ML20217N346
| ML20217N346 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/31/1998 |
| From: | Heffley J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JMHLTR:#98-0085, JMHLTR:#98-85, NUDOCS 9804090082 | |
| Download: ML20217N346 (17) | |
Text
- 1. . Q>mmonwealth rxlism O>mpany Drculen Generating Station 6500 Nonh Drexten Road Moms. IL 60150 Tel H15-9412920 t
- March 31,1998 JMHLTR: #98-0085 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
Subject:
Design Basis Initiative Program Dresden Nuclear Power Station Containment Cooling Service Water Failure of the Dresden Lock and Dam NRC Docket Nos.50-010. 50-237 and 50-249 Reference (a) J. M. Heilley (Comed) to USNRC letter dated March 13,1998 regarding " Design Basis Initiative Program" The purpose of this letter is to provide the basis for Dresden's conclusion that the dam failure coincident with a LOCA is beyond the design basis of Dresden Units 2 & 3.
Furthermore, a Licensing Amendment is not necessaiy and clarifications to the UFSAR may be made through the provisions of 10CFR50.59.
Dresden Station personnel identified several discrepancies in Section 9.2.5 of the Updated Final Safety Analysis Report (UFSAR) which addresses dam failure coincident with a loss-of-coolant-accident (LOCA). Specifically, all existing LOCA analyses are based on 2 Containment Cooling Service Water (CCSW) pumps starting within 10 minutes following a LOCA, not i CCSW pump starting 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a LOCA as described in Section f 9.2.5.3.2. In reference (a), Comed indicated that a Licensing Amendment would be submitted in March 1998 to clarify the licensing basis with respect to dam failure. These I discrepancies are similar to those noted by NRC Region Ill inspectors in Inspections 50-237/249-97021 and 50-237/249-98007.
Upon funher review of this issue, Comed Design Engineering has concluded that a Licensing Amendment is not necessary and that clarifications to the through the provisions of 10CFR50.59. Considering a dam failure coincident with a LOCA is beyond the design basis of Dresden Units 2 & 3. The coping scenario for a dam i failure coincident with a LOCA currently described in Dresden's UFSAR Section i 9.2.5.3.2 was not used as the basis of acceptability in the Safety Evaluation Report (SER) by the NRC for original plant licensing. The description of the Ultimate Heat Sink (UHS) in UFSAR Section 9.2.5 was mentioned in the SER for Technical Specification 3/4.8.C. 1 1
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USNRC March 31,1998 JHMLTR: #98-0085 Page 2 Therefore, Comed must continue to assure that no plant changes will occur that reduce the margin of safety implied in the coping scenario for this beyond design basis event.
This conclusion is based upon the analysis and information contained in Attachment I to this letter. Therefore, a Licensing Amendment, as previously identified, will not be required to address the failure of the Dresden Lock and Dam. Additionally, no design basis calculations will be generated.
As a result, Comed is taking the following actions to resclve the failure of the Dresden Lock and Dam:
- 1. The Dresden Station UFSAR Section 9.2.5, Ultimate Heat Sink will be revised to resolve discrepancies identified in Section 9.2.5.3.2 during the Design Basis Initiative Program and other discrepancies found in Section 9.2.5.3.1.
- 2. Dresden Station Procedure DOA 0010-01, Dresden Lock and Dam Failure, will be revised to incorporate the changes to UFSAR Section 9.2.5.
- 3. Dresden Station will evaluate developing a seismically qualified or verified path to obtain water from the UHS and deliver it to the shell of the isolation condenser of each unit. (NTS 237-123-98-00100B)
- 4. Dresden Station will notify the NRC of the results of the above evaluation.
If you have any questions, please contact Frank Spangenberg, Dresden Regulatory Assurance Manager at (815) 942-2920 extension,3800.
Sincerely, Jidd Vice Pr ident Dresden Stati Attachments cc: A. Bill Beach, Regional Administrator, Region III NRC Resident Inspector's Office L. W. Rossbach, Dresden Project Manager j
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Before Amendments 9/10, the original FSAR did not address how Dresden Station would achieve safe sh.ndown or mitigate the consequences of a LOCA following a dam failure.
