactor Water Level B 2(3F0287-105A 120 second time delay, Relief Valve Opening A 2(3F0287-105B 120 second time delay, Relief Valve Opening B l D-59 Verificanon screening of Key Parameters, Rev, hon 0
i l i AUf031ATIC DEPRESSURIZATION SYSTEh! (ADS) Operational Modes ! Emergency Core Cooling Afode - <tutomatic Depressuri:ation Initiation when low-low reactor water level is coincident with high drywell pressure (all conditions , must be met) j
- high drywell pressure (four switches - one-out of-two-twice logic) coincident with j low-low reactor water level (four switches one-out-of-two-twice logic)and 120 second timer timed out (one-out-of-one logic) and any LPCI or Core Spray pump running with 2100 psig discharge pressure (one-out-of-two I logic) l 4
Initiation with only low-low reactor water level (all conditions must be met) I i low-low reactor water level (four switches one-out of two-twice logic) continuous for 8.5 i minutes 7 permissive from 8.5 minute time delay (one-out-of-one logic) ; any LPCI or Core Spray pump running with 2100 psig discharge pressure (one-out-of-two I
- logic)
Function l l - The above logic initiates the ADS system to provide reactor depressurization to allow operation of j low pressure coolant systems. l 1 Station Blackout bfode i
- i No additional requirements j itppendit R Afode No additional requirements Key Parameters Parameter Value Parameter Calculation !
Reference Reference l High Drywell 2 2 psig Ref. 8 Sect. 5.3.1 Note: GE Report GE-NE-187-661291 l Pressure Permissive Ref. 2 Table 3.2.B-1 providesjustification for increasing the ! Analytical Limit for the high drywell pressure initiation to 3 psig. The Updated ; Principal LOCA Analysis Parameters for ; Dresden Units 2 & 3 dated September s 1996 will be based on a high drywell pressure initiation of 2.5 psig. ; (Instrument Trip Point) 53" we NED-I EIC-0091, Rev. O,"Drywell ! Pressure Switches (ECCS) Setpoint Error f Analysis at Normal Operating i Conditions" Unit 2 & 3 J (instrument Trip Point) 49" wc NED-I-EIC-0091, Rev.1,"Drywell ! Pressure Switches (ECCS) Setpoint Error i Analysis at Normal Operating l Conditions" Unit 2 & 3. Rev 1 was not implemented, See ECCS system discrepancy. D-60 ver so ..r x.y r., uri, it i e
l I AlHO51ATIC DEPRESSURIZATION SYSTENI (ADS) Parameter Value Parameter Calculation 1 Reference Reference ' Low-Low Reactor 2 84" above TAF Ref.2 Table 3.2.B 1 Note: TSUPs defines TAF as 360"above Water Level 444"(above vesset zero) Ref. 8 Sect. 5.1.4 vessel zero. Reference 11, Permissive 59" (RWL ret"crenced to Ref. I Sect. 6.3.2.3.2 257HA350AM, Nuclear Boiler System l Instrun.ent zero) Data Sheet shows instrument zero at 503" above vessel zero, or 143" above TAF. Therefore these values are consistent. (Instrument Trip Point) 52" (unit 3) NED-I EIC-0100, Rev.4," Reactor Water Level ECCS Initiation /HPCI Turbine Trip indicating Switch Setpoint Error Analysis i at Normal Operating Conditions" Unit 3. l l See ECCS system discrepancy. (Instrument Trip Point) 52.45" (unit 2) NED-I-EIC-0121, Rev. 5," Reactor Water I Level ATWS RPT/ARI Logic and ECCS Initiation Setpoint Analysis, Reactor Pressure ATWS RPT/ARI Logic and Setpoint Analysis" Unit 2. l ADS Relief Valve 120 seconds Ref. I Sect. 6.3.2.4.2 DIS 0250-04, Rev.16," Automatic Opening Time Delay Ref. I Sect. 5.2.2.4 Depressurization System Logic System Ref. I Sect. 7.3.1.4 Functional Test". This procedure Ref. 2 Table 3.2.B-1 functionally tests the time delay relay between I minute 35 seconds to I minute 55 seconds. ADS Initiation Time 8.5 minutes Ref. I Sect. 6.3.2.4.2.D DIS 1400-05, Rev 16," Core Spray Delay for Low-Low s 10 minutes Ref. I Sect. 7.3.1.4 System Logic System Functional Test - ! Reactor Water 510 seconds Ref. 2 Sect. 3.2.B-1 Sebsystem 1". This procedure Level without High Ref. 8 Sect. 4.5.3.2 functionally tests the time delay relay
- Drywell Pressure between 8 minutes to 9 minutes. Note
j The 8.5 minute timer bypasses the low reactor pressure permissive in the Core l Spray and LPCI initiation, and the DW pressure and 120 second timer in ADS, so they will start on low-low reactor water l level only. LPCI Pump 2100 psig Ref. I Sect. 5.2.2.4 NED I-EIC-0086, Rev. 3," Auto Pressure 2100 & 5150 psig Ref. I Sect. 7.3.1.4 Blowdown Permissive, Low Pressure ADS Permissive 2 50 & $ 100 psig Ref.2 Table 3.2.B 1 Coolant Injection, and Core Spray Pumps 106 psig Ref. 5 Table 3.2.2 Discharge Pressure Switches Setpoint Error Analysis " Unit 2 & 3. See (Instrurnent Trip Point) Potential Discrepancy note 1. Core Spray Pump 2100 psig Ref. I Sect. 5.2.2.4 NED-I EIC-0086, Rev. 3," Auto Pressure ADS Ref. I Sect. 7.3.1.4 Blowdown Permissive, Low Pressure i Permissive 2100 & 5150 psig Ref. 2 Table 3.2.B 1 Coolant Injection, and Core Spray Pumps 2 50 & $ 100 psig Ref. 5 Table 3.2.2 Discharge Pressure Switches Setpoint i (Insuument Trip Point) 112 psig Error Analysis " Unit 2 & 3. See f PotentialDiscrepancy note 1. D-6I Vetfeuuon semems or Key Pararse,en Itenson 0
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l AUTOMATIC DEPRESSURIZATION SYSTEM (ADS) Potential Discrepancies *
- 1. The old Technical Specifications have a different pressure value for Low Pressure Core Cooling Pump Discharge Pressure for initiation of ADS, then the Updated Tech Specs. The old Tech Specs have a pressure range of 2 50 & 5100 psig and the TSUPs have a pressure range of 2100 & $ 150 psig.
Dresden Units 2 and 3 Principal LOCA Analysis Parameters, EMF-89-065, Rev. 3, July 1995 , reference the old Tech Spec. Table. An engineering self-assessment has been initiated to assess the implementation of the Setpoint Control Program. This assessment is scheduled to be completed by the end of March,1997.
- j. List of Calculations l 1. Unit 2 & 3," Auto Blowdown Permissive, Low Pressure Coolant Injection, and Core Spray Pumps Discharge Pressure Switches Setpoint Error Analysis" NED-I-EIC-0086, Rev. 3
- 2. Unit 3," Reactor Water Level ECCS Initiation / HPCI Turbine Trip Indicating Switch Setpoint Error Analysis at Normal Operating Conditions, NED-I EIC-0100, Rev. 4 l 3. Unit 2," Reactor Water Level ATWS RPT/ARI Logic and ECCS Initiation Setpoint Analysis, Reactor Pressure ATWS RPT/ARI Logic and Setpoint Analysis" NED-I EIC-0121, Rev. 5 l
- 4. Unit 2 & 3,"Drywell Presst:re Switches (ECCS) Setpoint Error Analysis at Normal Operating Conditions" NED-i-EIC-0091, Rev. O
! 5. Unit 2 & 3,"Drywell Pressure Switches (ECCS) Setpoint Error Analysis at Normal Operating Conditions" NED-t-ElC-0091, Rev.1 (this serpoint not implemented) Referensta l 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP) January 13,1997
- 3. Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis) January 13,1997
- 4. Dresden Administrative Technical Requirements (DATR) September,1996
- 5. Dresden Technical Specifict.tions December 1996
- 6. Design Specification for Nuclear Boiler System 257HA350, Rev. O,6/30/67 l
- 7. Not used.
i l
- 8. Dresden Units 2 and 3 Principal LOCA Analysis Parameters, EMF-89-065, Rev. 3, JtJy !995
- 9. Automatic Depressurization System Logic System Functional Test, D!S 0250-04, Rev.16
- 10. Core Spray System Logic System Functional Test - Subsystem I, DIS 1400-05, Rev.16
- 11. Nuclear Boiler Specification Data Sheet 257HA350AM, Rev.10,11/12nl D-62 v.nc- s.m .t w.r.,wm am o I
4 kV SAFETY RELATED AUXILIARY POWER SYSTEM I
System Description
i The 4 kV Safety Related Auxiliary Power System provides power to safety related and non-safety related i electrical equipment. Normal power supply to the station is provided by the Unit Auxiliary Transformers I (UATs) and Reserve Auxiliary Transformers (RATS) to the 4 kV auxiliary power system. Emergency power is provided by onsite diesel generators. Station Blackout power is supplied by independent onsite diesel generators. I Key System Components Component EPN Descriotion l 2-6300 Unit Auxihary Transformer 21 2-6400 Reserve Auxiliary Transformer 22 3 6300 Unit Auxiliary Transformer 31 3-6400 Reserve Auxiliary Transformer 32 Non-segregated Phase Bus Duct 2 6723 4 kV Switchgear 23 2-6723 1 4 kV Switchgear 231 2 6724 4 kV Switchgear 24 l 2-6724-1 4 kV Switchgear 241 3 6733 4 kV Switchgear 33 3-6733 1 4 kV Switchgear 331 3-6734 4 kV Switchgear 34 l I 3-6734-1 4 kV Switchgear 34-1 Operational Modes Operating Mode l Initiation Not applicable; the 4 kV system is normally operable regardless of plant status. l Function The 4 kV auxiliary power system provides power to safety related and non-safety related electrical l equipment. Control power for the operation of the 4 kV circuit breakers is provided by the 125 Vdc system. l l Station Blackout Mode Afler a Station Blackout (SBO) event, the 4kV system is required atter the Alternate ac (AAC) SBO diesel generators are started. Appendix R Mode The Dresden Appendix R Analysis primarily relies on the procedural tripping of a particular list of associated circuits for a fire in a particular fire zone and takes no credit for coordination of electrical protection. i 4 4 l D-63 Venfmacon Scrum%g of Key Paranetors, Revueon 0
. m ___ _ _ . _ _ _ . _ . _ _ . _ . _ _ . - _ _ _ _ . _ . _ _ _ _ _ _ . _ . . __ _ . _ _ . _ _ _ _
4 kV SAFETY RELATED AUXILIARY POWER SYSTEM l l Kev Parameters i Parameter Value ParameretReferenes falculation Reference Rated Voltage 4160 Engineering Design SL-4544 j Minimum Motor Running Voltage (90% of 4000V, calculated value must exceed minimum) Bus 23 3600V Ref 6, NEMA MG-1 note 1 Bus 231 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198-1819-1, R1 l Bus 24 3600V Ref 6, NEMA MG 1 note 1 l Bus 241 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198-1819-2, RI l Bus 33 3600V Ref 6, NEMA MG 1 note 1 Bus 331 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198-18 19 3, R1 l Bus 34 3600V Ref 6, NEMA MG 1 note I Bus 341 3600V Ref 6, NEMA MG 1 3820V in Calc. 91981819-4, R1 l (3820 is the voltage that must be maintained to ensure proper operation of 480V safety related loads) Minimum Motor Starting Voltage (90% of 4000V, calculated value must exceed minimum) Bus 23 3600V Ref 6, NEMA MG-1 note i Bus 231 3600V Ref 6, NEMA MG-1 3820V in Calc. 9198 18191, Rt ! Bus 24 3600V Ref 6, NEMA MG 1 note 1 Bus 24-1 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198 18 19-2, R1 Bus 33 3600V Ref 6, NEMA MG 1 note i Bus 331 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198 1819 3, RI Bus 34 3600V Ref 6, NEMA MG 1 note i
. Bus 34-1 3600V Ref 6, NEMA MG 1 3820V in Calc. 9198-1819-4, RI l (3820 is the voltage that must be maintained to ensure proper operation of 480V safety related loads) ist Level Undervoltage (2930V, +/- 146V, senses loss of voltage, set below 2nd level UV)
Bus 23 No operating reference 2930V in RSO pgs 125 & 126 Bus 231 2930(+/ 5%) UFSAR Table 8.3-7 sensed by Bus 23 relays l Bus 24 No operating reference 2930V in RSO pgs 178 & 179 Bus 241 2930(+/ 5%) UFSAR Table 8.3-7 sensed by Bus 24 relays Bus 33 No operating reference 2930V in RSO pgs 127 & 128 Bus 331 2930(+/-5%) UFSAR Table 8.3 7 sensed by Bus 33 relays Bus 34 No operating reference 2930V in RSO pgs 180 & 181 Bus 341 2930(+/ 5%)UFSAR Table 8.3 7 sensed by Bus 34 relays 2nd Level Undervoltage (nominal drop out) Bus 231 3835(+/ 7V)UFSARTable 8.3 7 3872V in Calc. 8982-13 19-6, R4 l Bus 241 3820(+/- 7V)UFSAR Table 8.3 7 3872V in Calc. 8982 13 19-6, R4 Bus 331 3884(+/- 7V)UFSAR Table 8.3-7 3872V in Calc. 8982 1719-2, R3 i Bus 341 3870(+/- 7V)UFSAR Table 8.3 7 3872V in Calc. 8982-1719-2, R3 l 2nd Level Undervoltage (ma).imum drop-out) Bus 231 No operating reference 3907V in Calc. 8982 13 19-6, R4 Bus 24-1 No operating reference 3907V in Calc. 8982 13 19-6, R4 2 Bus 331 No operating reference 3907V in Calc. 8982-17-19 2, R3 Bus 34-1 No operating reference 3907V in Calc. 8982-17-19 2, R3 i i l O-64 Venrwance Scrwsmag o(Key Parammars. Reisaan 0
-. - - . , .._ . _ _ . _ _ -_ . .m . ._ , ._ - . = - . _ . , . _ . _ , _
i E 4 kV SAFETY RELATED AUXILIARY POWER SYSTEM
- Parameter Value Parameter Reference Calculation Reference 2nd Level Undervoltage (UV relay setting)
Bus 23 1 3784V TSUP Table 3.2.B l 3871 (+/ 7V), RSO 7/lI/96 Bus 241 3784V TSUP Tabb 3.2.B-1 3872 (+/ 7V), RSO 7/l1/96 j Bus 331 3832V TSUP Table 3.2.B.1 3884 (+/ 7V), RSO 4/28/94 . Bus 341 , 3832V TSUP Table 3.2.B l 3871 (+/- 7V), RSO 4/28/94 a 2nd Level Undervoltage (UV relay time delay setting) l Bus 231 7 sec (+/ 1.4 sec)TSUP Table 3.2.B-1 7 sec (+/- 20%), RSO 7/11/96 Bus 241 7 sec (+/- 1.4 sec)TSUP Table 3.2.B 1 7 sec (+/- 20%), RSO 7/11/96 j Bus 331 7 sec (+/- 1.4 sec)TSUP Table 3.2.B l 7 sec (+/ 10%), RSO 4/28/94 i Bus 341 7 sec (+/- 1.4 sec)TSUP Table 3.2.B-1 7 sec (+/- 10%), RSO 4/28/94 l (The 7 second timer initiates a five minute timer) l 2nd Level Undervoltage (nominal pick-up) Bus 23 1 No operating reference 3891V in Calc. 8982 13 19-6, R4 Bus 24-1 No operating reference 3891 V in Calc. 8982-13 19-6, R4 l Bus 33-1 No operating teference 3891V in Calc. 8982 1719 2, R3 Bus 341 No operating reference 3891V in Calc. 8982 1719-2, R3 l 2nd Level Undervoltage (maximum pick up) Bus 231 No operating reference 3927V in Calc. 8982 13 19-6, R4 Bus 241 No operating reference 3927V in Calc. 8982 13-19-6, R4 Bus 33-1 No operating reference 3927V in Calc. 8982 1719-2, R3 Bus 341 No operating reference 3927V in Calc. 8982 1719 2, R3 Bus Overvoltage (110% of motor rated voltage, calculated value must be below "value") Bus 23 4400V NEMA MG 1 note 3 Bus 231 1400V NEMA MG 1 note 2 Bus 24 4400V NEMA MG 1 note 3 Bus 241 4400V NEMA MG-1 note 2 Bus 33 4400V NEMA MG-1 note 3 Bus 331 4400V NEMA MG 1 V .-110% max in Calc. 9198-69-19-1, R0 Bus 34 4400V NEMA MG 1 note 3 Bus 341 4400V NEMA MG 1 V.,=110% max in Cale. 9198-69-19-1, R0 Protective Relay Settings yes Relay Settin3 Orders (RSOs) note I: This number is determined by the ELMS AC calculations and depends upon the offsite voltage level, auxiliary power system configuration and plant operating status. ELMS AC does not provide a summary of the approximately 12,000 pages of calculations. These buses feed the ( 1) buses so if the voltage at the ( 1) bus is acceptable, then the voltage at this upstream bus is also acceptable, note 2: Procedure DOA 6500-11, Rev. O, sets limits on the 4kV bus voltage to l 10%. note 3: This number is determined by the ELMS-AC calculations and depends upon the offsite voltage level, auxiliary power system configuration and plant operating status. ELMS-AC does not provide a summary of the approximately 12,000 pages of calculations. D-65 Venticanoe 5mening of Key Parameten, Rension 0
l l l 4 kV SAFETY RELATED AUXILIARY POWER SYSTEM l 1 htential Discrepancies ! 1. Table 8.31 of the RUFSAR lists the feeder circuit breakers on 4 kV buses 23 1 and 241 as having ratings of 350 MVA,1200A. Thi circuit breaker rating is inconsistent with the purchase ! specification rating of the subject circuit breakers. The purchase specification rating of the circuit breakers is 250 MVA,1200A. Additionally, these circuit breakers have been analyzed in the short circuit calculations with a rating of 250 MVA, 3200A. As such, this error is suggestive of a typo and does not place the plant in an unanalyzed condition since the correct ratings have been utilized in the auxiliary power short circuit calculations (cales: 7317-43-19-1, rev.18, Unit 2 ELMS-AC, and 7317-43 19-2, rev.16, Unit 3 ELMS-AC). I This discrepancy is being tracked with a PIF.
- 2. Buses 23,24 33, & 34 should be included in DOA 6500-11 to preclude overvoltage to the safety related motors on those buses.
The Dresden 4 kV Bus Overvoltage DOA 6500-11, rev. O may not be complete. This procedure is intended to be utilized in order to mitigate overvoltage conditions on the 4 kV safety related bus 23-1 (331) and 24-1 (34-1). This procedure may be incomplete since it does not address the i voltage levels on the upstream SWGR 23 (33) and 24 (34) which also feed safety-related 4 kV ! Containment Cooling Service Water (CCSW) Pumps A, B, C, and D. He procedure does not l address the upstream bus voltage on which an overvoltage condition could exist and cause safety-l related motor winding damage. The setpoint voltage at which an overvoltage condition is alarmed I is 4.3 kV, this is a computer alarm point. A cursory review of ELMS calculation 7317-43 19-1 resulted in a typical voltage drop between SWGR 23 and 231 to be approximately 10 Vac under ; summer loading conditions, this is suggestive of minimal voltage drop between the buses. In the l subject procedure, the relationship between adding system load and the subsequent bus voltage reduction is not clear and is not entirely consistent with reference 1.0 which determines minimum system KVA load to preclude 4 kV bus overvoltage conditions. It is understood that the intent of l the referenced calculation is only to recommend a minimum system KVA load which will limit i overvoltage conditions. l The long term effects of motor overvoltages could accelerate equipment aging and should be addrested by 4 kV voltage monitoring which includes all safety-related buses. The upstream voltage at SWGR 23 (33) and 24 (34) can be monitored in the control room. This potential problem could be remedied by revising the procedure to have the operator monitor the bus voltage ; at the upstream SWGR if an alarmed overvoltage condition exists. An alarmed overvoltage l coridit:on is defined as a 4 kV bus 23 1 (33-1) or 24-1 (34 1) voltage greater than or equal to 4300 Vac. The subject procedure does recommend adequate operator actions which will reduce an overvoltage condition. The deficiencies in the subject procedure do not pose operability issues. This discrepancy is being tracked with a PIF. D-66 ven6canos $aeenmg of Key Parameurs. Revism 0
l l 4 kV SAFETY RELATED AUXILIARY POWER SYSTEM NOTES
- 1. Switchgear MVA Rating:
UFSAR Table 8.3.1 lists all the breakers on Buses 23 & 24 as 250MVA. The mod approval letter, Chron# 0303335,is for upgrading Buses 23,24,33 & 34 to 350MVA. This is being processed by DFL 95-035 which will update UFSAR for Buses 23,24,33, and 34. l
- 2. Degraded Voltage: l UFSAR Table 8,3-7 lists the degraded voltage setpoints as Bus 23-1,3835V; Bus 24 1,3820V; I Bus 33-1,3884V; Bus 34-1,3870V TSUP Table 3.2.B 1 lists the degraded voltage setpoints as Unit 2,3784V; Unit 3,3832V. The S&L calculations 8982 13-19-6, Rev 4 & 8982 17 19-2, Rev 3, list 3872 as the nominal dropout voltage.
This is being processed by DFL 96064, DFL 96065 & see note 6.
- 3. DBD DR-014," Auxiliary Power System", makes frequent reference to DBD-DR-076," Protection and Coordination", which does not exist.
l Deletion of references to non-existent DBD's is an ongoing station program. 4 The UAT 21 transformer H X & H-Y impedances used in calculation 7317 38-19-1, Rev 2, ELMS AC input data, are approximately 4% higher than actual, i Sargent & Lundy has an analysis that conel ides this is acceptable.
- 5. DAP 07 34, R3,11/6/95, Section G , states "Under certain loading configurations not previously evaluated, there exis*.s the possibility that a three phase bolted fault would cause the current to rise and exceed the interrupt capabilities of the breaker feeding the fault". The loading conditions in the procedure are not identified, but in 1982 Comed was advised that an analysis of the aux power system showed the breakers in some 4kV buses had published interrupting ratings less than the available fault current.
The term " interrupt capability" is not an ANSI defined term. The mod numbers in Section G do not agree with the mod numbers in Section W. This is a procedure that identifies a design basis concern and permits operation outside the design basis and is an issue that has been discussed with the NRC. l l
- 6. DOA 6500-07, R2, Section F states the 2nd level undervoltage set point is 3820 volts. UFSAR l l Table 8.3.7 lists higher values for Buses 23 1,33-1 & 34 1. The calculations for these set points, l 8982 13-19 6, Rev 4 & 8982 17-19-2, Rev 3, list a nominal value and a maximum value (based on relay error). The value used in the DOA should be identified.
The value in the DOA is correct, this is a recommendation to consider expanding the definition of the value. t i i i D-67 Verifwaison $mesmg of Key Paruneners. Iten.non 0 l
l { l 4 kV SAFETY RELATED AUXILIARY POWER SYSTEM l l 7. ER9605522 is for the addition of a 345kV line from Collins to Dresden in 1997. This will have an effect on the min & max voltage and min & max fault curre.nt in the 345kV switchyard at ! Dresden. The impact of these effects on the 4kV system must be analyzed prior to the line being energized. The normally conservative method to analyze an auxiliary power system is to use a switchyard , source impedance based on the switchyard circuit breaker interrupting rating, this permits l switchyard alterations without having to reanalyze the fault duty of the equiptr,ent in the auxiliary power system. The existing auxiliary power system analyses use the existing calculated availabi-switchyard fault current, which is significantly less than the circuit breaker rating. This requhes a reanalysis of the auxiliary power system every time the calculated voltage levels, fault current and/or method of operating the switchyard changes from what was previously analyzed. No procedure has been found to implement this process. The station design engineering group is aware of this item, which will be resolved prior to the line being energized.
- 8. A documented basis for selecting 3820V as the second level undervoltage setpoint could not be found., it appears in the cales as the analytical voltage limit without reference as to where it came from. The starting & running voltages calculations that use 3820V as the minimum bus voltage conclude that that number is acceptable as a minimum voltage (9198 18-19-1, Rev 1; 9198-18-19 2, Rev I; 91981819 3, Rev i & 9198-18-19-4, Revi).
The RSO for the Bus 33 12nd level UV relay has a setpoint of 3884 (+/- 7 volts) inconsistent with calculation 8982-1719-2, Rev 3, page 20, which states a nominal dropout value of 3872 V. The calculations document the value of 3820 is acceptable.
- 9. The RSOs for the 2nd level undervoltage relays for Buses 23-1 & 24-1 have an issue date of 7/11/96 but no completion date. These relays should be checked to ensure they are correctly set.
The RSO for the 2nd level undervoltage relays for Bus 33-1 has an iss e date of 3/16/94 and for Bus 34-1 an issue date of 4/8/94 and have values different than the values in calculations 8982 13-19-7, Rev 4, & 8982 1719-2, Rev 3. These relays should be checked to ensure they are correctly set. The station confirmed that the relays are correctly set and revisions to the RSOs are in progress. References
- 1. Dresden Design Basis Document (DBD), DBD-DR-014, Revision A,"Auxil;ary Power System", dated 10/29/95.
- 2. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 3. S&L Specification K 2175,1/31/1967,4160 Volt Switchgear
! 4. DTR 3081 AMH,4/20/1994, Report of Design Tests on CBSI AMHG 350 MVA Switchgear
- 5. RSO refers to relay setting orders.
- 6. 21 A5580, Rev.1,10/31/72, Motors General Requirements, (GE purchase specification) l D-68 Ventbsam Scen,ag of Key Paramem Restuon 0
480 VAC SAFETY RELATED AUXILIARY POWER SYSTEM j 1 System Descripilen I i ne 480 Vac Safety Related Auxiliary Power System provides power to safety related and non-safety j related electrical equipment. Only the safety related 480 V SWGR and MCC's are listed as key system 1 components. The normal power supply to the 480 Vac switchgear is provided by the 4 kV system through ! 4160 V + 480 V unit substation transformers. 4 System Comnonents l Comoonent EPN Descriotion l j J 2-7328 480 Vac Switchgear Bus 28 I 2 7828-3 480 Vac MCC 28 3 ' 2 7828-2 480 Vac MCC 28-2 2 7828 1 2-7828-7 480 Vac MCC 28-1 480 Vac MCC 28 7 (
\
2-7329 480 Vac Switchgear Bus 29 j 2-7829-3 480 Vac MCC 29-3 L 2 7829 8 480 Vac MCC 29-8 I 2 7829-4 480 Vac MCC 29-4 i 2 7829-2 480 Vac MCC 29 2 j 2 7829-1 480 Vac MCC 291 2 7829 9 480 Vac MCC 29-9 2 7829-7 480 Vac MCC 29-7 3-7338 480 Vac Switchgear Bus 38 3 7838 7 480 Vac MCC 38-7 I 3-7838-2 480 Vac MCC 38-2 j 3-7838-3 480 Vac MCC 38 3 i 3 7838-4 480 Vac MCC 38-4 l 3 7838-1 480 Vac MCC 381 ' 3 7339 480 Vac Switchgear Bus 39 3 7839 1 480 Vac htCC 391 3-7839-2 4SO Vac MCC 39-2 3-7839-7 480 Vac MCC 39-7 i Operational Modes ! t I Operating Mode initiation Not applicable; the 480V system is normally operable regardless of plant status. Function The 480 Vac auxiliary power system provides power to safety related and no-safety related ! electrical equipment. Control power for the operation of the 480 Vac circuit breakers is provided ! by the 125 Vdc system. Manual operation of the circuit breakers is controlled by procedures. I Station Blackout Mode Afier a Station Blackout (SBO) event, the 480V system is required after the Attemate ac (AAC) SBO ' diesel generators are started. D-69 v.nr- so . .e x.y e- n. a.- o I
t 480 VAC SAFETY RELATED AUXILIARY POWER SYSTEM Appeadix R Mode ! The Dresden Appendix R Analysis primarily relies on the procedural tripping of a particular list of ' associated circuits for a fire in a particular fire zone and takes no credit for coordination of electrical l protection. l Key Parameters Parameter Value Parameter Reference Calculation Reference Rated Voltage 480 Engineering Design ref. 3
- Minimum Running Voltage (90% of 460V, motor rated voltage)
Bus 28 414 NEMA MG 1 433 at bus in Calc. 9198 18-191, R1 Bus 29 414 NEMA MG 433 at bus in Calc. 9198 18 19-2, R1 Bus 38 414 NEMA MG 1 433 at bus in Cale. 9198 18-19-3, RI Bus 39 414 NEMA MG 1 437 at bus in Calc. 919818 19-4, R1 Metor Starting Voltage (90% of 460V, calculated value must exceed minimum) Bus 28 414 NEMA MG 1 417 at bus in Calc. 9198-18 19-1, R1 Bus 29 414 NEMA MG 1 41I at bus in Calc. 9198-18 19-2, Rt Bus 38 414 NEMA MG 1 418 at bus in Calc. 9198 18 19-3, R1 Bus 39 414 NEMA MG 1 413 at bus in Calc. 919818-19-4, R1 i l note: Where calculated values are lower than minimum, the calculations document its acceptance. ist Level Undervoltage (70% of 480V, senses loss of voltage, set below 4kV 2nd level UV) ! Bus 28 336 BTP PSB l of NUREG 0800 335 in Calc. 7927 9819-1, R0 Bus 29 336 BTP PSB 1 of NUREG 0800 335 in Calc. 7927 98-19-1, R0 Bus 38 336 BTP PSB 1 of NUREG 0800 335 in Calc. 7927 98-191, R0 Bus 39 336 BTP PSB 1 of NUREG 0800 335 in Cale. 7927-98-19 1, R0 l l 1st Level Undervoltage (UV relay setting) Bus 28 no operating reference 335 (+/- 5%) RSO 12/13/88 Bus 29 no operating reference 335 (+/- 5%) RSO 12/13/88 Bus 38 no operating reference 335 (+/- 5%) RSO 12/13/88 Bus 39 no operating teference 335(+/ 5%)RSO 12/13/88 Bus Overvoltage (based on 110% NEMA standard 460V motors) Bus 28 * $06 NEMA MG 1 506 in Cale. 9198-6919 1, R0 Bus 29
- 506 NEMA MG-1 506 in Calc. 9198-69-19-1, R0 Bus 38
- 506 NEMA MG 1 506 in Calc. 9198-69-19 1, R0 Bus 39
- 506 NEMA MG-1 506 in Cale. 9198-69-19-1, R0
? D-70 Verificauen Screening of Key Parerwtern, Ramme 0 __ _ - _ _ .__ - _ ~ .___ __ _ _ _ _ _ _-.
_ _ . _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ . . _ . _ . . _ _ _ - ._ . . __ __. _ m_ _ _.__ m ._ l I 480 VAC SAFETY RELATED AUXILIARY PO%IR SYSTEM l
- Buses 28,29,38, and 39 feed MCCs with 440V motors that require their terminal voltage be limited to i 484V (DOA 6500-11, R0, concludes that operating the 440V motors at elevated vottages is not expscted to prevent them from performing their function.) j i
l Potential Discrenancies 1 l Notes:
- 1. DBD.DR-014," Auxiliary Power System", makes frequent reference to DBD-DR-076," Protection and Coordination", which does not exist.
Deletion of references to non-existent DBDs is an ongoing station program.
- 2. An operating procedure to limit the potential for overvoltages on the 480V system as required by calculation 9198 69-19 1, R0, has not been identified.
During normal plant operation the loading on these 1500 kVa transformers is in excess of 800 kVa.
- 3. The 600 A frame 480V switchgear breakers have two levels ofinterrupting ratings, depending on l
whether the breaker trip has a time delay or not. This is not a concern as long as the people making trip unit adjustments are aware of the design. References i 1. Dresden Design Basis Document (DBO), DBD-DR-014, Revision A. " Auxiliary Power System", dated l 10/29/95.
- 2. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 3. Sargent & Lundy Report SL-4544,"Dresden Unit 2 and 3 Review of the Electrical Aux Power System Performance", dated 03/15/90, and Supplement 1, dated 03/10/94 l
l l t I l l t D 71 vaif.cmion Screemag of Key Paremsen Revnaea 0
ISOLATION CONDENSER (including Makeup Water)
System Description
The Isolation Condenser system provides core decay heat removal in the event the reactor vessel becomes isolated from the main condenser by Main Steam isolation Valve (MSIV) closure. The system consists of a heat exchanger with two separate tube sections and in a single shell, piping, valves, instrumentation and controls that are used to accept steam from the reactor pressure vessel, and then return condensed steam to the reactor through the Shutdown Cooling and Reactor Recirculation systems. Also included within this system are the makeup water sources to the shell section cf the heat exchanger. There are three different sources of makeup water that can be utilized: (1) clean demineralized water, supplied by either the diesel driven makeup pumps or the clean demineralizer water pumps;(2) river water supplied by the diesel driven fire pump; and (3) contaminated condensate supplied by the condensate transfer pumps or condensate transfer jockey pumps. The system (except for the shell side of the heat exchanger) is at reactor pressure during normal plant operation with only one condsnsate retum valve in the system loop closed. The shell side of the heat exchanger is vented to the atmosphere. For system operation, this condensate return valve opens to retum condensed reactor steam back to the reactor vessel, forming a closed cooling loop. Water makeup to the shell side of the heat exchanger is not required for a minimum of 20 minutes. Kev System Comnonents Comoonent EPN Descriotion 2(3F1302 Isolation Condenser Fleat Exchanger 2(3F1301-3 Isolation Condenser Condensate Return Valve 2(3Fl301-10 Isolation Condenser Makeup Supply Valve 2(3F3319A Condensate Transfer Pump 2(3F3319B Condensate Transfer Pump 2(3F3320 Condensate Transfer Jockey Pump l-4100 Diesel Driven Fire Pump 2/3-4101 Diesel Driven Fire Pump 2(3F4102 Fire Main Makeup Supply Valve 2/3-4303A Clean Demineralized Water Pump 2/3-4303B Clean Demineralized Water Pump 2/3-43122A isolation Condenser Diesel Driven Makeup Pump A 2/3-43122B isolation Condenser Diesel Driven Makeup Pump B 2/3-43123A Isolation Condenser Makeup A Pump Driver 2/3-43123B isolation Condenser Makeup B Pump Driver 2(3F4399-74 Condensate Transfer Makeup Supply Valve 2(3 FPS-0263-53A Reactor Sustained Fligh Pressure Switch 2(3 FPS-0263-$3B Reactor Sustained High Pressure Switch 2(3 FPS-0263 53C Reactor Sustained F{igh Pressure Switch 2(3 FPS 0263-53D Reactor Sustained Fligh Pressure Switch 2(3F595-117A Time Delay Relay 2(3F595-117B Time Delay Relay 2(3F595117C Time Delay Relay 2(3F595-117D Time Delay Relay D 72 Vaificasai 5.:reenmg of Kev Parsmeters Rerson 0
l l ISO 1ATION CONDENSER (including Makeup Water) Operational Modes Operating Atode-NormalPlant Operation The system is in standby during normal plant operation with the reactor pressure greater than 150 psig. In this mode, sufficient water inventory is to be maintained in the shell side of the isolation condenser for the first 20 minutes of operation. Reactor Pressure Vessel (RPQ lsolation Afode initiation automatic on high reactor pressure after a time delay, provided a Group V isolation signal is not present
- capable of being manually initiated l Function 1 The isolation Condenser System provides reactor core cooling following RPV isolation from the l main condenser. Condensed steam is returned by gravity to the reactor via the Shutdown Coolmg i
and Reactor Recirculation Systems. Shell side water is evaporated and released through a vent l line to the atmosphere. Makeup water is available from a number of difference sources. Station Blackout Afode The Station Blackout Mode of the Isolation Condenser System is essentially the same as the RPV isolation Mode. The primary difference is that the makeup is from the fire pumps for the first hour and then from the condensate transfer pumps after the attemate a-c power supply becomes available. Appendix R Afode The Appendix R Mode of the Isolation Condenser System is essentially the same as the RPV ! solation Mode. The primary difference is that the makeup is from the contaminated condensate storage tanks for the first two hours. Dam Failure Afode The Dam Failure Mode (non LOCA) of the Isolation Condenser System is essentially the same as the RPV ! solation Mode. The primary difference is that the ultimate source of makeup is from fire protection system. FloodProtection Afode The Flood Protection Mode of the isolation Condenser System is essentially the same as the RPV isolation Mode, except that it would not be initiated until 6 to 7 hours after reactor shutdown with the reactor depressurized. Long term makeup is provided by portable gasoline driven pumps. Key Parameters Parameter Value Parameter Calculation Reference Reference Heat Removal Rate 252.5 x 10' l (Sec. 5.4.6.1) None identified (Decay Heat at 300 sec) Brulhr & 3 (3/4.50) (Note 1) Initiation Pressure 1070 psig 2 (3/4.2D) & 33,34,35 & 36 Pressure Switch 2(3)-PS-0263 53A D 3 (3/4.2D) (Notes 2,3, & 10) D-73 Venikauan $ceensag of Key Parammers. Remon 0
l ISOLATION CONDENSER (including Makeup Water) Parameter Value Parameter Calculation Referencs Reference initiation Time Delay 17 sec 3 (3/4.5D) 33,34 & 36 Time Delay Relay 2(3)-595117A D (Notes 2 & 3) Makeup Water Flow Decay Heat I (Sec. 5.4.6.3) 13,14,15,18,23, ! at 20 min. & 3 (3/4.5D) 25 26,27,28,30, 1 31 & 34 (Note 4) Latest Analysis Basis 430 gpm 34 - 34 Condensate Transfer Pump 2(3)-3319A B 500 gpm I (Sec, 9.2.6.2) None Identified l Condensate Transfer Jockey Pump 2(3) 3320 70 gpm 1 (Sec. 9.2.6.2) Nonc Identified Diesel Driven Fire Pump (Unit 1) 1-4100 2500 gpm 4 (Sec. 4.1.2.1) 43 Diesel Driven Fire Pump (Unit 2/3) 2/3 4101 2000 gpm 4 (Sec. 4.1.2.1) 43 Clean Demin. Water Pump 2/3-4303A B 225 gpm 38 38 ISCO Diesel Makeup Pump 2/3-43122A B 860 gpm 27 28 - Portable Gasoline Driven Pump 150 gpm 41 None Identified I 1 Minimum Shell Side Water 20 min. of I (Sec. 5.4.6.3) 11,21,22,33 & Volume /I'emp. Operation 34 (Note 5) Onsite Water Storage Volume None Identified None identified 4,8,15, 19,24 & 32 (Note 6) Isolation Condenser Return Valve 2 1301 3 Min. Operator Terminal Voltage 236 39 40 & 42 (Note 8) Thrust Adequacy 2998 lb 39 39 Isolation Condenser Retum Valve 3 1301 3 Min. Operator Terminal Voltage 218 39 40 & 42 (Note 8) Thrust Adequacy 2993lb 39 39 Isolation Condenser Makeup Valve 2 1301 10 Min. Operator Terminal Voltage None Identified None Identified 40 & 42 Thrust Adequacy None Identified None Identified None identified l Isolation Condenser Makeup Valve 3 1301 10 Min. Operator Terminal Voltage None identified None Identified 40 & 42 Thrust Adequacy None Identified None identified None Identified Fire Main Makeup Supply Valve 2-4102 Min. Operator Terminal Voltage None Identified None Identified None Identified Thrust Adequacy None Identified None Identified None Identified Fire Main Makeup Supply Valve 3-4102 l Min. Operator Terminal Voltage None Identified None Identified None Identified j Thrust Adequacy None Identified None identified None Identified l D-74 v.nnoo som.< x, e-= w o l
ISOLATION CONDENSER (including Makeup Water) Parameter Value Parameter Calcubtion Reference Reference Condensate Transfer Supply Valve 2-4399-74 Min. Operator Terminal Voltage 178 44 40 Thrust Adequacy 1860 lb 44 44 Condensate Transfer Supply Valve 3-4399-74 Min. Operator Terminal Voltage 180 44 40 Thrust Adequacy 1860 lb 44 44 Condensate Transfer Pump 2(3)-3319A-B Min. Operator Terminal Voltage None Identified None Identified None Identified Start Voltage None Identified None Identified None Identified Run Voltage None Identified None Identified None Identified NPSH at 500 gpm 18.5 ft 25 25 4 Condensate Transfer Jockey Pump 2(3)-3320 Min. Operator Terminal Voltage None Identified None Identified None Identified Start Voltage None Identified None Identified None Identified Run Voltage None Identified None identified None Identified NPSH None Identified None Identified None Identified Clean Demin. Water Pump 2/3-4303 A-B Min. Operator Terminal Voltage None Identified None Identified None Identified Start Voltage None Identified None Identified None Identified Run Voltage None Identified None Identified None Identified NPSH Pump Curve 37 37 (Note 7) ISCO Diesel Makeup Pump 2/3-43122A-B l NPSH Pump Curve 26 26 (Note 7) l 4 Potential Discrepancies
- 1. No formal calculations have been found that provide a basis for the heat removal rate of the isolation condenser as stated in the UFSAR. See Note i for further information.
[ Note: When the NTS Item is developed, it should be indicated that the original plant d.: cay heat curve is very conservative, and its use can magnify the problems associated with other parameters (water storage and makeup). The use of a more reasonable decay heat curve (e.g., ANS-5.1 - 1979) should be considered to achieve an appropriate balance in the overall isolation condenser design. This approach would also minimize the impact of water carryover on the makeup system performance requirements.]
- 2. The TSUP requires that for initiation of the isolation condenser that there must be a sustained high I pressure signal of 1070 psig for 17 seconds. No formal calculations have been found that provide a l basis for the initiation of this time delay. See Note 2 for further information. l D-75 Venfmanen seasosag of Key Parameers Revaana 0
- . _. _ ~ . -. --. l l ISOLATION CONDENSER Uncluding Makeup Water) [ Note: When the NTS Item is developed, it should be indicated that the calculations performed referencing a 1070 psig initiation pressure and a 15 second delay indicate that the maximum inventory loss is 12,000 lb incorporate a very consersative decay heat curve but neglect the system startup time (including the uncertainty in time delay setting and accuracy) and fuel sensible heat. Any water level calculations perfoimed to demonstrate that there will not be a ECCS initiation signal generated should include these factors.]