Section 9.2.5.3.2 in the current UFSAR is based on a coping scenario which was provided to the NRC on February 28,1969 in response to Question I.F in Amendment 9/10 of Dresden's original FSAR. Question I.F reads:
"With regard to A.10., above, we understand that failure cf the Dresden Island Lock and Dam could reduce the Kankakee . River to a level below the intake of the main cooling water canal. If this failure were to occur as a result of an eanhquake which disabled all Class II systems, it appears that the availability of an ultimate heat sink for all units wousd be uncertain. On this basis a complete safety evaluation of the consequences of such a failure should be provided for our review.
The evaluation should include elevation drawings showing the dam, the intake structures, and the levels at which service water pumps would lose suction.
- 1. Provide an evaluation of the seismic design of the lock and dam indicating whether the structure is considered Class I.
- 2. Provide an evaluation of the ability of Units 2 and 3 to cope with the effects of an earthquake during normal operation which causes coincident failures of the dam, all Class Il systems, and offsite electrical power. How would the comequences be affected by the availability of offsite power?
- 3. Repeat the evaluations assuming in addition that the earthquake results in the design basis loss-of-coolant accident in one of the two units."
The coping scenario in FSAR Amendment 9/10 (attached) was reviewed by the NRC prior to the NRC issuing Safety Evaluation Reports for Dresden Unit 2 and Unit 3 dated October 17,1969 and November 18,1970.
Section 2.3 Hydroloev ofthe Unit 2 Safety Evaluation Report contains the following l_ relevant paragraph:
l "The site elevation is 516 feet as compared with the maximum historical flood I
elevation of 506.4 feet and the normal pool elevation of the river as controlled by Page1of9 JMHLTR: #98-0085
the Dresden Dam of 505 feet. The facility is designed so that tuflicient water to assure safe shutdown will be impounded in the intake and discharge canals for coeling in the event of a failure of the Dresden Dam and a subsequent lowering of the pool elevation of the rivers."
This paragraph does not indicate that water impounded in the canals needs to mitigate the consequences of a LOCA.
Section 2.4 Geology and Seismology of the Unit 2 Safety Evaluation Report contains the following relevant paragraph:
"The facility was designed to withstand the effects of an earthquake corresponding to a maximum horizontal ground acceleration of 0.10g. Facility components and structures are designed such that the loads caused by an earthquake of this magnitude in combination with opereGng loads do not exceed code allowable stresses and for ground accelerations of 0.20g, there will be no loss of function of critical structures and components necessary to assure a safe and orderly shutdown. We and our consultant, Dr. Newmark (Nathan M. Newmark Consulting Engineering Services), have reviewed the seismic design of the facility."
This paragraph does not indicate that a LOCA must be considered coincident with an earthquake.
! The Unit 3 SER did not contain any relevant paragraphs since only the difTerences from Unit 2 were addressed.
- Based on these statements in the SER, it is concluded that postulating an earthquake that l causes a dam failure coincident with a LOCA is beyond the design basis of Dresden Station Units 2 & 3. Thare is no indication that the coping scenario provided in response to Question 1.F.3 was used as the basis for acceptability in the SERs for Dresden Units 2
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Systematic Evaluation Program & SER's The availability of water impounded in the canals was evaluated during the Systematic Evaluation Program (SEP) under Topic II-3.C " Safety-Related Water Supply". The Franklin Rssearch Center (FRC) produced a Technical Evaluation Report (TER) for Ilydrological Considerations dated May 7,1982 which addressed Topic II-3.C. The TER indicated the criteria by which the UIIS was evaluated were taken from Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants" The criteria included the ability of the UliS to provide suflicient cooling for at least 33 days (a) to permit simultaneous safe shutdown and cooldown of all nuclear reactor units that it serves and to maintain them in a safe shutdown condition, and (b) in the event of an accident in one unit, to limit the effects of that accident safely, to permit simultaneous and safe shutdown of the remaining units, and to maintain them in a safe shutdown condition.