- 3. It is believed that calculation BSA-D-95-07 should be used to establish the design basis makeup flowrate of 430 gpm based on NUREG-0800 or 361 gpm based on ANSI-5.1. The isolation system notebook and calculation 282 Y-M 03 should be revised to reflect these makeup flowrates. See Note 4 for further information.
- 4. Currently there is no design basis value for onsite storage of water for continued operation of the isolation condenser system. Onsite storage requirements were given in the original FSAR, but were subsequently removed. He initial fire protection analysis required 8 hour onsite storage, but this was later revised to 2 hours. The required onsite storage supplies needs to be established and calculations should be perfomied to provide the basis for this requirement. See Note 6 for further information.
- 5. Calculation BSA-D-95-07 provides a good basis for the isolation condenser water level and temperature requirements. This document should supersede RSA-D-93-06, which provided an initial basis before BSA-D-95-07 was completed. See Note 5 for further information.
- 6. Several Dresden documents discuss makeup water to the isolation condenser for a single unit case.
However, there are accident scenarios where both units would be affected simultaneously and the isolation condenser would be used concurrently for both units. These documents need to be revised to address dual unit operation of the isolation condenser system. See Note 4 for further information.
- 7. He NPSH calculations for the isolation condenser diesel driven makeup pumps and condensate storage transfer pumps were performed at the rated flow conditions for single pump operation. The calculations should be revised to consider the potential system runout conditions, which could potentially reduce the available margin. See Note 7 for further infonnation.
- 8. The minimum operator terminal voltage for MOVs 2(3) 1301 3 and 2(3) 4399 74 is less than was assumed in the thrust adequacy calculations. The valve opening calculations demonstrate that there is a substantial margin relative to the required thrust. Therefore, these valves should be able to open as required to initiate the isolation condenser system. The thrust adequacy calculations should be revised to reflect the lower terminal voltage. See Note 8 for further information.
- 9. A number of mechanical parameter references or supporting calculations for the isolation condenser system could not be found and retrieved. These parameter references and calculations include: (1) system performance calculations;(2) MOV thrust adequacy; and (3) pump NPSH calculations. See Note 9 for further information.
- 10. A number of ek :trical parameter references or supporting calculations for the isolation condenser system could not be found and retrieved. These parameter references and calculations include: (1)
MOV minimum terminal voltage calculations; and (3) pump minimum terminal voltage, start voltage, and run voltage calculations. See Note 9 for further information. D-76 Venflucoe $creen886 or Key Paremeters, Revision 0
ISOLATION CONDENSER (includingMutp Water) l l
- 11. The pressure switch settings process for the isolation condenser imtiation may contain an error in the l
treatment of the head correction term. As a result, the current instmment settings procedures could
)
increase the potential for spurious system initiation when the total instrument setting uncertainties and ! margin are taken into account. This potential inconsistency should t e evaluated to determine the l acceptability of the inst:ument settings. In additiun, the pressure switch setting process needs to be l incorporated into the isolation condenser performance calcuhtions to assure that the analysis covers the instrument uncertainties. See Note 10 for further information. l l
- 12. Three errors were identified in the DBD. These errors were related to: (1) the description of features not included in the design of Dresden 2 end 3: (2) the identification of event criteria that are not I appropriate for controlling parameters for the isolation condenser system; and (3) references that are not applicable to Dresden 2 and 3.
Heferences l l
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A and Pending l Changes.
- 2. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP), January 13,1997.
3, Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis,(TSUP Basis), January 13,1997
- 4. Dreden Administrative Technical Requirements (DATR), September,1996.
- 5. Not L;ed.
1
- 6. Dresden Design Basis Document (DBD), DBD-DR-171, Revision A," Isolation Condenser System",
dated 01/30/96. l
- 7. System Notebook for Isolation Condenser System.
- 8. Fire Protection Reports Volume 2.
- 9. Station Blackout Docur.ients, Volume 1, Station Blackout Analysis.
- 10. Unit 2(3) Operator's Daily Service Log - Revision 50.
I1. GEK-9542," System 1300 - Isolation Condenser for Dresden-2 Nuclear Power Station, Commonwealth Edison Company," April 1969. (DBD Reference 03.007)
- 12. 21 A1608 Revision 4 - Isolation Condenser - Purchase Specification for Dresden 2 & 3. (DBD Reference 05.001)
- 13. NEDE-10430. (DBD Reference 03.019)
- 14. NEDE-10692. (DBD Reference 03.020) l 15. Memo.," Isolation Condenser Shell Makeup," August 16,1965. (DBD Reference 06.002)
- 16. Memo.," Isolation Condenser System - Dresden 11," June 4,1968. (DBD Reference 06.004)
D-77 Ver&auon screemog of Key Paranweers, Rewman 0
I l
- ISOLATION CONDENSER (including Makeup Water) i l 17. Memo., " Isolation Condenser System - Dresden Ill," July 23,1968. (DBD Reference 06.010) l
- 18. Letter, GE to Comed,"Dresden Units 2 and 3 Isolation Condenser " Dry Pipe" Moisture Separator, May 2,1967 (DBD Reference 06.020)
- 19. Letter, Comed to NRC,"SEP Topic: 113.8, Flooding Potential and Protective Requirements;113.B1, l l capability of operating plants to cope with Design Basis Flood condition; and !!-3.C, Safety Related {'
Water Supply (Ultimate Heat Sink)," November 17,1982. (DBD Reference 06.027)
- 20. Letter, Comed to NRC,"Dresden Station Units 2 and 3 Additional Information on Safe Shutdown Paths (Appendix R)," January 9,1986. (DBD Reference 06.028)
, 21. RSA:94-149," Isolation Condenser Operating Level Recenciliation," October 28,1994. (DBD Refe:ence 06.036) I' j 22. Memo.," Clarification ofInterim Administrative Control of Isolation Shell Side Level," March 17, j 1995. (DBD Reference)
- 23. Dresden 2 -Isolation Condenser (GE Document Undated). (DBD Reference 08.004)
] 24. Sargent & Lundy Evaluation No. D-01001MF. (DBD Reference 08.007) , 1 1 l l 25. Sargent & Lundy Calculation PMED 8982-62-02," Calculations for Condensate Transfer Pump i Suction Piping Pressure Loss and Net Positive Suction Head." (DBD Reference 08.007) . 26. Bechtel Calculation DR 150-M-001, %o-Condenser Suction Line Reroute, Determination of NPSH 4 for New Line Routing." (DBD Referene 08.008)
- 27. Letter," Transmittal of Confirmatory Calculation of Minimum Flow Requirement for isolation Condenser Makeup Feed," January 4,1990 with attached NUS Calculation 282Y M-03, Clean l Demineralizer Pump Upgrade Effects of Changing Impeller Size on Existing Pumps." (DBD I Reference 08.009) I
- 28. Stone & Webster Calculation Ol849.00-M(Cl)-001," Conceptual Pressure Drop for Isolation l Condenser New Diesel Driven Makeup Water System." (DBD Reference 08.012) i
- 29. Stone & Webster Calculation Ol849.00-M(Cl)-003,"lSCO Makeup Pump Useable Volume from Clean Demineralized Water Storage Tank (T-105B)." (DBD Reference 08.010)
- 30. Bechtel Calculation DR-150-M-002," Iso-condenser Make up Pump, Minimum Flow Recire Line Pressure Drop for Orifice Sizing."
- 31. Sargent & Lundy Calculation ATD-0130 Rev. O," Determination of Resistances of the Backup Supply Lines to the Isolation Condensers."
- 32. Original FSAR(Sec.4.5).
- 33. RSA-D-93-06 Revision 2,"Dresden Nuclear Station isolation Condenser Extended Heat Removal Capability."
- 34. BSA D-95-07 Revision 0,"Dresden Isolation Condenser Performance."
D-78 Venocanon Screeneag of Key Pwanwws, llewunm 0
ISOLATION CONDENSER (including Makeup Water)
- 35. NED-I-EIC-120,"Sastained High Reactor Pressure Switch Setpoint Error Analysis at Normal Operating Conditions (DIS 1300-1)," Rev. 0, Unit 2 & 3.
- 36. DIS 1300-01," Sustained High Reactor Pressure Calibration," Revision 07.
- 37. NUS Calculation 282Y-M 01,"Demineralizer Makeup to the isolation Condenser."
- 38. NUS Calculation 282Y-M-02," Clean Demineralizer Pump, Pump Head Estimate, Pump and System Curve."
- 39. Bechtel Calculation 004-MN-305 Revision 0," Thrust Window Calculations 1301 System Valves."
- 40. DRE-96-0126," Motor Terminal Voltage Calculation for Dresden 250 Vdc Motor Operated Valves,"
Revision 0.
- 41. DOA-0010-04," Floods," Revision 9.
- 42. NED-EIC-MOV-DR-0005, Valve Actuator Motor Terminal Voltage Calculation," Rev. O.
- 43. Document #0004987875," Operability Determination of Fire Suppression Water Systems."
- 44. Bechtel Calculation DR-265-M-001,"MOV Differ. Pressure Thrust Window 2(3)-4399 74," Rev.1.
Notes:
- 1. Heat RemovalRate The basic heat removal requirement is for the system to remove decay heat at 5 minutes (300 seconds) following a scram. References 1,3 and 7 identify the required heat removal rate for the isolation condenser as 252.5 x 10' Btu /hr. Reference 7 states that the heat removal rate is approximately 3% of the rated power level. No calculations have been found that identify the basis for this value. However,it has been confirmed that this heat removal requirement was in the original FSAR (Reference 32) that was used as the basis for the original plant operating license.
Based on the discussion in Reference 7, the decay heat curve used in the design of isolation condenser had a value of approximately 3% at 300 seconds. Using a decay heat rate of 3% of 2527 MWt results in a heat removal rate of 258. 7 x 10' Bru/hr. References 6 and 12 establish the isolation condenser design heat transfer rate as 258.7 x 10' Btu /hr with a maximum steam flow rate of 209,270 lb/hr for each tube bundle. As a result, the specified heat removal rate (Reference 12) is conservative relative to the required heat t transfer rate (References I and 3). References 16 and 17 provides the results of calculations that l demonstrate that the tube side of the isolation condenser will pass 4I8,540 lb/hr of flow with about 10% margin. Confirmation that the isolation condensers met the required heat removal rate with supporting calculations was provided during the startup test program (References 13 and 14). These tests demonstrated that the condensate return valve required throttling in order to limit the heat removal capability of the isolation condenser. I I l D 79 Venficanos sawung o(Key Pararwn Revisam 0
l 1 i
- ISOLATION CONDFNSER (including Makeup Water) l j lt should be noted that the decay rate used in the design of the isolation condenser is conservative relative i
3 to the current safety analysis methods. Therefore, it is important to establish what the current requirements i are relative to the required isolation condenser heat removal capability. For example, is the key parameter the decay heat at 300 seconds using the current safety analysis methods or 252.5 x 10' B u/hr as established by the license commitments. His is particularly important relative to the water level and makeup requirements, j b 2. Initbti>n Pressure !. References 2,3,6 and 7 identify the initiation pressure as a reactor dome pressure of 1070 psig. He basic { requirement for the combination of the initiation pressure (PS-0263-53 A D) and time delay (595-l l7A-D) is that the initiation pressure needs to be set high enough that spurious initiations are avoided and low d enough to assure system functioning when required. Thus, the initiation pressure setpoint should be greater than the high pressure scram setpoint and below the relief valve opening setpoint. Further, the delay time should be sufficient to allow the feedwater and bypass systems to fulfill there functions, if they are available. Thus, the initiation pressure and delay time requirements are based on engineeringjudgment and
- supported by the results of the safety analysis. With the current pressure switch setting procedures and
- calculated uncertainties, the initiation pressure could be between about 1033 and 1069 psig vessel dome pressure. As a result, the isolation condenser initiation setpoint could be below the high pressure scram
- setpoint. This situation results because of a potential misinterpretation of the technical specification
'. requirements (Note 10). It would appear that the current setpoints for the initiation pressure and the l initiation time delay satisfy the requirements associated with limiting the loss of reactor inventory. ! However, there are no ca!culations that specifically support this conclusion. $ As discussed above, the initiation pressure and delay time need to be considered together. Reference 6 i states that the initiation pressure is based on a rated RPV dome pressure of 1015 psia and a typical transient j minimum peek pressure rise of 50 psi. Automatic initiation of the isolation condenser at the initiation pressure of 1070 psig ensures that a potential event is in progress. The 15 second delay time ensures that 1 initiation occurs for longer transients and notjust a momentary pressure transient. Note that TSUP bases l allows for a 17 second delay time and the uncertainties in the time delay are not addressed. With the current time delay setting procedures and calculated uncertainties, the time delay could be between about 6 , and 21 seconds. } j Reference 3 states that the isolation condenser system actuation instrumentation is provided to assure , adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater
- flow without providing actuation of any of the core cooling equipment. He Reference i design evalua
- ion j does state that during the first 5 minutes following a reactor scram, the isolation condenser provides i sufficient cooling such that only a few thousand pounds of water is lost from the reactor via blowdown
! through the relief valves. Reference 34 provides the results of calculations referencing a 1070 psig initiation pressure and a 15 second time delay. These calculations indicate that the maximum inventory loss is 12,000 lb. The calculations incorporate a very conservative decay heat curve but neglect the system startup time (including time delay uncertainties) and fuel sensible heat. The conservative decay heat tends 3 to compensate for the other factors, and it is unlikely that an ECCS initiation signal would be generated. j However, a more accurate level calculation should be developed for the design basis conditions to demonstrate that the technical specification bases is satisfied.
- 3. Initiation Time Delay
) References 3 identifies the initiation time delay as 17 sec. See initiation pressure, above, for more
! infonnation relative to the requirements related to time delay.
l 4 D-80
- vwo-so-w.gu a e i
i l LSOLATION CONDENSER (including Makeup Water) l l t Makeup Water Flow l l References I,3,6,7,8 and 9 establish that sufficient inventory is to be maintained in the isolation ! condenser so that it can perform its function for 20 minutes without makeup. The Reference i states that ! with no makeup water, the water stored above the isolation condenser tubes is depleted in 20 minutes. The j decay heat evaluation was based on the original decay heat curves for the station. Following that time period, sufficient makeup needs to be provided, so that it can fulfill its long term cooling requirements. For Dresden 2 and 3, there are three sources of makeup water to the isolation condenser: (1) clean l demineralized water from the demineralized water tank (200,000 gallons); (2) river water; and (3) ! condensate from the contaminated condensate storage tanks (2 - 250,000 gallon common tanks and the l 200,000 gallon Unit I tank). Clean demineralized water can be supplied by either of the two common clean demineralized water transfer pumps or the two common isolation condenser diesel driven makeup pumps. River water can be supp!ied by either the common Unit 2/3 diesel driven fire pump or the Unit I diesel driven fire pump. Condensate can be supplied by 2 unitized condensate transfer pumps or 1 unitized l jockey pump. Demineralized water is the preferred source of makeup water. River water, because ofits ' lower quality, is the backup source. The least desirable source is condensate, because ofits entrained radioactivity. A number of calculations have been identified that are related a the makeup flow requirements. Many of these calculations are historical in nature and do not reflect current isolation performance, particularly the amount of carryover experience during startup testing. There is no comprehensive calculation that establishes the required makeup flow. Based on the latest calculation (Reference 34), a flow of 430 gpm per isolation condenser, initiated at 20 minutes after reactor shutdown is required to satisfy the isolation condenser makeup flow requirement. , To understand the current situation, it is necessary to review the historical precedents. Reference 15 identifies the isolation condenser makeup requirements for early BWRs. The reference design was for a gravity feed makeup tank with an 8 hour storage capacity to supplement the approximate one hour storage capacity contained in the isolation condenser shell. The back pressure was identified as 5 psi. Reference 15 states that, based on the reliability of pumping system power supplies, alternate isolation condenser shell feed systems which rely on external power sources are acceptable. For a pumped system to be acceptable, two ac motor driven makeup pumps, which are connectable to emergency diesel generators are required. Further, the diesel driven fire pump was to be used to put service water into the isolation condenser she'l in the event that all available demineralized water was used and core cooling was still required by the isolation condenser. Condensate transfer pumps having a capacity of 250 gpm were identified as being acceptable for Oyster Creek. Also, a fire pump could be used ifits capacity was 250 gpm and it met seismic design requirements. Reference 15 applied these requirements to Dresden 2. The purchase specification for the isolation condenser (Reference 12) requires that the water carryover not exceed 3%. Reference 17 provides confirmation of the original design requirement as 3% by weight. These original requirements are apparently the reason that many analyses and evaluations refer to a 250 gpm makeup requirement. However,250 gpm is inconsistent with the 20 minute without makeup requirement. The decay heat at 20 minutes in the original design basis decay heat curve is greater than the evaporative loses for a 250 gpm makeup. Reference 23 provides a calculation that establishes a 350 gpm makeup capability. This calculation is based on a reactor decay heat rate of 2.2% (190 x 10' Bru/hr) at 20 minutes with no moisture carryover. The original FSAR (Reference 32) used as the basis for the initial operating license states that, when operating normally, makeup to the shell side of the isolation condenser is added from the condensate storage system at 350 gpm, which is sufficient to remove decay heat 35 minutes after scram. Calculation supporting the Reference 32 makeup requirements have not been retrieved.. However, the Reference 32 capability is consistent with the makeup capability identified in the fire protection evaluations (Reference 8). D-81 Venrksw $creensp(Key Paramesers. Rawson o
l l ISOLATION CONDENSER (including Afakeup Water) As a result of the initial startup testing on the isolation condenser system program (References 13 and 14), l it was determined that the water carryover fraction was substantially greater than the initially specified l value. As a result, additional calculations were performed to determine the required system makeup. l Reference 27 provides the results of sensitivity calculations performed to determine the required makeup I rate. These sensitivity calculations use the 10CFR50 Appendix K decay heat requirements for the loss of coolant accident analysis and assumed water carryover fraction of from 34% to 50%. The calculation concludes that for a 50% carryover fraction a 500 gpm of makeup flow starting at time ofisolstion condenser initiation is adequate. If the carryover fraction drops to 40% within 8 minutes,500 gpm of makeup initiated 4 minutes after scram is adequate. Based on this study, it was determined that a minimum , flow rate of 860 gpm was required for makeup initiated at 20 minutes. This conclusion was based on the l use of the 10CFR50 Appendix K decay heat requirements and a water carryover fraction of 34% at 20 minutes. The most recent system operability calculations (Reference 33) indicate that a makeup rate of 430 l gpm is required a: 20 minutes for one isolation condenser. 'Ihis calculation uses a decay heat curve that bounds the one referenced in the UFSAR and a water carryover model based on startup test data. t Each of two new diesel driven demineralized wmer makeup pumps were designed to provide 860 gpm of water makeup to either one of the isolation condensers. Reference 28 provides the calculations to demonstrate that the diesel driven demineralized water makeup pump can deliver the required flow to the isolation condenser with the isolation condenser shell side pressure at 25 psig. Reference 26 provides the NPSH calculations for the diesel driven demineralized water makeup pumps. The Reference 26 and 28 analyses support the operation of one pump at a time. Reference 30 provides the calculations for the diesel driven demineralized water makeup pump minimum flow recirculation line. The UFSAR identifies the rated flow rate for the condensate transfer pumps as 500 gpm. Reference 25 provides the NPSH calculations for condensate transfer pumps supplying makeup water to the isolation condenser at 500 gpm. Reference 25 also identifies the system total developed head as 165 ft No system performance calculations were retrieve. However, based on other calculations, it would appear that the condensate transfer pumps should be able to deliver about the system rated flow to the isolation condensers. The condensate jockey pump flow rate is too small to support the isolation condenser ma'<eup requirements, except at low decay heat rates. Reference 31 provides the calculation of the resistance in the supply lines to the isolation condensers from the diesel driven fire [ sumps. The calculations are provided for a 1000 gpm of makeup flow to a single isolation condenser. No specific fire protection system calculations have been retrieved that demonstrate the capability of the fire protection system to provide the isolation condenser makeup requirements. Based on the pump test requirement (Reference 4) and the system resistance calculations, the fire protection system has adequate capability to satisfy the isolation condenser makeup requirements as a backup supply. A possible exception is for the makeup requirements during a fire. A operability assessment has been performed to demonstrate a makeup capability of 350 gpm at 20 minutes assuming a fire that has the potential to fail the condensate transfer pumps. Based on the available calculations, adequate makeup flow (currently assumed to be 430 gpm) is available from either the diesel generator makeup pumps or the fire pumps to support the operation of one isolation condenser. However, calculations have not been found to support the simultaneous operation of both Unit 2 and Unit 3 isolation condensers, if this is a requirement. Based on event scenarios, simultaneous operation of both isolation condensers can be postulated for: (1) a loss of offsite power;(2) fire;(3) station blackout; (4) dam failure; and (5) probable maximum flood. As a result some reconciliation calculations should be performed to demonstrate that all required scenarios can be accommodated by the current j diverse makeup capability. These reconciliation calculations should also consider the makeup water l storage requirements. i D-82 Ventkanon $ueenmg o(Key Paramters, Reision 0
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ISOLATION CONDENSER (including 51akeup Water) For postulated floods (Reference 41), provi6ons are made for isolation condenser makeup flow being supplied by two portable 150 gpm gasoline driven pumps. These provisions are consistent with commitments mcde in response to NRC questions during the Systematic Evaluation Program (Reference 19). There is no i pecific time frame in which the portable pumps are required. The minimum time suggested in the ii the response is greater than 6 to 7 hours. No calculations were retrieved that demonstrate the s dequacy of the portable pumps.
- 3. Atinimum ShellSide Water Volume / Temperature The basic water volume and temperature requirement established in Reference I is for no makeup water to be required for the first 20 minutes of an event. This is further clarified in that, with no makeup water, the water stored above the isolation condenser tubes is depleted in 20 minutes. TSUP (Reference 2) requires that the isolation condenser be demonstrated operable at least once per 24 hours by verifying the shell side water volume and the shell side water temperature to be within limits. References 6 and 10 identify the minimum water level required in the isolation condenser as 6.8 ft. Reference 10 also identifies a maximum water temperature of 150 *F. Reference 10 also references the operability technical specification.
This 6.8 ft level is consistent with the operating level reconciliation memo (Reference 21). Reference 21 concludes that even though there will be some level boildown into the tube region to achieve the desired integrated heat removal, the mixture level will still remain above the top of the tubes. The Reference 21 conclusion is based on an NFS calculation (Reference 33), which establishes a 7.1 ft level at 100 *F initial shell side temperature and a 7.5 ft level for elevated temperatures. The level requirements are further clarified in Reference 22. Reference 22 states that preliminary calculations (Reference 33) have been performed supporting an interim acceptable band from 6.6 ft to 7.6 ft for isolation condenser level to assure design basis heat transfer capability. For this range to be valid, the ini tial shell side water temperature must be less than or equal to 120 *F and the heat removal rate must be no greater than 275 x 10' Bru/hr. Further, Reference 22 recommends that Operations attempt to maintain a normal water level at greater than 6.8 ft and within the given range. This is consistent with the current requirements identified in the daily surveillance checks (Reference 10) to demonstrate compliance with Reference 2. It should be noted that the current 6.8 ft requirement (about 22,000 gallons of storage) represents a significant change from the original design. The purchase specification (Reference 12) required a minimum water storage volume above the tubes in the condenser shell of i 1,300 gallons (~ 43 in). The i1,300 gallon was based on the original design requirements that allowed for a 3% carryover fraction. As a result of the startup tests (References 13 and 14), it was identified that there was a potential for a significantly increased carryover fraction. For example, in the calculation for Test No. 2 in Reference 14, it was concluded that with an initial water level of 6.8 ft, the isolation condenser can remove the required amount of heat without makeup, if the isolation condenser retum valve is throttled to give 117% rated hett removal capacity. However, based on Test No. 3, it was recommended that the water level in the isolation condenser be maintained above 7.5 ft. Reference 11 identifies 150* F as the high temperature alarm setpoint for the shell side water volume in the isohtion condenser. A more recent system performance calculation (Reference 34) supersedes the preliminary calculation (Reference 33) and provides a number of acceptable water level and temperature combinations. The acceptable initial water level and temperatures are dependent on the assumed decay heat curve and allowable level penetration into the isolation condenser tube bundle. Reference 34 demonstrates the acceptability of the current daily surveillance check requirements. The preliminary calculation should be superseded and the more recent calculation identified as the design basis calculation relative to shell water level and temperature. As a part of this process, the design basis decay heat curve and allowable water level at 20 minutes should be defined. D-83 Venfman Susensas erKey Pwamesers, Rawsam 0
ISOLATION CONDENSER (including Makeup Water)
- 6. WaterStorage Volume No specific onsite water storage volume requirements have been identified. However, there are inferred onsite water storage requirements associated with makeup system sources and availability. Also, onsite water storage capability can be associated with some of the postulated event scenarios.
Reference 15 indicates that 8 hours of onsite storage was required for the original design. He original FSAR (Reference 32) stated that a total of 90,000 gallons was available in the condensate storage tank for l makeup purposes to the isolation condenser. His quantity of water was sufficient for removing decay heat I for over 8 hours. No calculations supporting this statement have been identified. Reference 29 does indicate that the water storage required for 8 hours ofisolation condenser operation is 90,000 gallons. To meet this volume and prevent vortexing at the pump suction for the diesel driven makeup pumps,73% of ) the volume of the clean demineralized water storage tank is required for single pump operation. l Dresden Administrative Technical Requirements (DATR) for fire protection (Reference 4) establish a l , requirement for a condensate storage of 130,000 gallons / unit. Reference 24 is the calculation supporting l this requirement. This calculation is based on the water makeup required for leakage from the primary l system (25 gpm for 72 hours) and water shrinkage in the primary system due to cooldown. No provision is l included in the analysis for water makeup to the isolation condenser. (Note: There is an implied assumption that the HPCI is available for makeup purposes or the fire protection procedures address actions to transfer CRD supply when the volume is reduced below 90,000 gallon reserved storage level.) Based on the above evaluation, there is no dedicated makeup water supply for fire protection purposes. However, the tire protection reports (Reference 8) state that the long term reactor pressure control and decay heat removal is the isolation condenser system. Further, it is stated that initial makeup will be supplied from the condensate storage tanks via the condensate transfer pumps. Also, it is stated that should it become apparent that long term operation (up to 72 hours) of the isolation condenser is necessary, the opemtor will place a priority on the condensate storage, reserving it for reactor makeup and will establish makeup to the isolation condenser from the service water system (supplied by the diesel driven fire pumps, if necessary). This is further amplified by the statement that, due to the length of time involved (at least 8 hours) before service water is necessary, the fire is assumed to be out and any suppression systems that may have activated are manually isolated from the main header. Reference 19 establishes that water makeup from the service water (fire protection) system J.' not be needed for at least 2 hours after isolation condenser initiation. Further, guidance is te : A vided n, gr ding the availability of sufficient condensate storage water. Rus, there is an inferred coinmitment for a 2 to 8 hour water supply onsite for isolation condenser makeup for fire protection purposes. This may be applicable to both units if it is assumed there is a loss of offsite power to the other unit. In the safety analysis process, credit is taken for the isolation condensers for a loss of offsite power and station blackout. No source of makeup water to the isolation condenser is identified for these events. If no onsite storage is assumed, the makeup would have to be supplied to both units by the diesel driven fire pumps. As discussed relative to the makeup flow requirements above, no calculations have been identified that demonstrate the diesel driven fire pumps can simultaneous support the requirements for both units. A postulated dam failure is a design requirement for Dresden 2 and 3, which requires operation of the isolation condensers. Reference I establishes a requirement for isolation condenser makeup of 2.5 x 10' gallons / unit for a postulated dam failure. The makeup water is supplied by the diesel driven fire pumps or service water pumps. Sufficient water is trapped in the intake and discharge fiumes for the long term decay heat removal. No supporting calculations for this statement were found. D-84 VanfWe %sunmg of Key Paranutm, Rewion 0
__ __ m _ . _ _ _ . _ _ . ____ _ _ _ _ . __.____..____q I J i j ISOLATION CONDENSER (including Makeup Water) i The ability of Dresden 2 and 3 to cope with the probable maximum flood is a requirement. Reference 19 I establishes the commitments on the isolation condenser relative to the probable maximum flood. The requirement is to maintain shutdown for an indefinite period following depressurization to 212 *F in 6 to 7 4 hours. The source and quantity of makeup water to the isolation condenser is not identified. In summary, no onsite water storage requirements for the isolation condenser have been identified. Calculations supporting this position have not been found. Based on event scenarios and licensing 2 documentation, there are some inferred onsite storage requirements. As a result, some reconciliation
- calculations should be performed to demonstrate that all required scenarios can be accommodated. These l reconciliation calculations should also consider the makeup water flow requirements that are discussed above.
- 7. NPSH Calculations
- The NPSH calculations for the isolation condenser diesel driven makeup pumps (Reference 26) were j performed at the rated flow conditions for single pump operation. For these conditions, minimum NPSH l margin is 0.6 ft. No calculations were performed for potential system runout conditions, which could
- potentially reduce the margin further.
l The NPSH calculations for the clean demineralizer water pumps (Reference 37) were performed for a l , variety of conditions to determine the effect of changing pump suction piping sizes. For the current piping l
- configuration, there is no calculated NPSH margin at rated conditions for single pump operation. The i i potential for system runout was not evaluated.
i , 8. MOV Terminal Voltage I The minimum operator terminal voltage (References 40 and 42) for MOVs 2(3) 1301 3 is less than was l assumed in the thrust adequacy calculations (Reference 39). The valve opening calculations demonstrat . 4 that there is a substantial margin relative to the required thrust. Therefore, these valves should be able to l l open as required to initiate the isolation condenser system. It should be noted that References 40 and 42 l
- are inconsistent with respect to calculated motor terminal voltage. The controlling document should be l identified and the other document superseded.
The minimum operator terminal voltage (Reference 42) for MOVs 2(3) 4399-74 is less than was assumed ! in the thrust adequacy calculations (Reference 44). The valve opening calculations demonstrate that there is a substantial margin relative to the required thrust. Therefore, these valves should be able to open as required to allow makeup water to be provided from the clean demineralized water storage tank.
- 9. Calculations Not Retrieved During the review process for the key parameters for the isolation condenser system, a number of
- parameter references or supporting calculations could not be found and retrieved. These parameter
{ references and calculations include: 7 a. A calculation supporting the design basis isolation condenser heat removal rate (Note 1). j b. A system performance calculation for the condensate transfer pumps.
- c. A system performance calculation for the condensate transferjockey pump.
! d. A system performance calculation for the portable gasoline driven pump.
i e. Thrust adequacy calculations for Valves 2(3)-1301-10.
- f. Thrust adequacy and minimum operator terminal voltage calculations for Valves-2(3)-4102.
.l g. Minimum terminal voltage, start voltage, and run voltage for Pumps 2(3)-3319A & B.
- h. Minimum terminal voltage, start voltage, run voltage, and NPSH for Pumps 2(3). 2320.
I D-85 > Wn&asami $ersensag o(Key Paramews Revi,em o
)
ISOLATION CONDENSER (including 5fakeup Water)
- t. Minimum terminal voltage, start voltage, run voltage, and NPSH for Pumps 2/3 4303 A & B.
- 10. PressureSwitchSettings The pressure switch settings process for the isolation condenser initiation may contain an error in the treatment of the head correction term. The technical specifications are established for steam dome pressure. In the pressure switch setting calculations and procedures (References 35 and 36), it is i
established that there is a significant water head at the pressure switch. This results in a significant i reduction in the isolation condenser system initiation pressure relative to the steam dome pressure. As a result, the current instrument setting procedures could increase the potential for spurious system initiation when the total instrument setting uncertainties and margin are taken into account. This potential inconsistency should be evaluated to determine the acceptability of the instrument. Further, the pressure switch setting process needs to be incorporated into the isolation condenser performance calculations to assure that the analysis covers the instrument uncertainties. l D-86 unc-so .tx,r-c.am o
! i l OFFSITE POWER 1
System Description
l i The Dresden offsite power system source consists of the Mid America Interconnected Network (MAIN) . Ris system provides a reliable source for the Dresden 138 and 345 kV switchyards and plant auxiliary 1 power. For the purposes of the Dresden auxiliary power system, the offsite source boundaries consists of the respective switchyards (345 and 138 kV), transmission lines and buswork to the high side of the Reserve Auxiliary Transformers (RAT) and the Unit Auxiliary Transformers * (UAT). The 345 kV j switchyard is connected to the Unit 2 (3) Main Power Transformers (MirT) and the Unit 3 RAT. He 138 kV switchyard is connected to the Unit 2 RAT. The 345 and 138 kV switchyards are interconnected via autotransformers TR81 and TR83. The 345 kV switchyard transmission system consists of six lines on , different rights-of-way. The 138 kV switchyard transmission system consists of six lines on different
- rights-of.way (ref. 6).
) i note: The offsite feed to the UAT is only when backfeeding through the MPT. Normally, the UAT is 3 fed from the main generator. 1 i l j Kev System Components 4 I Comoonent Descriotion 345 kV switchyard Supplies 345 kV to sta' ion: , a Unit 3 RAT q- e can backfeed Unit 3 UAT through MPT 3
- e can backfeed Unit 2 UAT through MPT 2 i
l 138 kV switchyard Supplies 138 kV to station: l
- Unit 2 RAT 2-6200 (MPT2) General Electric 17.1/345 kV,952 MVA Power Transformer J 3-6200 (MPT 3) Westinghouse 17.3/345 kV,986 MVA Power Transformer
] 2(3)-7700 Isolated Phase Bus Duct Operational Modes l Overatine Afode Initiation: Not applicable, the offsite power system is intended to be operable regardless of unit operating status.
- Function
i The function of the offsite power system is to provide a reliable source of 345 kV (Unit 3 RAT 32) and 138 i kV (Unit 2 RAT 22).The system is capable of backfeeding each units respective Unit Auxiliary
- j. Transformer (UAT) through the MPT. The UAT is normally fed from the generator when the unit is in j operation and connected to the MAIN system.
, Station Blackour l During Station Blackout (SBO) mode, it is assumed that the offsite power system is not available. 4 i 1 D-87 v.noaa so . .< u, rma,.m. am o I
)
l 4 i ! ! OFFSITE POWER i hvendir R Mode l' It is assumed in the fire protection report that if offsite power is available, it will be utilized for shutdown. However, for the analysis, Isolation Condenser and HPCI are utilized for unit shutdown when offsite power
- is not available (ref.17) I
{ Key Parameters ' Parameter Value Parameter Reference" Calculation Reference 4 345 kV (Red. Blue) ) Minimum Voltage 342 kV ref.1, section 8.2 ref.13 ] Maximum Voltage 362 kV ref.1, section 8.2 ref.16 Schedule Voltage 354 kV ref.1, section 8.2 ref.9 ) 4 i
- 138 kV ired. Bluel j Minimum Voltage 130,136 kV ref. I, section 8.2 ref.13 l Maximum Voltage 145 kV ref.1, section 8.2 ref.16
- Schedule Voltage unknown *" ref.1, section 8.2 unknown j I
j " note: The switchyard voltage levels which are not explicitly specified in the UFSAR but the associated 1 sections are listed for information. f l *" note: This voltage was not identified in ref. 9. 1 The 345 kV switchyard voltage should be operated at or above liq.kV (ref.10). The 138 kV switchyard voltage should be operated at or above 116 kV (ref.10). These minimum recommended switchyard j voltages are utilized in coniunction with a 4 kV degraded voltage reset point of 3.903 kV in order to { j maintain the plant essential 4 kV bus above its respective reset point. It should be noted that the specific 2 j reset point may be different than 3.903 kV; refer to the 4 kV key parameters for additional details. ! i . Availability of Offsite Source f I Loss of I Source: I with one of the two offsite sources (138 or 345) inoperable, the station shall enter a 7 day LCO to recover j the source or be in HOT SHUTDOWN within 12 hours, COLD SHUTDOWN within the following 24 { hours (ref. 7). Q. l Loss of Both Sources: l With both of the two offsite sources (138 and 345) inoperable, the station shall enter a 24 hour (I day) j LCO to recover one of the two sources or be in HOT SHUTDOWN within 12 hours, COLD SHUTDOWN ! within the following 24 hours. Restore both sources (138 and 345) within 7 days from the time of the initial I loss of power or be in HOT SHUTDOWN within 12 hours, COLD SHUTDOWN within the following 24 l hours (ref. 7). l I i D-88 v.nn.:.- sm .t x, en ac,w o
)
i i . OFFSITE POWER i Potential Discrepancies j e SPOG l-1 A (ref 10) makes a reference to thc Dresden .I k Vsecondlevet undervoltage resetpoint of i 3.903 k V, this value is not consistent with the degraded voltage setpoint (see 4 k V Key Parameter Assessmentfor details).Thefollowing was identiped in a Plf:
- SPOG l 1 A utilizes a second level reset point of 3.903 kV, this value differs from the actual setpoint i
of 3.891 (nominal reset). i Notes:
- There exists a proposed 345 k V transmission line addition betwen Dresden and Collins Stations.
1 This line addition will change the voltage conditions ofthe 345 kVand 138 k Vswitchyards. The 138 switch >urdisfedfrom the 345 k V bus through autotransformers TR8I and TR83. The design ] information regarding: stability, fault currents, and voltage profilefor the proposed line addition j are notreadily available. i i note:Information on theproposed transmission line eddition will be afuture condition which does not j presently exist. However, the afects ofthe new line on the Dresden auxiliarypower system and
- switchyards should be addressedprior to energi
- ation ofthe line.
- A calculation ofthe 123 Vdc voltage dropfrom the main battery to the switchyard cannot befound it l
Is understood that a modification will be pet in place to address the dispos! tion ofthe switch >urdload \ from the main battery. As such, a specupc voltageprofile may be developedfor the switchyard loads as part ofthe subject modification. note: This is not a signspcant discrepancy since the circuit load to the switchyard has been addressed in a 123 Vdc bus voltage drop calculation (8982-66-191, rev. 3). It is recommendedthat the voltage drop to the switchyard relay house be addressedprior to the removal or transfer ofthe 123 Vdcfeed l from the main battery to a local battery in the relay house. l t e SPOG 21 (ref 13) states the voltage levels expectedin the switchyards under various conditions. The j estreme switchyard voltages (max / min)for the nuclear stations (page 7) are unclearfor emergency \ conditions. The subject SPOG states the voltage under normal conditions but does not clearly identify the system load at which time the low voltage occurs. note: This is a comment on the clarity ofthe data and not necessarily a discrepancy.
?
l i l 1 l D-89 Venikatson $memes of Key Pvamete s, Itemnion 0
}
l OFFSITE POWER ' References l
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision Ol A.
2. l Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP); January 13,1997.
- 3. Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis); January 13,1997.
- 4. Dresden Administrative Technical Requirements (DATR); September,1996.
- 5. Dresden Technical Specificati ons; December,1996.
- 6. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A, section 8.2, 1 Offsite Power Systems. l
- 7. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP) January 13,1997, section:
3.9/4.9, dated 03/16/96. This reference was utilized for offsite power loss mitigation data.
- 8. DOP 6400-08,345 kV Voltage Control.
- 9. System Planning Operating Guide (SPOG) 1 1, rev.1, Generating Station Operating Voltage Levels.
- 10. SPOG l-1-A., rev. 0, Operating Nuclear Stations At Reduced Voltage.
I1. SPOG l-2 E, rev. O, Station 12, Dresden 345 kV and 138 kV Bus Operation.
- 12. SPOG l 3, Generating Station Stability.
- 13. SPOG 2-1, rev. O, Transmission System Operating Voltage Levels.
- 14. MAIN Guide No. 2, rev. 3, Simulation Testing Of The MAIN Bulk Power System To Assess Adequacy And Reliabid'y.
- 15. DOP 6100-21, Backfeeding. of Transformer 2 (3) arid Unit Transformer 21 (3 I).
- 16. SPOG 5-3, Operating Procedures During Low System Voltage Conditions.
- 17. Fire Protection Report, Volume 2," Safe Shutdown Report", Amendment 4, dated August 1987 l
l l l l l D-90 Ventkanon Screase g of Key Pwameew,. Rennm 0
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J ? EMERGENCY CORE COOLING SYSTEM (ECCS) INITIATION LOGIC System Descriptian The Emergency Core Cooling Systems (ECCS) are designed to transfer heat from the reactor core following any LOCA at a rate such that fuel and clad damage that would interfere with continued etTective a core cooling is prevented and clad metal water reaction is limited to negligible amounts. The ECCS ' i Initiation Logic consists ofinstrumentation designed to initiate several independent, redundant systems. ' l' The ECCS systems are: Core Spray (CS) System; Low Pressure Coolant injection (LPCI) System; High Pressure Coolant Injection (HPCI) System; Automatic Depressurization (ADS) System. The ADS is l addressed in a separate section. Key System Components The following components are the process sensors that are used to initiate ECCS. Comoonent EPN Descrintion 2(3 FPS-0263 52A LPCU Core Spray Initiation Low Reactor Pressure Switch A l 2(3)-PS-0263 52B LPCU Core Spray Initiation Low Reactor Pressure Switch B ] 3-LIS-0263-72A ECCS Initiation Reactor Vessel Low-Low Level Switch A i 3-LIS-0263 72B ECCS Initiation Reactor Vessel Low-Low Level Switch B l 3-LIS-0263 72C ECCS Initiation Reactor Vessel Low-Low Level Switch C 3 LIS-0263 72D ECCS Initiation Reactor Vessel Low Low Level Switch D 2 LT-0263-25A1 ECCS Initiation Reactor Vessel Low Low Level Transmitter Al
- 2 LT-0263-25A2 ECCS/ ADS Initiation Reactor Vessel Low Low Level Transmi'tet A2
- 2-LT 0263-25BI ECCS Initiation Reactor Vessel Low-Low Level Transmitter B 1
! 2 LT-0263 25B2 ECCS/ ADS Initiation Reactor Vessel Low Low Level Transmitter B2 4 2 LT-0263-25Cl ECCS Initiation Reactor Vessel Low Low Level Transmitter CI i 2 LT-0263 25C2 ECCS/ ADS Initiation Reactor Vessel Low. Low Level Transmitter C2 l 2 LT-0263 25DI ECCS Initiation Reactor Vessel Low Low Level Transmitter D1 4
2 LT-0263 25D2 ECCS/ ADS Initiation Reactor Vessel Low Low Level Transmitter D2 f' 2(3)-0287124A 8.5-minute timer, Initiate ECCS on Low-Low Reactor Water Level A 2(3)-0287-124B 8.5 minute timer, Initiate ECCS on Low-Low Reactor Water Level B 4 2(3)-PS-1632-A LPCU Core Spray / HPCI Initiation High Drywell Pressure Switch A 2(3)-PS 1632-B LPCU Core Spray / HPCI Initiation High Drywell Pressure Switch B j 2(3)-PS 1632-C LPCU Core Spray / HPCI Initiation High Drywell Pressure Switch C . 2(3)-PS-1632-D LPCI! Core Spray / HPCI Initiation High Drywell Pressure Switch D l l Operational Modes i Emergency Core Cooling Mode - High Pressure Coolant injection Initiation (any one of the following is required for initiation)
- high drywell pressure (four switches one-out-of two-twice logic) low low reactor water level (four switches one-out-of-two-twice logic)
Function The above logic initiates the HPCI system to provide reactor pressure vessel core cooling. 5 D-9i Vent!carwe $anness of Key Paramen Raws.on 0
~. - - .-- . . - . - - - . - . - . . - . . - - --
EMERGENCY CORE COOLING SYSTFM (ECCS) INITIATION LOGIC ! Emergency Core Cooling Mode - Core Spray Initiation (any one of the following is required for :nitiation) high drywell pressure (four switches - one-out of two-twice logic) 1 low low reactor water level (four switches - one-out-of two-twice logic) coincident with low I reactor pressure (two switches - one-out-of two logic) l low low reactor water level (four switches one-out-of-two-twice logic) coincident with 8.5 l minute time delay (one-out of-one logic) l Function The above logic initiates the Core Spray system to provide reactor pressure vessel core cooling. Emergency Core Cooling Mode - Low Pressure Coolant injection initiation (any one of the following is required for initiation) high drywell pressure (four swit:hes one-out-of-two-twice logic) low-low reactor water level (four switches one-out-of two-twice logic) coincident with low reactor pressure (two switches - one-out-of-two logic)
- low low reactor water level (four switches one-out-of-two-twice logic) coincident with 8.5 minute time delay (one-out-of-one logic)
Function J The above logic initiates the LPCI system to provide reactor pressure vessel core cooling. l l Station Blackout Mode No additional requirements Appendis R Mode No additional requirements j 1 Key Parameters l Parameter Value Parameter Calculation I Reference Reference i i CS Initiation 2 84" above TAF Ref.2 Table 3.2.B-1 Note: TSUPs defines TAF as 360" Low Low Ref. 5 Table 3.2.2 above vessel zero. Reference 7, Reactor 444"(abon wsset rem) Ref. 6 Sect. 5.1.6 257HA350AM, Nuclear Boiler ' Water Level 59" (RW1. referenced to Ref. I Sect. 6.3.2.3.2 System Data Sheet shows instrument la5tmment2) zero at 503" above vessel zero, or 143" above TAF. Therefore these values are consistent. (Instmment Trip Point) 52" (unit 3) NED I-EIC-0100, Rev. 4," Reactor Water Level ECCS Initiation /HPCI Turbine Trip Indicating Switch Setpoint Error Analysis at Normal Operating Conditions" Unit 3. See Potential Discrepanty note 2. (Instrument Trip Point) 52.45" (unit 2) NEG ! EIC n!2!, Rev. 5," Reactor Water Level ATWS RPT/ARI Logic and ECCS Initiation Setpoint Analysis, Reactor Pressure ATWS RPT/ARI Logic and Setpoint Analysis" Unit 2. D-92 vena sm ..nor = m, % o
_ - - _ . _ . _ . . ___.__._._____.._._-_.--_____m...._--___ _ _ . . _ l I E31ERGENCY CORE COOLING SYSTE31 (ECCS) INITIATION LOGIC j l Parametet Value Parameter Calculation i Reference Reference Low Reactor 300 psig Ret 6 Sect. 5.2.2 Pressure 350 psig Ref. I Sect. 6.3.3.4.2.1 2:300 & 5 350 psig Ref. 2 Table 3.2.B 1 (Instrument Trip Point) 338.5 psig NED-I EIC-0099, Rev.2," Reactor l I Low Pressure (350 PSIG)ECCS Permissive Setpoint Error Analysis at Normal Operating Conditions" Unit 2 & 3. l l High Drywell 52 psig Ref. 6, Sect. 5.3.3 Note: GE Report GE-NE-187 66-l Pressure Ref. 2 Table 3.2.B-1 1291 providesjustification for l Ref. 5 Table 3.2.2 increasing the Analytical Limit for j Re f. Sect. 7.3.1.2.1 the high drywell pressure initiation to 3 psig. Note: The Updated PrincipalLOCA Analysis Parameters for Dresden Units 2 & 3 dated September 1996 will be based on a high drywell pressure initiation of 2.5 psig. (Instrument Trip Point) 49" we ) NED-I E!C-0091, Rev.1,"Drywell i Pressure Switches (ECCS) Setpoint ' Error Analysis at Normal Operating Conditions" Unit 2 & 3. See Potential Discrepancy note 1. (Ins.rument Trip Point) 53" we NED I EIC 0091, Rev.O"Drywell Pressure Switches (ECCS) Setpoint Error Analysis at Normat operating Conditions" Unit 2 & 3. See Potential Discrepancy note 1. CS Initiation 8.5 minutes Ref.1 Table 6.3 1 Note i DIS 1400-05, Rev 16," Core Spray Time Delay with Ref. I Sect. 7.3.1.2.1 System Logic System Functional Low Low $ 10 minutes Ref. 2 Table 3.2.B-1 Test - Subsystem 1" His procedure Reactor Water tests setting of time delay relay Level between 8.0 minutes to 9.0 minutes. Note: De 8.5 minute timer bypasses the low reactor pressure permissive in the Core Spray and LPCI initiation, so they will start on low-low reactor water level only. Note:This timer is in the ADS logic. Core Soray 3 seconds Ref. I Sect. 6.3.2.1.4.1 Note: This time response is for' Subsystem Logic initiation atler sensing Logic Response Low-Low Reactor Water Level or Time. High Drywell Pressure. I D-93 v.nt- so . .exc.n , n.m .
l l t ESIERGENCY CORE COOLING SYSTESt (ECCS) INITIATION LOGIC l [ Parameter Value Parameter Calettation j Reference Referengs 4 LPCI Initiation > 84" above TAF Ref. 2 Table 3.2.B 1 Note: TSUPs defines TAF as 360"
- Low Low Ref. 5 Table 3.2.2 above vessel zero. Reference 7, ;
Reactor 444* (above s esset zero) Ref. 6 Sect. 5.1.5 257HA350AM, Nuclear Boiler l Water level 59" (RWL referenced to Ref. I Sect.6.3.23.2 System Data Sheet shows instrument instrument zero) ] q zero at 503" above vessel zero, or 143" above TAF. Therefore these l } values are consistent. j (Instrutnent Trip Point) -52" (unit 3) NED I-EIC-0100, Rev. 4," Reactor - l, Water Level ECCS Initiation /HPCI Turbine Trip Indicating Switch Setpoint Error Analysis at Normal
- Operating Conditions" Unit 3. See
}' Potential Discrepancy note 2. (Instrutnent Trip Point) 52.45" (unit 2) NED-I EIC-0121, Rev. 5," Reactor i Water Level ATWS RPT/ARI Logic !- and ECCS Initiation Setpoint i Analysis, Reactor Pressu e ATWS RPT/AR' Logic and Serpoint l Analysis" Unit 2. Low Reactor 300 psig Ref. 6 Sect. 5.23 ] Pressure 350 psig Ref. I Sect. 633.4.2.1
> 300 & 5 350 psig Ref.2 Table 3.2.B-1 (Instrument Trip Point) 338 $ psig NED-I EIC-0099, Rev.2," Reactor Low Pressure (350 PSIG) ECCS l
- Permissive Setpoint Error Analysis at l Normal Operating Conditions" Unit 1 2&3 i
j High Drywell $ 2 psig Ref. 6, Sect. 53.2 Note: GE Report GE NE 187 1 Pressure Ref.1, Sect. 73.1.2.1 1291 providesjustification for Ref.2 Table 3.2.B 1 increasing the Analytical Limit for Ref. 5 Table 3.2.2 the high drywell pressure initiation to l, 3 psig. Note: The Updated Principal l LOCA Analysis Parameters for Dresden Units 2 & 3 dated September 1996 will be based on a high drywell ptessure initiation of 2.5psig. (Instrument Trip Point) 49" we NED-I EIC-0091, Rev. I,"Drywell Pressure Switches (ECCS)Setpoint Error Ar.alysis at Normal Operating Conditions" Unit 2 & 3. See Potential Discrepancy note 1. (Instrument Trip Point) 53" we NED 1.EIC-0091. Rev. 0 "Drywell Pressure Switches (ECCS)Setpoint Error Analysis at Normal Operating Conditions" Unit 2 & 3. See Potential Discrepancy note 1. D-94 Verikanee susuung or Key Partneters, llewmoa 0
l
.i 1
EMERGENCY CORE COOLING SYSTEM (ECCS) INITIATION LOGIC i 1 Parameter yalus Parameter Calculation ! Egfsrence Reference ! LPCI Initiation 8.5 minutes Ref. I Table 6.31 Note DIS 1400-05, Rev 16. " Core Spray Time Delay with 1 System Logic System Functional
, Low Low 510 minutes Ref. I Sect. 7.3.1.2.1 Test Subsystem I" This procedure 1
Reactor Water Ref. 2 Table 3.2.B 1 tests setting of time delay between I i- Level 8.0 minutes to 9.0 minutes. Note: The 8.5 minute timer bypasses the low reactor pressure permissive in the CS and LPCI initiation, so they will a 2 start on low low RX water level only. This timer is in the ADS logic. HPCI In.fiallan > 84" above TAF _ Ref.2 Table 3.2.B 1 Note:TSUPs defines TAF as 360" f Low Low above vessel zero. Reference 7, l Reactoe Water 444"(above vesset zero) Ref. 6 Sect. 5.1.3 257HA350AM, Nuclear Boiler Level -59" (RW1. referenced to Ref. I Sect. 6.3.2.3.2 System Data Sheet shows instrument in58 ""'z'm) zero at 503" above vessel zero, or j 143" above TAF. Therefore these values are consistent. (Instrument Trip Point) -52" (unit 3) NED-I EIC-0100, Rev. 4 " Reactor Water Level ECCS Initiation /HPCI
- Turbine Trip Indicating Switch Setpoint Error Analysis at Normal
- Operating Conditions" Unit 3. See Potential Discrepancy note 2.
(Instrurnent Trip Point) -52.45" (unit 2) NED I EIC-0121, Rev. 5," Reactor 4 Water Level ATWS RPT/ARI Logic j and ECCS Initiation Setpoint j Analysis, Reactor Pressure ATWS RPT/ARI Logic and Setpoint 1 Analysis" Unit 2. j . High Drywell 5 2.0 psig Ref. 6, Sect,5.3.4 Note: GE Report GE NE 187-66
- Pressure Ref.1, Sect. 6.3.2.3.2 1291 providesjustification for
] Ref. 2 Table 3.2.B-1 increasing the Analytical Limit for j Ref. 5 Table 3.2.2 the high dr>well pressure initiation to } 3 psig. Note: The Updated Principal j LOCA Analysis Parameters for Dresden Units 2 & 3 dated September 1996 will be based on a high drywell pressure initiation of 2.5 psig. (Instrument Trip Point) 49"we NED-I EIC-0091, Rev.1,"Drywell Pressure Switches (ECCS)Setpoint {l Error Analysis at NormalOperating Conditions" Unit 2 & 3. See i Potential Discrepancy note 1. l . D-95 4 v.,com so-.y a x ,,--n, n-- o
DIERGENCY CORE COOLING SYSTDI (ECCS) INITIATION LOGIC l Parameter Yalue Parameter Calculation Referenqg Reference (Instrurnent Trip Point) 53" we l NE D-I EIC-0091, Rev. 0 "Dryw ell i Pressure Switches (ECCS) Setpoint Error Analysis at Normal Operating Conditions" Unit 2 & 3. See Potential Discrepancy note 1. , Potential Discrepancies * '
- l. The Analytical Limit for ECCS initiation on High drywell pressure is listed as 2 psig. in the TSUP's and the original Tech Specs. The existing setpoint for high drywell pressure initiation of Core Spray, LPCI, HPCI and ADS in the ECCW IDS data sheets is 53" we., equivalent to 1.92 psig. Instrument setpoint calculation NED-I EIC-0091, Rev 0, dated !2-28-92, has a setpoint of 53", which results in a negative margin. An operability determination was performed on 12 30-92, with input provided by GE Report GE-NE-187-66-1291. This report provided a safety evaluation to justify increasing the analytical limit (AV) for the Drywell high pressure signal to 3 psig, stating there would not be any significant effect on the timing of the reactor scram & ECCS initiation because of the rapid pressurization rares for postulated events.
On 4-16 93, Revision I was made to this calculation, revising the setpoint to a value of 49" we. (equivalent to 1.77 psig) with a positive margin. However, this revised setpoint was not implemented. The Instrument calibration procedure DlS 1600-04, Revision 13, still establishes the instrument setpoint as 53" 51" we. This calculation also applies a head correction for drywell atmosphere (air). Although the head correction established is negligible, it should not have been used in this application, sine the instrument is located above the instrument tap, and the sensing line is considered to be dry. Note: NED-I-EIC 0091 is being revised to address the potential discrepancies described above.
- 2. In calculating the margin for the unit 3 RX water level initiation signal for ECCS, instrument setpoint calculation NED I-EIC-0100 used an incorrect setpoint value of 51" RWL instead of the correct value of-52" RWL. The margin that was calculated is +2.61 "RWL for instruments LIS-003-0263 72A/C, and + 3.04" RWL for instruments LIS-003-0263 72B/D. If the correct setpoint value of 52 " RWL is used, the margin would be less, but should still be positive. This calculation should be revised as appropriate to determine the correct margin for this setpoint.
An engineering self assessment has been initiated to assess the implementation of the Setpoint Control Program. This assessment is scheduled to be completed by the end of March,1997. List of Calculations
- 1. Unit 3, " Reactor Water Level ECCS Initiation / HPCI Turbine Trip indicating Switch Setpoint Error Analysis at Normal Operating Conditions, NED-I EIC-0100, Rev. 4
- 2. Unit 2," Reactor Water Level ATWS RPT/ARI Logic and ECCS Initiation Serpoint Analysis, Reactor Pressure ATWS RPT/ARI Logic and Setpoint Analysis" NED-I EIC-0121, Rev. 5
- 3. Unit 2 & 3," Reactor Low Pressure (350 PSIG) ECCS Permissise Setpoint Error Analysis at Normal Operating Conditions" NED-I EIC-0099, Rev. 2 D-96 Vedcasma $creemag of Key Parenscars Re* men 0
.. . . _ _ ~ . _ . _ . ._._ ____._ - - . . _ _ . _ _ . _ . _ _ . _ _ . _ . . _ . . _ _ _ . - . _ .
4 EMERGENCY CORE COOLING SYSTEM (ECCS) INITIATION LOGIC 1 i 4. Unit 2 & 3 "Drywell Pressure Switches (ECCS) Setpoint Error Analysis at Normal Operating 1 Conditions" NED l EIC-009i, Rev.1(this setpoint not implemented i
- 5. Unit 2 & 3,"Drywell Pressure Switches (ECCS) Setpoint Error Analysis at Normal Operating Conditions" NED 1 ElC-0091, Rey,0
- References l
- l. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Unit 2 & 3 Technical Specifications Upgrade (TSUP) January 13,1997 j 3. Dresden Station Unit 2 & 3 Technical Specifications Upgrade Basis (TSUP Basis) January 13,1997 i
- 4. Dresden Administrative Technical Requirements (DATR) September,1996 t
- 5. Dresden Technical Specifications December 1996 l
- 6. Dresden Units 2 and 3 Principal LOCA Analysis Parameters, EMF-89-065, Rev. 3, July 1995
! 7. Nuclear Boiler Specification Data Sheet 257HA350AM, Rev.10, i1/12nI i i ! l I 4 i i k i i 1 4 4 D-97 Ver&auon 5creenes of Key Parameeers, Re.wice 0 4___
1 1 1 HIGH PRESSURE COOLANT INJECTION (HPCI)
System Description
The HPCI system is designed to maintain reactor water coolant inventory to provide adequate core cooling in the event of a loss-of coolant accident (LOCA) which does not result in rapid depressurization of the pressure vessel. The loss-of-coolant may be due to a loss of reactor feedwater or a small line break which l does not cause immediate depressurization of the reactor vessel. The HPCI system allows for plant ! shutdown by maintaining sufficient reactor water inventory until the reactor is depressurized to a point where the Low Pressure Coolant Injection (LPCI) system and/or Core Spray (CS) system can inject water. The HPCI system includes a steam turbine driving a two-stage high pressure main pump, a gear-driven, single-stage booster pump, valves, high pressure piping, water sources, and instrumentation. The HPCI system is normally aligned to take suction from the Condensate Storage Tank (CST) which provides a clean source of cooling water. However, the HPCI suction is automatically aligned to the pressure suppression pool which provides a safety related source of water for injection to the reactor core upon tank low water level or high suppression pool water level. Kev System Components Comnonent EPN Descriotion 2(3F2302 HPCI Main and Booster Pump 2(3F2303 HPCI Turbine 2(3F2301-3 HPCI Turbine Steam Supply Valve 2(3)-2301-6 HPCI Pump Condensate Storage Tank (CST) Suction Isolation Valve 2(3)-2301-7 HPCI Pump Discharge Check Valve 2(3F2301-8 HPCI Pump Discharge Shutoff Valve 2(3)-2301 14 HPCI Pump Minimum Flow Bypass Valve 2(3) 2301-28 HPCI Steam Supply Drain Pot Suppression Pool Drain Valve 2(3F230129 HPCI Steam Supply Drain Pool Main Condenser inboard Drain Valve 2(3F230130 HPCI Steam Supply Drain Pool Main Condenser Outboard Drain Valve 2(3)-2301-35 HPCI Pump Suppression Pool Suction Outboard Isolation Valve 2(3F230136 HPCI Pump Suppression Pool Suction inbard Isolation Valve 2(3F230164 HPCI Turbine Stop Valve inboard Drain Valve 2(3F2301-65 HPCI Turbine Stop Valve Outboard Drain Valve 2(3F2301-AOP HPCI Auxiliary Oil Pump (AOP) 2(3F2301-STPV HPCI Turbine Stop Valve 2(3F2320-GSEF HPCI Gland Seal Exhaust Fan (GSEF) 2/3-LS-2350-A Condensate Storage Tank (CST) Low Water Level Switch A 2/3 LS-2350-B Condensate Storage Tank (CST) Low Water Level Switch B 2/3-LS-2350-C Condensate Storage Tank (CST) Low Water Level Switch C 2/3 LS-2350-D Condensate Storage Tank (CST) Low Water Level Switch D 2(3FLS-2351 A Suppression Pool High Water Level Switch A 2(3FLS-2351 B Suppression Pool High Water Level Switch B 2(3FDPT-2352 HPCI Steam Supply Line High Flow Differential Pressure Transmitter 2(3FDPT 2353 HPCI Steam Supply Line High Flow Differential Pressure Transmitter 2(3FFS-2354 HPCI Discharge Bypass High! Low Flow Switch D 98 Verneuce Screemag of Ke Paranen Revuon 0
- - . - . - . - . - .- . . - - - - = - - - - -- - - - -- - - --- - - - - -~- - HIGH PRESSURE COOLANT INJECTION 0[PCI Comoonent EPN Descriotion 2(3FTS 2370 A HPCI Steam Line Detection High Area Temperature Switch A l 2(3FTS-2370-B HPCI Steam Line Detection High Area Temperature Switch B l 2(3FTS 2370-C HPCI Steam Line Detection High Area Temperature Switch C 2(3FTS 2370-D HPCI Steam Line Detection High Area Temperature Switch D 2(3FTS 2371 A HPCI Steam Line Detection High Area Temperature Switch A 2(3FTS-2371 B HPCI Steam Line Detection High Area Temperature Switch B 2(3FTS-2371 C HPCI Steam Line Detection High Area Temperature Switch C l 2(3FTS 2371 D HPCI Steam Line Detection High Area Temperature Switch D I 2(3FTS-2372 A HPCI Steam Line Detection High Area Temperature Switch A 2(3FTS 2372 B HPCI Steam Line Detection High Area Temperature Switch B i 2(3FTS 2372-C HPCI Steam Line Detection High Area Temperature Switch C l 2(3FTS 2372 D HPCI Steam Line Detection High Area Temperature Switch D 2(3FTS 2373-A HPCI Steam Line Detection High Area Temperature Switch A 2(3FTS-2373 B HPCI Steam Line Detection High Area Temperature Switch B 2(3FTS-2373 C HPCI Steam Line Detection High Area Temperature Switch C l 2(3FTS-2373-D HPCI Steam Line Detection High Area Temperature Switch D ; 2(3FPT 2389-A HPCI Steam Supply Line Low Pressure Transmitter A ' 2(3FPT-2389 B HPCI Steam Supply Line Low Pressure Transmitter B 2(3FPT 2389-C HPCI Steam Supply Line Low Pressure Transmitter C 2(3FPT-2389-D HPCI Steam Supply Line Low Pressure Transmitter D Operational Modgs injection Mode Initiation low low reactor water level
- high drywell pressure
- manually initiated Function The HPCI system provides adequate reactor core cooling to maintain reactor water coolant inventory without assistance from other mechanical systems.
1 Auto-Isolation Mode (Group 4 Isolation) Initiation (isolation of the steam supply will occur if any of the following conditions are present)
- high steam line flow
- high HPCI toom temperature
- low steam line pressure Function The steam supply to the HPCI system is isolated to minimize damage of the HPCI equipment and allow for ADS and LPCI/ Core Spray actuation and limit inventory loss through the break.
D-99 v.nra.= so -. on , en..m %.a o
HIGH PRESSURE COOLAST INJECTION (HPCD Turbine Trip Mode Initiation (closure of the turbine stop valve will occur if any of the following conditions are present)
- turbine overspeed
- low pump suction pressure
- high turbine exhaust pressure
- reactor vessel high water level
- tripped manually Function ne steen supply to the HPCI system is isolated to minimize turbine damage, pump damage, and prevent v.'ssel overfill.
Station Blackout of de
~
No additional requirements i AppenditR Afode l During a fire on the isolation condenser floor of the reactor building (Elev. 589' 0"), the isolation l condenser pipe chase, and Shutdown Cooling Pump Room (Unit 2) or Transversing In-core Probe (TIP) Room (Unit 3); the HPCI system is used for shutdown. He HPCI system is used by the operator to control reactor pressure and maintain reactor water level. During this mode, the HPCI system must have the capability of restart as needed by the operator. Kev Parameters Parameter Value Parameter Reference c alculation Reference HPCI Injection Ilow at 5000 gpm at Reference 9 Discrepancy 11 Reactor Pressure reactor pressure Reference 16 between 150 psid and i150 psid (differential pressure is between reactor vessel and primary i containment) HPCI Pump Net Positive 25 feet Reference 6,7 Reference 13 Suction Head (NPSH) (at 5,600 gpm) Reference 1 - table 6.3 7 Required at Design Flow HPCI Pump Net Positive Suction Head (NPSH) Available When aligned to CST $9 feet Reference 11 Discrepancy 2 When aligned to torus 30 feet Reference 10 - section 6.3.3.4.3 D:screpancy 3 HPCI Turbine Supply 1125 psia Reference 6,7 Discrepancy 4 High Pressure and Flow 145,000 3s/hr Reference 1 - table 6.3 7 Rate D 100
- v. ire so .txn r, % e
- - . . _ - . ~ _
- _ _ - . - - . . - - . _ . - - - . . . - . - . _ _ - _ . . . - . . . - ~ . - - ~ . - - - .
1 l HIGH PRESSURE COOLANT INJECTION (HPCD i Parameter Valut Pinimeter Reference C;dculation Reference I j HPCI Turbine Supply 155 psia Reference 6,7 Discrepancy 4 i Low Pressure and Flow 102,500 lbs'hr Reference 1 - table 6.3 7
, Rate HPCIinitiation Time 25 seconds Reference 6,7,8 Reference 31 Referecce 1 table 6.3 7 i,
Referen:e 9 I i HPCIblinimum Flow Bypass Valve Setpoints l HPCI Pump Discharge 600 gpm Reference 2 table 3.2.B 1 Reference IS Low Flow Bypass Reference 6,7
- Setpoint j (Instrument Trip Setpoint) 715 gpm l
HPCI Pump Discharge 1200 gpm Reference 6 Reference 18 l l High Flow - Bypass
- Setpoint (Instrurnent Trip Setpoint) I127 ppm l
}
- l HPCISystem Auto isolation Serpoints (Group 4 isolation)
High Steam Line 300 % rated Reference ! - table 7.3 1 Reference 20 l Flow steam flow Reference 2 - table 3.2.A 1 q Reference 10 section 7.3.2.3 1 l High HPCI Room 200*F Reference I table 7.31 Reference 17 ! j Temperature Reference 2 - table 3.2.A 1 !
- (Instrument Trip Setpoint) 180*F Reference 10 section 7.3.2.3 l j Reference 6 j Low Steam Line 100 psig Reference 1 - tabis 7.3-1 Reference 19 i Pressure Reference 2 - table 3.2.A 1 Discrepancy 5
! (tastrument Trip Serpoint) 130.1 psig Reference 10 - i (ineluding 25 psig section 6.3.2.3.3.4,7.3.2.3 head correction) Reference 6 l HPCI Pump Suction Switch Overfrom Condensate Storage Tank (CST) to Suppression PoolSerpoints CST Low Water Level 10,000 gallons Reference 2 - table 3.2.B 1 These level switches
- i. Reference 6 are not calibratable.
! Reference 4 - table 3.12 1 Setpoint calculations do not apply to these f type oflevel switches. i Suppression Pool 15' - 5" above Reference 2 table 3.2 B 1 These levelswitches i High Water Level bottom ofchamber Reference 6 are not calibratable. Reference 4 - table 3.121 Setpoint calculations j do not apply to these , type oflevel switches. D 101 ! v.nnoo sm .tx. yen.,m a. - e l
3 4 1 HIGH PRESSURE COOLANT INJECTION (HPCI) a j Earameter Value Parameter Reference Calculation Reference i j Alotor Termina! Voltagefor Active Components During injection Stade (Reference 14) ] 23013 (to open) ! Unit 2 186 Vdc No Design Reference Reference 22 l 1 ] Reference 23 Reference 24 l Discrepancy 6 I Unit 3 18$ Vdc No Design Reference Reference 22 ! Rcference 23
. Reference 24
] Reference 25 3 Discrepancy 6 Discrepancy 7 l 2 230l-8 (to open) i j Unit 2 166 Vdc No Design Reference Reference 22 l 5 Reference 26 ! Reference 24 Discrepancy 6 l Unit 3 161 Vdc No Design Reference Reference 22 l
- Reference 26 i i Reference 24 Discrepancy 6
{ 2301 STPV (to open) j (solenoid SV 8 energize) ! Unit 2 No Design Reference Discrepancy 10 Unit 3 No Design Reference Discrepancy 10 230123 (de-energize) The solenoids for these air operated valves (AOVs) all de energize upon i 230129 (de energize) HPCI initiation. Therefore, no motor terminal voltage requirements apply
- 2301 30 (de-energize) even though these are active valves.
. 230164 (de energize) l 230165 (de-energize) (solenoids) { -l i 1 1 i
+
i D 102 1 v.no s- ..< rn e % o 1 i
. . . . .. . . . . - . . - - - . -~.. ._.- -..--_-. -
.- . . ~ . . . . --. - -
i a l j HIGH PRESSURE COOLANT INJECTION (IIPCI) i
; Parameter Value Parameter Reference Calculation Reference Afotor Terminal Voltagefor Active Components During injection Afode (Reference 14)(con't) 2301 AOP(to start) i Unit 2 191.0 Vdc No Design Reference Reference 33 i Unit 3 191.0 Vdc No Design Reference Reference 33 i
R:ference 33 provides j the armature voltage of i the terminals of the loads fed by the 250 . Vdc batteries. 'Ihis
- calculation gives the i
voltage value for time l 1 periods after the initiation point of the i 250 Vdc system. The .t time period where the i lowest voltage is f available is from 0-1 j minute after system j initiation for the 4 Auxiliary Oil Pump (AOP). Alltime ! periods after 1 minute i have available voltages . of greater than 207
- volts. Per Reference 33, j the AOP is designed to
! operate with voltages as I low as 207 volts. l Therefore, operation of this motor is satisfactory for all time periods except during from 01 minute after initiation. However, per Reference 33, this motor is able to withstand a short time overload of 50% of rate armature current for 1 minute. Therefore, undervoltages due to the starting of blocks of motors during the first minute is not a concem. D-103 Verdicanon Screeneng of Key P a uneters, Revision 0
HIGH PRESSURE COOLANT INJECTION (HPCD Parameter Value Parameter Reference Calculation Reference
&fotor Terminal Voltagefor Active Components During injection Afode (Reference 14) (con't) 2320-GSEF (to start)
Unit 2 197.9 Vdc No Design Reference Reference 33 Unit 3 201.4 Vdc No Design Referer.ce Reference 33 Reference 33 provides the armature voltage of the terminals of the loads fed by the 250 Vdc batteries. This calculation gives the voltage value for time periods after the initiation point of the 250 Vdc system. The time period where the lowest voltage is available is from 01 minute after system ) initiation for the Gland Seal Exhaust Fan (GSEF). All time periods aner 1 minute i have available voltages of greater than 207 volts. Per Reference 33, the GSEF is designed to operate with voltages as low as 207 volts. Therefore, operation of this motor is satisfactory for all time periods except during from 0-1 minute after initiation. However, per Reference 33, this motor is able to withstand a short time overload of 50% of rate armature current for 1 minute. Therefore, undervoltages due to the starting of blocks of motors during the first minute is not a concem. D-104 Veri 6sanon $creening a(Key Parameters. Revision 0
_ _ . . _ _ _ _ _ . .._..--__.___.._.___.__..__._.__.__._._.._._-__..-m, s.__ l l l HIGH PRESSURE COOLAhT INJECTION (HPCD Parameter - Value { Parameter Reference Calculation Reference ! Motor Terminal Voltagefor Active Components During Auto isolation Mode (Reference 14) 2301-4 (to close)
- Unit 2 394 Vac No Design Reference Reference 22
- Reference 23 j Unit 3 390 Vac No Design Reference Reference 22 Reference 23 a
i 23015 (to close) Unit 2 173 Vdc No Design Reference Reference 22 l{ . Reference 23 Reference 24 j Discrepancy 6 i 8 Unit 3 171 Vdc No Design Reference Reference 22
- j. Reference 23 j Reference 24
, Discrepancy 6 230135 (to close) Unit 2 178 Vdc No Design Reference iteference 22
- . Reference 29
, Reference 24
- Dist.repancy 6 1
Di<crepancy 8 Unit 3 188 Vdc No Design Reference Reference 22
- Reference 29 Reference 24 l Discrepancy 6 230136 (to close)
Unit 2 175 Vdc No Design Reference Reference 22 l Reference 29 i- Reference 24 { Discrepancy 6
- Unit 3 165 Vdc No Design Reference Reference 22 Reference 29
- Reference 24 j Discrepancy 6 i
j Motor Terminal Voltagefor Active Components During Switch Overfrom Condensate Storage Tank (CST) to Suppression Poolfor Pump Suction (Reference 14)
- i 2301-6 (to close) i Unit 2 175 Vdc No Design Reference Reference 22 i Reference 28
- Reference 24 Discrepancy 6 Unit 3 187 Vdc No Design Reference Reference 22 Reference 28 Reference 24 1
Discrepancy 6 I J t
- D-105 v.nnc so ..ex,y % ne o 1
i .., . . - - - _ _
- . _ _ . _ _ _ . - _ _ . _ _ - . . . _ - . . - . ._m . . = _ . __ _. . . . . _ _ . _ . _ . . -__
J HIGH PRESSURE COOLANT INJECTION (HPCI) Parameter yA Parameter Reference Calculation Reference 230135 (to open)
- Unit 2 178 Vdc No Design Reference Refennce 22 Refertace 29 j Refereace 24 Discrepmcy 6
- Discrepa.,cy 8 i Unit 3 188 Vdc No Design Reference Reference 22
- Reference 29 Reference 24 Discrepancy 6 l
230136 (to open) Unit 2 175 Vdc No Design Reference Referenca 22 Reference 29 Reference 24
- Discrepancy 6
- Unit 3 165 Vdc No Design Reference Reference 22 Reference 29 Reference 24 Discrepancy 6 l Motor Terminal Voltagefor Active Components During Bypass Flow (Reference 14) 2301-14 (to close)
Unit 2 172 Vdc No Design Reference Reference 22 Reference 27 3 Reference 24 Reference 32 " Discrepancy 6 Unit 3 162 Vdc No Design Reference Reference 22 Reference 27 Reference 24 Reference 32 l Discrepancy 6 2301-14 (to open) i Unit 2 172 Vdc No Design Reference Reference 22 Reference 27 Reference 24 Reference 32 Discrepanty 6 Discrepancy 9 Unit 3 162 Vdc No Design Reference Reference 22 Reference 27
. Reference 24 Reference 32 i Discrepancy 6 f
Discicpancy 9 h D-106 Venfwanon Scremnas of Key P _ ., Revueen 0
l l HIGH PRESSURE COOLANT INJECTION (HPCI) Potential Discrepancies
- 1. The lower pressure range in which the HPCI system is required to deliver 5,000 gpm differs between section H.I.a(l) of Reference 12 and section 4.5.A.3.b.I of Reference 2. A Performance Improvement 1 Form (PIF) has been generated to address the issue. It is believed the correct pressure range is from 150 to 350 psig.
- 2. Adequate NPSH is demonstrated for the HPCI pump when aligned to the Condensate Storage Tank l (CST) with significant margin by Reference i 1. However, this document is an " evaluation", and a l formal calculation could not be located. Reference 21 calculated the maximum static pressure l available to the HPCI pump suction when aligned to the CST, However, the conditions used in this I cal:ulation should not be used when determining NPSH available for the HPCI pump. Furthermore, l
this calculation only determines static pressure, which is just one part of determining NPSH. Although I the data in Reference 11 appears to be correct, a formal calculation needs to be perfonned to provide the appropriate basis for pump NPSH when aligned to the CST. A PIF has been generated to address this issue.
- 3. No formal calculation could be identified that provides a basis for the NPSH available to the HPCI pump when aligned to the torus. A value is given in section 6.3.3.4.3 of Reference 1, the section that provides the analysis of determining the NPSH available to Emergency Core Cooling System (ECCS) l pumps. However, this analysis applies to the Low Pressure Coolant injection (LPCI) and Core Spray i pumps. Reference 21 calculated the maximum static pressure available to the HPCI pump suction when aligned to the torus. However, the conditions used in this calculation should not be used when l determining NPSH available for the HPCI pump. Furthermore, this calculation only determines static i pressure, which isjust one part of determining NPSH. An analysis similar to the one performed in Reference 1, documented in a formal calculation, needs to be performed to ensure adequate NPSH available to the HPCI pump when aligned to the suppression pool. A PIF has been generated to address this issue.
1
- 4. No calculation could be identified that provid:s a basis for the mass flowrate and steam pressure at the turbine inlet for the pressure range specified. Per Reference 6 and 7, the estimated friction line loss from the reactor vessel to the turbine inlet is 10 psi. However, there is no formal calculation that validates that number. A formal calculation needs to be performed to ensure that the reactor pressure vessel can provide steam to the turbine at the design conditions. A P!F has been penerated to address this issue.
- 5. The setpoint for the HPCI system isolation on low reactor vessel pressure is identified in References I, 6,10,19 as 100 psig. However, Reference 2 identifies the setpoint as 80 psig. A letter from General Electric (GE) was provided in Reference 19 stating that the correct setpoint is 100 psig. Reference 19 also states that Reference 2 should be revised to reflect the correct setpoint of 100 psig. A Nuclear Tracking System (NTS) number is tracking this correction.
1 D-107 Wnfwanon Screenmg of Key Paramates, Rewmon 0
IIIGli PRESSURE COOLANT INJECTION (IIPCI)
- 6. Reference 22 calculates the motor tecminal voltages for the HPCI system 2301 motor operated valves (MOVs). Reference 24 calculates motor terminal voltages for MOVs that are powered by the 250 Vdc system. The MOVs that are listed in both documents are the HPCI (2301) system MOVs that are powered by the 250 Vdc system. These valves are 2(3)-2301-3,5,6,8,9,10,14,15,35,36,48, and 49 For all of these valves, the value for the motor terminal voltage is lower in Reference 24 than the value in Reference 22. The reason for this appears to be the added conservatism used in Reference 24, as the motor terminal voltages calculated are determined to be " worst case". However, References 23, 26,27,28, and 29 used the motor terminal voltage value from Reference 22 to perform its thrust calculations. If the values calculated in Reference 24 are to be used for design input with other calculations, then Reference 22 should be revised to ensure that the motor terminal voltage values for the above valves are obtained from Reference 24. In addition, References 23,26,27,28, and 29 should be revised, or a new thrust calculation should be performed to supersede References 23,26,27, 28, and 29; to ensure that the motor terminal voltage produces the required torque to actuate the valve.
A PIF has been generated to address this issue.
- 7. The minimum required thrust to close (and open) the 3-2301 3 valve is greater in Reference 23 than Reference 25. This is most likely due to the calculation methodology and the added conservatism in Reference 23. The maximum available thrust is the same in both calculations and exceeds the minimum required thrust in both cases. However, Reference 25 should be revised so that Reference 23 is used as design input in the future to ensure design control. A PIF has been generated to address this issue.
- 8. The maximum available thrust to close and open the 2-2301-35 valve differs between Reference 29 and Reference 30. It is unclear at this point which value is correct or more conservative. The minimum required thrust is the same in both calculations and is exceeded by the maximum available thrust in both cases. However, the correct maximum available thrust should be determined and the necessary documents revised to ensure one value is used in the future. A PIF has been generated to address this issue.
- 9. The motor terminal voltage at valve 2(3)-2301-14 differs between References 22 and 32. Reference 32 was an older calculation that was prepared to determine the voltage drop from the battery to the valve terminals. This methodology is different from the one used in Reference 22, and is not as conservative. Also, the motor terminal voltage from Reference 32 was not used for the valve thrust
- calculations. Reference 12 should be revised or superseded to ensure design control. A PIF has been generated to address this issue.
- 10. The turbine stop valve solenoid valve SV-8 is powered by the 125 Vdc batteries. No calculation could be found that determines the minimum voltage available at the solenoid valve terminals, or if this voltage is adequate to properly operate the valve. A voltage drop calculation for both Unit 2 and 3 needs to be performed for the 125 Vdc system, similar to the one prepared for the 250 Vdc system in Reference 33, to ensure adequate voltage is available at the turbine stop valve solenoid valve terminals. A PIF has been generated to address this issue.
J
- 11. Dresden Calculation DRE96-0210."HPCI Pump Margins with Respect to Appendix K Curve," was performed and demonstrates that the HPCI system can deliver flow to the reactor vesselin excess of what is required in the Appendix K LOCA analysis as specified in References 9 and 16. However, this calculation has not been completed and signed prepared, reviewed, and approved. This calculation needs to be completed and signed to ensure a formal calculation exists that can provide a basis for HPCI design flowrate. A PIF has been generated to address this issue.
D-108 Wn6canoa kroening of Key Parwneien, Revision 0
HIGH PRESSURE COOLANT INJECTION (IIPCI) References
- 1. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 01 A.
- 2. Dresden Station Unit 2 and 3 Technical Specifications Upgrade (TSUP), January 13,1997,
- 3. Dresden Station Unit 2 and 3 Technical Specification Upgrade Basis (TSUP Basis), January 13,1997.
- 4. Dresden Administrative Technical Requirements (DATR), September,1996.
- 5. Dresden Technical Specifications, December,1996.
- 6. GE Specification 257HA353 AB, "HPCI System Data Sheet," Revision 3, dated 10/21/69.
- 7. GE Drawing 729E299,"HPCI System Process Flow Diagram," Revision 4, dated 2/17/67.
- 8. GE Specification 257HA353,"HPCI System Design Specification," Revision 5, dated 10/21/69.
- 9. Nuclear Fuel Services (NFS) Safety Analysis Input EMF-93-176, Revision 2," Updated Principal LOCA Analysis Parameters for Dresden Units 2 and 3", dated September,1996.
- 10. Dresden Rebaseline Updated Final Safety Analysis Report (UFSAR), Revision 02, with approved change DFL# 96070.
I 1. System Based Instrumentation and C ontrol Inspection (SBICI) Tracking Number T-079,"Dresden HPCI NPSH Evaluation - CST Suction," CHRON #211450, dated 8/30/94.
- 12. Dresden Operating Surveillance (DOS) Procedure 2300-03, "HPCI System Operability Verification,"
Revision 40, dated 12/19/96.
- 13. Byron Jackson Pump Curve Number T-29649-3, HPCI Booster Pump Serial Number 671-S-0998, Model Number 12xl4x23 DVS, dated 5/21/68.
- 14. GE Drawing 728E911, Sheets 1,2, and 3,"HPCI System Functional Control Diagram," Revision 10, dated 7/13/72.
15 Fire Protection Report, Volume 2," Appendix R Non-Conformance (Sections III.G, III.J, III.L), Safe Shutdown Report.
- 16. Nuclear Fuel Services (NFS) Safety Analysis Input EMF-89-065, Revision 3,"Dresden Units 2 and 3 Principal LOCA Analysis Parameters", dated July,1995.
List of Calculations 17 Dresden Calculation NED-I-EIC-0108,"High Pressure Coolant Injection Turbine and Pump Area Temperature Switch Setpoint Error Analysis," Revision 1, dated i1/07/96.
- 18. Dresden Calculation NED-I-EIC-0109,"High Pressure Coolant Injection (HPCI) Pump Discharge Flow Loop Accuracy and Minimum Flow Setpoints," Revision 2, dated 3/26/96.
D-109 Venfication $creenmg o(Key Parvnerars, Revisson 0
i l HIGH PRESSURE COOLANT INJECTION (HPCI) 19 Dresden Calculation NED-I-EIC-0110,"High Pressure Coolant Injection Low Reactor Pressure Isolation Error and Serpoint Analysis," Revision 3, dated i1/02/96. 20 Dresden Calculation NED-I-EIC-Olli,"High Pressure Coolant Injection (HPCI) Steam Line High Flow Isolation Setpoint Error Analysis," Revision 3, dated 11/4/96. I i
- 21. Dresden Calculation 87-981/ 982,"HPCI Static Pressure and Total Developed Pressure," Revision I, dated 12/28/87.
- 22. Dresden Calculation NED-EIC-MOV-DR-0007," Valve Actuator Motor Terminal Voltage Calculation," Revision 0, dated i1/21/94.
- 23. Dresden Calculation 004 MN-311," Thrust Calculation for Functional Group HPCI 1", Revision 11, dated 6/22/94.
- 24. Dresden Calculation DRE96-0126," Motor Terminal Voltage Calculation for Dresden 250 VDC Motor {
Operated Valves," Revision 0, dated 7/22/96.
- 25. Dresden Calculation 004-MN-302," Thrust Window Calculations for MOVs 3-23013,3-2301-4,3- l 2301 5 & 3-2301 15," Revision 1, dated 10/16/91.
- 26. Dresden Calculation 004-MN-315," Thrust Windows for Dresden Function Group HPCI 3," Revision 4, dated 8/2/93. j i
- 27. Dresden Calculation 004 MN-316," Thrust Windows for Dresden Function Group HPCI-5," Revision !
2, dated 5/17/93. l
- 28. Dresden Calculation 004-MN-345," Thrust Windows for Dresden Function Group HPCI-2," Revision 1, dated 3/16/92.
- 29. Dresden Calculation 004-MN 349," Thrust Windows for Dresden Function Group HPCI-7," Revision 0, dated 3/10/92.
- 30. Dresden Calculation 004 M-058," Thrust Window Calculation for Motor Operated Valve Tag Number 2 2301 35," Revision 0, dated 8/22/90.
- 31. Dresden Calculation PMED-8982 30-01," Development of a Duty Cycle Based on a More Conservative Application of Coincident Starting Currents for 250 Vdc Battery System," Revision 11, dated 12/16/96.
- 32. Dresden Calculation 7328-00-191," Voltage Drop on 250 Vdc Motor Operated Valve," Revision 0, dated 1/20/88.
- 33. Dresden Calculation DRE96-0189,"250 Vdc Battery Voltage Drop Load Calculations," Revision 0, dated 11/1/96.
D-Il0 Vent'scation Screening of Key Parameters, Revision 0
i 1 1 DISCREPANCY
SUMMARY
1 ) Discrenancy Summary by System 1 I i Sofety-Related 12S Vdc System l . 1 Discreoancy #1 The UFSAR lists 1495 AH rating for NCX-21 lead calcium battery cells for the 125 V batteries .; which does not apply to them since their design basis "End of Discharge Voltage" is 1.81 volts
- instead of 1.75 volts per cell. Listing the battery manufacturer's nominal 8-hour,77 *F, rating of l 1495 AH for NCX 21 cells in the UFSAR and the system DBD is ambiguous and potentially
- misleading when the cells are not discharged to 1.75 Volts. For a higher (1.81V) end of discharge l voltage the cell capacity / usable rating will be less than 1495 AH.
i Discrenancy Resolution j The UFSAR and the DBD for the de system will be revised. Trackine Numbers l PIF #227A 12-1997-011977 (56) NTS #237-2019710701 1 Discreoancy #2 I , The following errors were identified in calculation 7056-00 19 5, Revision 31: i I. The assumption listed on page 2B (rev. 30) is no longer valid and should be deleted. t ] 2. On page 9, paragraph 2, the reference to 4 hour Station Blackout should be changed to one hour j and the referenced page number should be 269 (Rev. 8 summary) instead of 222. I 3. Page 32 (Rev. 30) needs to be revised to list the new SF6 breaker types and their current data
- for 4kV switchgear buses 23,24,33,34 to make it consistent with the data on page 40c. Also i the revision number on page 40c should be 31 instead of 24.
! 4. The current-time data shown on page 45 for Unit 2 ATWS panel feed (Circuit 13, Bus 2A-1) following modification M12 2-94-002 also applies to Circuit 10, Bus 28-1, listed on page 104 (Rev. 31). This should be made clear by adding a note on page 45, l
- 5. The load cycle period (end time) should also be shown on page 222.
- 6. The description of modification M12-2 94-002 (item 8, page 291C, Rev. 31) is not correct for the Unit 3 ATWS feed circuits as this modification is not yet implemented on Unit 3. The load j profile of 3 5 A from t=0 to 240 minutes applies to only Unit 2 circuits. For the Unit 3 ATWS
, circuits / feeds the stepped load profile (10.39/3.5 A) shown on page 134A, Rev. 31 is conservative and covers both the pre-mod and post mod loading. However, the Unit 2 j reference and modification number on page 134A needs to be corrected to show that it applies 1 to Unit 3 feeds.
J i
; E-1 Verifwasen screening of Key Parametert Ramos 0
Discrecancy Resolution The calculation will be revised. Tracking Numbers PIF #227A 121997-012563 NTS #237-140-97-02301 Discrenancy #3 The following discrepancies were identified in calculation 8684-41-19-1, Rev. 0: Part A of this calculation concerning the old FPS 23 battery cells is no longer relevant and it should be superseded in the next revision of this calculation. Parts B, C, and D of this calculation are not current since they do not utilize the present load data utilized in revision 31 ofcalculation 7056-00-19-5. Part C may be superseded by cross referencing to its updated 125 Vdc battery sizing calculation 7056-00-19-5, Revision 31 which validates adequacy of NCX 21 size cells. (The NCX-27 cells being about 30% !arger than the NCX-21 cells - 13 versus 10 positive plates - are enveloped by the adequacy of smaller, similar type and make NCX 21 cells). Parts B, D, F, and I of this calculation concerning the cable voltage drop and voltage profile may be incorporated in the current 125 Vdc Bus Voltage calculation 8982-66-19-1, (Present Rev. 5 of 8982-66-19 1 does not include feeds from the alternate batteries). Discrecancy Resolution PIF resolution will disposition this issue. Tracking Numbers The PIF addressing this discrepancy is being processed. Discrecancy #4 The following discrepancies was identified in calculation 8706-41 19 1, Rev.1: Parts B, C, and D of this calculation are not current since they do not utilize the present load data utilized in Rev. 31 of calculation 7056-00-19 5. Part C conceming sizing of the battery cells may be superseded by cross referencing it to the updated 125 Vdc battery sizing calculation 7056-00-19 5, rev. 31 which verifies adequacy of NCX 21 size cell. Parts B and D of this calculation concerning cable voltage drop and voltage profile may be incorporated in the current 125 Vdc Bus Voltage , calculation 8982-6619-1. (Present Rev. 5 of 8982-6619-1 does not include feeds from the ! alternate battenes. Discrecancy Resolution PIF resolution will disposition this issue. Trackine Numbers I The PIF addressing this discrepancy is being processed. i i E-2 l Wrifusoon screeneng e(Key Pernemers Revision 0
. . _ _ _ . . _ _ . - - _ _ _ _ _ . _ - . - _ . . - __.m.__ . _ _ . . . _ .- . - -- _ _ _ _. _- .
I Discreoancy N5 5
- The following discrepancy was identified in calculation 7056-00-19-5, Rev. 31
4 l This calculation does not address or verify adequacy of the 125 V batteries for the SBO condition with the current load data utilized in this revision. Discrenancy Resolution i j PlF resolution will disposition this issue. j Trackine Numbers i The PIF addressing this discrepancy is being processed. Discrecancy #6 i The following discrepancy was identified in calculation 5569-31-19-1, Rev. 2:
- This calculation determines the maximum fault current level at specific locations / buses in the 125 Vdc system when fed from the main batteries only, it would be more appropriate to include the attemate batteries also within its scope to make it more representative of the entire 125 Vdc system.
Discrenancy Resolution l ? PIF resolution will disposition this issue. Trackine Numbers l The PIF addressing this discrepancy is being processed. 4 I l Safety-Related 250 Vdc System I Discrecancy el Section 8.3.2.1.1 (pg. 8.3 20) of the UFSAR states that the 8-hour battery rating of 1495 ampere- ) hours is based on the lowest (minimum) expected electrolyte temperature of 65 'F, but TSUP , i Section 4.9.C.2.c lists a minimum of 60 *F average for all the connected cells in the battery. This implies that some cells may have less than 60 'F minimum electrolyte temperature. The battery { capacity / capability is based on a minimum 65 'F cell electrolyte temperature. Discrenancy Resolution - i The Technical Specifications with TSUP will be revised to clarify this issue, i
- Tracking Numbers e
i Plf #227A.121997-012037(46) I NTS #237 201-971200i f
)
1 1 E-3
%mfw4 ace Sr.reemag of Key W Reveos 0
i Discrecancy 82 i The 250 Vdc battery sizing calculation PMED 8982 30-01 does not address the station blackout j (SBO) requirements. An earlier 250 V battery sizing calculation,7056-00-19-4, Rev. 7, dated 5 22-i 91, addresses the 4 hour SBO in its Rev. 3 and the I hour SBO in its Rev. 5. This calculation is not l current and it does not reflect the current updated load data utilized in Rev.1I of Calc. PMED 8982 j 30-01. Also there is a duplication of Scope / Purpose concerning sizing of the 250 V battery cells j between these two calculation. I j_ Discrenancy Resolution i PIF resolution will disposition this issue. Tracking Numbers The PIF addressing this discrepancy is being processed. Low Pressure Coolant injection (LPCI) System Discrenancy #I The required Net Positive Suction Head (NPSH) for the LPCI pumps is identified in the LPCI DBD DBD DR 172A as 31 feet (for i pump operation - 5,000 gpm) and 37.5 feet (2 pump operation - 5,600 gpm). However, the LPCI pump curves identify the required NPSH to the pumps as 30 feet and 36 feet for 5,000 gpm and 5,600 gpm, respectively. Discrenancy Resolution i The DBD for the LPCI system will be revised. Trackine Numbers PIF #227A-12-1997-012047 (33) NTS #237 140-97-00801 Discrenancy #2 Dresden calculation DRE96-02 t l was performed to compare LPCI system resistance and pump c :;ves to the LOCA analysis for Unit 2. Attachment P, Figure I shows predicted 2 pump flow into the reactor vessel as a function of reactor vessel pressure and compares it to the flow used as input in the LCCA analysis. There is an error in this curve in that the minimum flow bypass was not accountea 'or in this curve. Discrecancy Resolution The calculation will be revised. Tracking Numbers P!F #227A-121997 011972 (81) OP EV #97 23 E-4 Wnficanoe S creeneag or Key Parameters Rev,sion 0
l l Discrecancy #3 l l l Calculation NED I EIC-0124 performed a setpoint error analysis for the LPCI minimum flow I bypass flow switch for the valve opening requirement or21,000 gpm. This calculation should be l revised to determine the flow ranges where the flow switch would close the minimum flow bypass i valve. This information would be used for calculation DRE96-02 t l, which calculated LPCI flow to the reactor vs. reactor pressure and compared that to what is required under the LOCA analysis. l Calculation DRE96-0211 did not include the minimum flow bypass in the calculation. Calculation l DRE96-021i should be revised to include the minimum flow bypass for the flow ranges specified in l calculation NED-I EIC-0124. l l' Discreoancy Resolution l The calculation will be revised to determine the necessary information. l Trackine Numbers An NTS number addressing this discrepancy is being processed. Discrecancy H4 l An evaluation performed by Nuclear Fuel Services (NFS), NFS:BSA: 96-140, specifies that 5,000 l gpm flow is required by the LPCI system through the containment cooling heat exchanger and to the l containment to ensure adequate containment cooling. However, no formal hydraulic calculations exist that demonstrate the LPCI system can provide this flow to ensure adequate cooling in the suppression pool cooling or containment spray mode. l l Discrepancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Trac':ine Numbers l PIF #227A 121997-012187(64) NTS #237140-97-01001 l Discrecancy #5 Section 6.2 and 6.3 of the UFSAR addresses containment response after a LOCA, and analyzes four containment cooling cases. However, no calculations exist that provides the basis for these scenarios. This section of the UFSAR should be revised to address those cases which are important from a licensing perspective (i.e., I LPCI pump / 2 CCSW pumps) and perform the necessary calculations to provide the basis for these scenarios. Discrepancy Resolution ne UFSAR will be revised and the remaining cases will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers l Pending UFSAR change DFL #97011 l E-S Veriraos Screenmg of Key Parametert Rethon 0
Discrecancy #6 , The LPCI system flow requirement for injection to the reactor vessel that is used in the LOCA analysis is 9,000 gpm at 20 psid. However, section 4.5 of the TSUP and Dresden Operating Surveillance (DOS) 1500-05 test for LPCI flow of 14,500 gpm. The TSUP and DOS 1500 05 should be evaluated for revision to test LPCI flow of 9,000 gpm as required by the LOCA analysis. Discrecancy Resolution The TSUP and surveillance procedure will be evaluated for change. Trackine Numbers Pif #227A-12-1997 012184(44) NTS #237-140-97-01801 Discrecancy #7 Dresden calculation DRE96-0211 demonstrates that the LPCI system can deliver flow to the reactor vessel in excess of the flow required by the Appendix K LOCA analysis. However, this calculation was only prepared for Unit 2. A similar calculation needs to be :ompleted for Unit 3. Discrecancy Re ch11gn . l This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis ! Calculations project. Tracking Numbers 4 PIF #227A-12-1997-012194(37) l NTS #249-140-97-01301 l Discreoancy #8 l l The available motor terminal voltage for 1501 series motor operated valves (MOVs) was calculated under calculation NED-EIC-MOV-DR-0001. For three valves, the thrust calculations used a motor l terminal voltage value greater than the motor terminal voltage available under calculation NED-EIC-MOV-DR-0001. These valves are 2 1501 13B (thrust calculation 004-MN 320),2 1501 19B i (for proposed modification scenario) (thrust calculation 004-MN 323), and 3 1501-27B (thrust I calculation 004-MN 331). l Discrecancy Resolution PIF resolution will disposition this issue. Tracking Numbers l l The PIF addressing this discrepancy is being processed. Discrecancy #9 l No calculations could be found that demonstrate there is sufficient motor terminal voltage for the LPCI pumps so the pumps can perform adequately. E-6 Venfwance Screensag of Key Parameters, Revisma 0
1 Discrecancy Resolution PlF resolution will disposition this issue. Tracking Numbers The PlF addresting this discrepancy is being processed. Discrecancy #10 Section 7.3.1.2.2 of the UFSAR and the LPCI system lesson plan state that a reactor vessel pressure permissive will delay the LPCI loop selection logic initiation until reactor pressure has dropped to a value less than or equal to 900 psig to allow for coastdown of any recirculation pump which has just been tripped. However, Dresden calculation NED-I-EIC-0114 and Dresden Instrument Surveillance (DIS) 1500-07 state that the instrument setpoint is 944 psig, and the setpoint error analysis was performed using an acceptance criteria of greater than or equal to 900 psig. Disoeoanc(11notution PIF resolution will disposition this issue. , Trackinq Numbers The PIF addressing this discrepancy is being processed. Discrecancy fit i Thrust calculation,004-MN 309, Rev. 7, dated 4/14/93, performed a thrust analysis for the Reactor Recirculation Discharge Valve 2-0202-5A. However, the motor terminal voltage used in this calculation is greater than the motor terminal voltage available as determined in calculation DRE96-0010, Rev. O, dated I/16/96. Discrecancy Resolution PIF resolution will disposition this issue. Trackinc Numbers The PIF addressing this discrepancy is being processed. Discrecancy #12 Dresden Instrument Surveillence (DIS) 1500-13 and 1500-14 used a milliamp value of 9.097 mA for 1155 gpm. However, the correct value should be 9.907 mA for 1155 gpm as stated in calculation NED-I-EIC-0124. Discrecancy Resolution The procedure will be revised. Trackinz Numbers PIF# 227A-12-1997-012189(74) NTS# 237140-97-01501 E7 v.nc- sm g or n.y r.n n n.m o
._- .-... - .-.-. - - - . - ~- =.
Containment Cooling SmIce li'ater (CCSFI) System Discrecancy #I Section 3.19.2 of the Dresden Technical Administrative Requirements (DATR) identifies the minimum water level in cribhouse CCSW suction bay of 499' - 0". However, section 3.8.C.l of the TSUP identifies the minimum water level as 500'- 0" Furthermore, the UFSAR references drawing M-10 which states the normal operating low water level in the suction bay is 501' - 0", and this value was used in calculation DRE96-0214. Discreoancy Resolution The calculation and DATR will be revised. Trackine Numbers PlF #227A-12-1997-012048 (32) NTS #237 140-97-00901 Discrep.ansyl! l No formal calculations exist that demonstrate adequate Net Positive Suction Head (NPSH) is I available to the CCSW pumps. Discrenancy Resolution i This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers P!F #227A-12 1997-012040 (79) NTS #237-140-97-00301 Discrecancy #3 Section 9.2.1.3 of the UFSAR specifies the minimum requirements for containment cooling include 2 CCSW pumps. However, section 9.2.5.3.2 indicate that only 1 CCSW pump will be available during a postulated dam failure. Currently there are no containment cooling or heat exchanger differential pressure calculations for a 1 CCSW pump scenario. Discrecancy Resolution PIF resolution will disposition this issue. Trackine Numbers The PIF addressing this discrepancy is being processed. E-8 Veticanon Screening of Key Parmem Revisaan 0
Discrecancy #4 No calculations could be found that demonstrate there is sufficient motor terminal voltage for the CCSW pumps so the pumps can perform adequately. In addition, no calculations have been found that show the CCSW pump motors are matched for power and speed requirements to the CCSW pumps. Discrecancy Resolution PIF resolution will disposition this issue. Tracking Numbers The PlF addressing this discrepancy is bei g processed. Discrecanev #5 1 There is a requirement that the CCSW side of the containment cooling heat exchanger be maintained at 20 psi above the LPCI side of the heat exchanger to preclude leakage from LPCI to CCSW. A report was generated by F. J. Mollerus and C. B. Johnson that shows why maintaining this pressure differential is adequate. This report should be formalized into a calculation. Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Trackine Numbers An NTS item addressing this discrepancy is being processed. Feedwater/ Condensate System Discrecancy #1 A number of key parameters that are used as input for the transient analysis, EMF-95-043, do not have a formal calculation to provide a basis for their values. A computer heat balance model is used to verify various operating parameters of the Feedwater/ Condensate system that is used while the plant is in operation. This model may provide the basis for some values in the transient analysis. l Discrenancy Resolution ' PIF resol
- ion will disposition this issue.
Trackine Numbers The PIF addressing this discrepancy is being processed. 1 E9 ww so-, .n., % :oi o l l
3 Turbine Building Closed Cooling IVater (TBCCIF) System Discrepancv # I No formal calculations could be found that provides a basis for the heat transfer rate of the TBCCW heat exchangers. Discreoancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers FIF #227A 12-1997-012041(78) NTS #237 140-97-00401 Discreoancy if 2 No formal calculations could be found that provides a basis for the flowrate through the TBCCW system. Discrecancy Resolution I This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis ! Calculations project. Trackine Numbers I h PIF #227A-12-1997-012035 (5 l} NTS #237-140-97-00101 i Main Steant Safety and Relief Valves i i Discrecancy #1 No calculation could be found that determines the minimum voltage at the solenoid terminals for the l target rock and electromatic relief valves, or if this voltage is adequate to operate the valve. A voltage d op calculation for both Units 2 & 3 needs to be performed for the 125 VDC system. Discrepancy Resolution PIF resolution will disposition this issue,
)
i Trackine Numbers 1
)
The PIF addressing this discrepancy is being processed. Discrecancy #2 The SRV valves have a nominal 10 second timer to inhibit reactivation to allow the reflood level in i the SRV discharge line to drop to normal level. The required time given in the UFSAR was 6.3 l seconds. The backup design calculation demonstrating that this value is conservative for all SRV j discharge lines needs to be either recovered or reconstituted. l l l E-10 Venfw.auon 5asening of Key Paramete1, Reisson 0
._. _ . _ _ _ - . ___..-._ __ _, .._..__.m_ mm...____ _, ... . . - . - . . _ . .
l l 1 h j Discrecancy Resolution j-PIF resolution will disposition this issue. l { Tracking Numbers
- 1 4 1 j The PIF addressing this discrepancy is being processed.
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i Service Water (SH) System \ i i NONE 2 l Automatic Depressurl:ation System (ADS) i Discrenancy #1 The old Technical Specifications have a different pressure value for LPCI pump discharge pressure for initiation of ADS, than the TSUP, table 3.2.B-l. The old Tech Specs have a pressure range of 250 & $100 psig and the TSUPs have a pressure range of 2100 & $150 psig. Dresden Units 2 and 3 Principal LOCA Analysis Parameters, EMF-89-065, Rev. 3, July 1995 reference the old Tech Spec Table. Discreoancy Resolution l The subject documents will be reviewed and the appropriate documents revised. i I Tracking Numbers i PIF# 227A-12-1997 012186(75) NTS# 237140 97-01601 4 k VSafety Related Auxiliary Power System Discrecancy #1 Table 8.3-1 of the UFSAR lists the feeder circuit breakers on 4 kV buses 23 1 and 24 1 as having ratings of 350 MVA,1200A. This circuit breaker rating is inconsistent with the purchase specification rating of the subject circuit breakers. 'lhe purchase specification rating of the circuit breakers is 250 MVA,1200A. Additionally, these circuit breakers have been analyzed in the short circuit calculations with a rating of 250 MVA,1200A. Discrenancy Resolution The UFSAR will be revised. Tracking Numbers PIF #227A-12 1997-012427(37) E-1I Wnriquea $creamns of Kay Pm Revmoa 0
i I L Discrecancy #2
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The Dresden 4 kV Bus Overvoltage DOA 6500 11, rev. 0 may not be complete. This procedure is intended to be utilized in order to mitigate overvoltage conditions on the 4 kV safety related bus 23 1 (33-1) and 24-1 (34 1). This procedure may be incomplete since it does not address the voltage levels on the upstream SWGR 23 (33) and 24 (34) which also feed safety-related 4 kV Containment i 4 Cooling Service Water (CCSW) Pumps A, B, C, and D. He procedure does not address the ! upstream bus voltage on which an overvoltage condition could exist and cause safety-related motor ! winding damage. He setpoint voltage at which an overvoltage condition is alarmed is 4.3 kV, this is a computer alarm point. A cursory review of ELMS calculation 7317-43 19 1 resulted in a typical ! ] voltage drop between SWGR 23 and 23 1 to be approximately 10 Vac under summer loading i conditions, this is suggestive of minimal voltage drop between the buses. In the subject procedure,
- the relationship between adding sys
- em load and the subsequent bus voltage reduction is not clear and is not entirely consistent with reference 1.0 which determines minimum system KVA load to
! { preclude 4 kV bus overvoltage conditions. It is understood that the intent of the referenced i calculation is only to recommend a minimum system KVA load which will limit overvoltage f conditions. l The long term effects of motor overvoltages could accelerate equipment aging and should be l addressed by 4 kV voltage monitoring which includes all safety-related buses. He upstream voltage at SWGR 23 (33) and 24 (34) can be monitored in the control room. This potential problem could be remedied by revising the procedure to have the operator monitor the bus voltage at the upstream SWGR if an alarmed overvoltage condition exists. An alarmed overvoltage condition is defined as a 4 kV bus 23 1 (33-1) or 24-1 (34-1) voltage greater than or equal to 4300 Vac.
-4 Reference 1.0 Calculation 9198-6919-1, rev. O, dated 02/25/94, Minimum Auxiliary System Loading to Prevent Safety.Related Overvoltages.
Discrenancy Resolution PIF resolution will disposition this issue. l Trackine Numbers 1 The PIF addressing this discrepancy is being processed. 480 Vac Safety Related Auxiliary Power System NONE Isolation Condenser System (including Makeup Water) Discrecancy #1 No formal calculations have been found that provide a basis for the heat removal rate of the isolation ! condenser heat removal rate as stated in the UFSAR. I Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. E-12 Venkanos Scresung or Key Parammes Revie=on 0
1 Tracking Numbers PIF #227A 121997 012046(49) NTS #237-140-97-00701 OP EV #97-19 I 1 l Discreoanev #2 i i The TSUP requires that for initiation of the isolation condenser requires a sustained high pressure I signal of 1070 psig for 17 seconds. No formal calculations have been found that provide a basis for j the initiation of this time delay. Discrenancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers j PIF #227A-12-1997-012045 (50) i NTS #237-140-97-0060I OP EV #97 20 l t Discrecancy #3 It is believed that calculation BSA D-95-07 should te used to establish the design basis makeup flowrate of 430 gpm based on NUREG-0800 or 361 gpm based on ANSI 5.1. The isolation system notebook and calculation 282 Y-M 03 should S revised to reflect these makeup flowrates, f Discrenancy Resolution 1 The calculation and system notebook will be revised. j Tracking Numben I PIF #227A-121997 012179(46) NTS #237140-97-0190I Discrecancy #4 1 Currently there is no design basis value for onsite storage of water for continued operation of the i' isolation condenser system. Onsite storage requirements were given in the original FSAR, but were subsequently removed. Initial fire protection ana'ysis required 8 hour onsite storage, but this was later revised to 2 hours. Onsite storage supplies needs to be established and calculations should be performed to provide the basis for this value. Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis j Calculations project. l l l l E 13 [ Ventkanon Screening of Key Pm Ravason 0 i i
Trackine Numbers PIF# 227A 121997-012354(17) NTS# 237140 97-02101 Discrecancy #5 The following three errors were identified in the isolation condenser system DBD-DR-I71:
- 1. Dresden isolation condenser is not designed to accommodate a complete loss of generator load or turbine trip without scram as stated on page 3-9.
- 2. Decay ratio is not an appropriate value of controlling parameters used as reference bounds for the design of the isolation condenser system as stated on page 4-1.
- 3. References 3.004 and 3.005 are Dresden Unit I references.
Discrecancy Resolution The DBD will be revised. Trackine Numbers PIF #227A-12 1997-012042 (77) NTS #237-140-97-00501 Discreoancy #6 Calculation BSA-D-95-07 provides a good basis for the isolation condenser water level and temperature requirements. This document should supersede RSA-D-93-06, which provided an initial basis before BSA D-95-07 was completed. Discrenancy Resolution The calculation will be revised. Tracking Numbers PIF #227A 121997 012185(38) NTS #237140-97-01701 Discreoancy 87 Several Dresden documents discuss makeup water to the isolation condenser for a single unit case. However, there are accident scenarios where both units would be atTected simultaneously and the isolation condenser would be used concurrently for both units. These documents need to be xvised to address dual unit operation of the isolation condenser system. Discrecancy Resolution The subject documents will be revised. E-14 Venficauon Screenmg of Key Paramecen Rennes 0
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Trackin9 Numbers PIF #227A 12-1997-012355(30) NTS# 237140-97-02001 Discrenancy #8 The NPSH calculations for the isolation condenser diesel driven makeup pumps (Reference 1) were performed at the rated flow conditions for single pump operation. For these conditions, the minimum NPSH margin is 0.6 fL Reference 3 provides the basis that the diesel driven pumps can deliver their rated flow. This calculation used conservative pressure drop assumptions, such as assuming the design pressure in the isolation condenser and a 10% increase in system resistance. If these conservatisms are removed, the pumps can deliver flow greater than its design flow, and thereby require greater NPSH. Reference i should be revised to address NPSH requirements for potential maximum flow conditions. The NPSH calculations for the clean demineralizer water pumps (Reference 2) were performed for a variety of conditions to determine the effect of changing pump suction piping sizes. For the current piping configuration, there is no calculated NPSH margin at rated conditions for single pump operation. In addition, NPSH for maximum flow conditions was not evaluated. Reference
- 1. Bechtel Calculation DR-150-M 001, " Isolation Condenser Suction Line Reroute, Determination l of NPSHA for New Line Routing."
- 2. NUS Calculation 282Y-M-01, Demineralizer Makeup to the isolation Condenser.
- 3. Stone and Webster Calculation Ol849.00-M(Cl)-001, " Conceptual Pressure Drop for Isolation Condenser Diesel Driven Makup System."
Discrecancy Resolution PIF resolution will disposition this issue. Trackine Numbers The PIF addressing this discrepancy is being processed. ! Discrepancy #9 4 I The minimum operator terminal voltage (References I and 2) for MOVs 2(3)-1301-3 and for MOVs 3 2(3)-4399-74 are less than was assumed in the trust adequacy calculations (References 3 and 4). 1 The valve opening calculations demonstrate that there is a substantial margin relative to the required ] thrust. Therefore, there is no operability concern because these valves should be able to open at the j reduced voltage, as required, to initiate the isolation condenser system. In addition, References I and 2 are inconsistent with respect to calculated motor terminal voltage. The controlling document appears to be Reference I due to added conservatisms that was included in this calculation. Reference 2 should be revised accordingly to ensure design control. D E-15 Venfwmoa Susensag of Kay P. ., Rawsum 0
J References
- 1. DRE-96-0126, biotor Terminal Voltage Calculation for Dresden 250 Vdc Motor Operated Valves, Revision 0.
- 2. NED EIC-htOV DR-0005, Valve Actuator hiotor Terminal Voltage Calculation, Rev. 0.
- 3. Bechtel Calculation 004 htN-305, Thrust Window Calculations 1301 System Valves.
- 4. Bechtel Calculation DR 265 ht-001, "MOV Differential Pressure Thrust Window 2(3) 4399 74",Rev.I Discreoancy Resolution PIF resolution will disposition this issue.
Trackin9 Numbers The PIF addressing this discrepancy is being processed. Discrecancy # 10 These following parameter references and calculation discreoancies were identified:
- a. A system performance calculation for the condensate trensfer pumps 2(3)-3319A & B.
- b. A system performance calculation for the condensate tran ferjockey pump 2(3)-3320.
- c. A system performance calculation for the portable gasoline driven pump.
Discrecancy Resolution l PIF resolution will disposition this issue. l Trackine Numbers The PIF addressing this discrepancy is being processed. Discrecancy #11 The pressure switch settings process for the isolation condenser initiation uay contain an error in the treatment of the head correction term. The technical specifications are esteblished for steam dome i pressure. In the pressure switch setting calculation and procedures (References I and 2), it is established that there is a significant water head at the pressure switch. This results in a significant reduction in the isolation condenser system initiation pressure relative to the steam dome pressure. l As a result, the current instrument settings procedures cculd increase the potential for spurious i system initiation when the total instrument setting uncertainties and margin are taken into account. l Using the correct instrument setting procedures and calculations, the initiation pressure could be I between 1033 and 1069 psig. This potential inconsistency should be evaluated to determine the acceptability of the instrument setpoint. More specifically, it needs to be determined if an isolation condenser initiation at a pressure less than the pressure scram setpoint is acceptable. Further, the pressure switch setting process needs to be incorporated into the isolation condenser performance calculations to assure that the analysis covers the instrument uncertainties. E 16 Venfwance Scnening of Key L- Revis,oe 0
I l P l References
- 1. NED I EIC 120, Sustained High Reactor Pres.sure Switch Serpoint Error Analysis at Normal Operating Conditions (DIS 1300-1), Rev. 0.
- 2. DIS 1300-01, Sustained High Reactor Pressure Calibration, Revision 07.
Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers Discrenancy #12 l These following parameter references and calculation discrepancies were identified:
- a. Thrust adequacy calculations for Valves 2(3)-1301 10. ;
- b. Thrust adequacy and minimum operator terminal voltage calculations for Valves-2(3) 4102.
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- c. Minimum terminal voltage, start voltage, and run voltage for Pumps 2(3) 3319A & B.
- d. Minimum terminal voltage, start voltage, run voltage, and NPSH for Pumps 2(3)- 3320.
- c. Minimum terminal voltage, start voltage, run voltage, and NPSH for Pumps 2/3-4303A & B.
H , Discrecancy Resolution PIF resolution will disposition this issue. 1 Tracking Numbers l I The PIF addressing this discrepancy is being processed. , Offsite Power.5?vstem Discrenancy #1 j l System Planning Operating Guide (SPOG) 1 1 A, rev. O, dated 04-01-93 " Operating Nuchar Stations At Reduced Voltage" references a Dresden second level (degraded voltage) undervoltage reset point of 3.903 kV, This reset point is not consistent with the degraded voltage reser (maximum , pickup) point of 3.927 kV which the latest relay settings reflect. SPOG l-1 A is the Bulk Power l Operations guide which is utilized to maintain nuclear stations switchyard voltages at adequate levels. l Discrecancy Resolution PIF resolution will disposition this issue. i 1 ! Trackine Numbers The PIF addressing this discrepancy is being processed. I f I E-17 Ve fa:auen Screening o(Key Parameen Reveson 0
Emergency Core Cooling System (ECCS) Initiation Logic Discreoancy #1 The Analytical Limit for ECCS initiation on High drpell pressure is listed as 2 psig. in the TSUP's and the original Tech Specs. The existing setpoint for high drywell pressure initiation of Core Spray, LPCI, HPCI and ADS in the EWCS IDS data sheets is 53" we., equivalent to 1.92 psig. Instrument setpoint calculation NED-I-EIC-0091 has a setpoint of 53", which results in a negative margin. An operability determination was performed on 12-30-92, with input provided by GE Report GE NE 187-66-1291. This report provided a safety evaluation tojustify increasing the analytical limit (AV) for the Drywell high pressure signal to 3 psig, stating there would not be any significant effect on the timing of the reactor scram because of the rapid pressurization rates for postulated events Revision I was made to calculation NED-I EIC-0091, revising the setpoint to a value of 49" we. (equivalent to 1.77 psig) with a positive margin. However, this revised setpoint was not implemented in Dresden Instrument Surveillance (DIS) 1600-04, Revision 13, which still establishes the instrument setpoint as $3" we. This calculation also applies a head correction for drywell atmosphere (air). Although the head correction established is negligible, it should not have been used in this application, since the instrument is located above the instrument tap, and the sensing line is considered to be dry. Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. Tracking Numbers PIF #227A 121997-011938 (63) OP EV #97-21 Discrecancy #2 Dresden calculation NED-I-EIC-0100 incorrectly used a reactor low water level setpoint of 51" RWL instead of the correct value of $2" RWL. This calculation needs to be revised to correct this error. Discrecancy Resolution The calculation will be revised. Tracking Numbers PIF# 227A-12 1997-012190(73) NTS# 249-140-97-01401 E-18 Venrwerma $asenemy of Key Nuneters. Reisroe 0
i I liigh Pressure Coolant injection (llPCI) System I Discrecancy el l l The pressure range in which the HPCI system is required to deliver 5,000 gpm differs between section H.I.a(1) of Dresden Operating Surveillance (DOS)2300-03 and section 4.5.A.3.b.1 of the TSUP. It is believed the correct pressure range is from 150 to 350 psig as stated in the TSUP. Discrecancy Resolution The procedure will be revised. Trackine Numbers PIF #227A.121997-012039 (80) NTS #237140-97-00201 Discreoancy #2 The setpoint for the HPCI system isolation on low reactor vessel pressure is identified the UFSAR as 100 psig. However, the TSUP identifies the setpoint as 80 psig. A letter from General Electric (GE) was provided in calculation NED-I EIC 0110 stating that the correct setpoint is 100 psig. This calculation also states that the TSUP should be revised to reflect the correct setpoint of 100 psig. Discrenancy Resolution The TSUP will be revised. l Trackine Numbers NTS #237100-96-20100 Discrecancy #3 Adequate NPSH is demonstrated for the HPCI pump when aligned to the Condensate Storage Tank (CST) with significant margin by an " evaluation", and a formal calculation could not be located. Calculation 87-981N82 determined the maximum static pressure available to the HPCI pump suction when aligned to the CST. However, the conditions used in this calculation should not be used when determining NPSH available for the HPCI pump. Furthermore, this calculation only determines static pressure, which isjust one part of determining NPSH. Although the data in the evaluation appears to be correct, a formal calculation needs to be performed to provide the appropriate basis for pump NPS!! when aligned to the CST. In addition, no formal calculation could be identified that provides a basis for the NPSH available to the HPCI pump when aligned to the torus. A value is given in section 6.3.3.4.3 of the UFSAR, the section that provides the analysis of determining the NPSH available to Emergency Core Cooling System (ECCS) pumps. However, this analysis applies to the Low Pressure Coolant Injection (LPCI) and Core Spray pumps. Calculation 87-981/982 determined the maximum static pressure available to the HPCI pump suction when aligned to the torus. However, the conditions used in this calculation should not be used when determining NPSH available for the HPCI pump. Furthermore, this calculation only determines static pressure, which isjust one part of determining NPSH. An analysis similar to the one performed in the UFSAR, documented in a formal calculation, needs to be performed to ensure adequate NPSH available to the HPCI pump when aligned to the suppression pool. E-19 Vaifwassee Saeoning of Key Parameurs, Reision 0
l 1 i Discreoancy Resolution ' This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. l Trackino Numbers 1 PIF #227A-121997-012191(65) NTS #237-140-97 01201 Discreoancy #4 No calculation could be identified that provides a basis for the mass flowrate and steam pressure at the turbine inlet for the pressure rarse specified. Per the HPCI system design specifications and i process flow diagram, the estimated friction line loss from the reactor vessel to the turbine inlet is 10 psi. Howe,er, there is no formal ca'culation that validates that number. A formal cah.ulation needs to be performed to ecsurt Lati:he te.Pr pressure vessel can provide steam to the terbine at the design conditions. Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis - Calculations project. Tracking Numbers PIF #227A 121997-012188(9) NTS #237140 97-01101 Discrepancy #5 Calculation NED-EIC-MO%DR-0007 calculates the motor terminal voltages for the HPCI system 2301 motor operated valves (MOVs). Calculation DRE96-0126 determines motor terminal voltages for MOVs that are powered by the 250 Vdc system. The MOVs that are listed in both documents are the HPCI (2301) system MOVs that are powered by the 250 Vdc system. These valves are 2(3). 2301-3, $,6,8,9,10,14,15,35,36,48, and 49. For all of these valves, the value for the motor terminal voltage is lower in calculation DRE96-0126 than the value in NED EIC-MOWDR 0007. This reason for this appears to be the added conservatism used in DRE96-0126, as the motor terminal voltages calculated are determined to be " worst case". However, thrust calculations used the motor terminal voltage value from NED-EIC-MO%DR-0007 to perform its thrust analysis. If the values calculated in DRE96-0126 are to be used for design input with other calculations, then NED E!C MOWDR-0007 should be revised to ensure that the motor terminal voltage values for the above valves are obtained from DRE96-0126. In additicn, thrust calculations should be revised, or a new thrust calculation should be performed to supersede existing ones; to ensure that the motor terminal voltage produces the required torque to actuate the valve. Discrenancy Resolution PIF resolution will disposition this issue. E 20 v.*=so aunm r mu o
Tracking Numbers The PlF addressing this discrepancy is being processed. Discrecancy #6 Discrepancies were found in thrust calculation for the failowing three cases:
- 1. The minimum required thrust to close (and open) the 3 2301-3 valve is greater in calculation 004-MN-311 than calculation 004 MN-302. This is most likely due to the calculation methodology and the added conservatism in 004-MN-311. The maximum available thrust is the same in both calculations and exceeds the minimum required thrust in both cases. However, 004-MN-302 should be revised so that 004 MN-31I is used as design input in the future to ensure design contro!.
- 2. The maximum available thrust to e 4,e and open the 2-2301-35 valve differs calculation 004 ;
MN-349 and calculation 004-M-058. It is unclear at this point which value is correct or more l conservative. The minimum required thrust b the same in both calculations and is exceeded by j the maximum available thrust in both cases. However, the correct maximum available thrust should be determined and the necessary documents revised to ensure one value is used in the future.
- 3. The motor terminal voltage at valve 2(3)-2301-14 differs between calculation NED-EIC-MOV-DR-0007 and 7328-00-19 1, 7328-00-19-1 was an older calculation that was prepared to determine the voltage drop from the battery to the valve terminals. This methodology is different from the one used in NED-EIC-MOV DR-0007, and is not as conservative. Also, the motor terminal voltage from 7328-00-19-1 was not used for the valve thrust calculations. This i calculation should be revised or superceded to ensure design control.
Discrecancy Resolution l PIF resolution will disposition this issue. ) J Tracking Numbers . The PIF addressing this discrepancy is being processed. I Discrecancy #7 The turbine stop valve solenoid valve SV-8 is powered by the 125 Vdc batteries. No calculation could be found that determines the minimum voltage available at the so'enoid valve terminals, or if l this voltage is adequate to properly operate the valve. A voltage drop calculation for both Unit 2 ) and 3 needs to be performed for the 125 Vdc system, similar to the one prepared for the 250 Vdc j system in calculation DRE96-0189, to ensure adequate voltage is available at the turbine stop valve solenoid valve terminals. 3 Discrecancy Resolution This activity will be evaluated and addressed under the Adequacy and Retrievability of Design Basis Calculations project. j I l E-21 Yentiauos $cnnens of Key Pwuneierii, Reviseos o J
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i l l Tracking Nnmhg
- PIF #227A 121997-012562 l NTS #237-140-97-02201 Discreoancy 88
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- Dresden Calculation DRE96-0210,"HPCI Pump Margins with Respect to Appendix K Curve," was i perfonned to demonstrate that the HPCI system can deliver flow to the reactor vessel in excess of
! what is required in the Appendix K LOCA analysis. - However, this calculation has not been } completed and signed prepared, reviewed, and approved. This calculation needs to be completed and signed to ensure a formal calculation exists that can provide a basis for HPCI design flowrate. 4
- Discrenancy Resolution 1
e PIF resolution will disposition this issue. i
- Trackine Numbers
- The PIF addressing this discrepancy is being processed.
t t l l. i i i i i l l 1 i l. ? I i i E 22 Verdcanon $mening of Kay Paraneeers, Revitee 0 u,___-----________ - _ - , ._ - . .-, - , - , . . . , - . . . - ,}}