The TER, however, only addressed the ability of the UHS to perform function (a) which is to permit simultaneous safe shutdown and cooldown of all nuclear reactor units that it serves and maintain them in a safe shutdown condition. The TER is silent on the ability of the UHS to limit the effects of an accident in one unit and safely shutdown the other.
The Dresden 2 Nuclear Generating Station, Safety Evaluation of Hydrology SEP Topics ll-3. A, II-3.B, ll-3.B.1 and ll-3.C was transmitted to Commonwealth Edison in a letter dated June 21,1982 from Paul O'Connor (NRC) to L. DelGeorge (Comed).Section VI.
Conclusions of the SER contains the following relevant paragraph:
"The Dresden Unit 2 UHS partially complies with the intent of Regulatory Guide 1.27. Specific areas of deviation are discussed in our evaluation (Section V) and in Section 3.4.3 of FRC's TER. The acceptability of the UHS is contingent on the development of appropiate and acceptable emergency procedures and technical specifications to cover flow scenarios that inundate safety equipment and seismic scenarios that fail the Dresden Island Lock and Dam and deicing line. These issues will have to be resolved during the integrated assessment."
. The June 21,1982 letter that transmitted the SER for Topic II-3.C stated that the ability of the safety related water supply (UllS) to meet the current NRC criteria is dependent on the acceptability of the Flood Emergency Plan (EPIP-200-11). Emergency procedures for l seismic scenarios that fail the Dresden Dam are not mentioned.
l l Section 4, Integrated Assessment Summarv. of NUREG-0823. Lntegrated Plant Safety I
Assessment. Systematic Evaluation Programm Dresden Nuclear Power Station. Unit 2, Page 3 of 9 JMHLTR: #98-0085
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dated February 1983 was reviewed. Discussion pertinent to the UHS following a dam failure is contained in Section 4.1.4 " Flood Emergency Plan" on pages 4-2 and 4-3:
"The existing flood emergency plan does not provide assurance that the ultimate heat sink can provide for a safe plant shutdown. Scenarios exist that the would result in the flooding of safety equipment. Also, failure of the Dresden Island Lock and Dam would result in low water levels that may affect the capability of the ultimate heat sink. Further, plant procedures require internal flooding of structures , which could result in a loss of all reactor cooling.
The staff has recommended that the licensee have the capability to install and operate an emergency pump above the PMF level capable of providing 100% makeup water to the isolation condenser and other cooling needs for the duration of the flood, including the time needed to restore the operation of flood-damaged components. The plant currently has the capability to use a portable pump to supply cooling water directly to the isolation condenser using a fire hose connection.
By letter dated November 17,1982(a), the licensee committed to revise the flood emergency procedure. Included in these changes will be flood predictions, gasoline pump connections and fuel supplies, intake canal level gauge, and clearer direction on the use of instruments."
These paragraphs and all other paragraphs in this section focused on the flood emergency plan. Emergency procedures for seismic scensrios that fail the Dresden Dam are not mentioned.
Commitments made in the November 17,1982 letter are implemented in DOA 0010-04,
" Floods" Therefore, Dresden Units 2 & 3 comply with the basis for acceptance during the SEP. Technical Specification 3.8.C " Ultimate Heat Sink" has a Limiting Condition for Operation (LCO) for water level and temperature in the cribbouse. DOA 0010-01 constitutes an emergency procedure to address a seismic scenario that fails the Dresden Lock and Dam and DOA 0010-04 constitutes an emergency procedure that addresses actions to be taken during floods.
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I Technical Specification Upgrade Program & SER's
.A new Technical Specification Section (3.8.C) for the Ultimate Heat Sink was added during the Technical Specification Upgrade Program. TS Section 3.8.C requires a minimum water level of 500 fl mean sea level (to be changed to 501 ft 6 in) and an average water temperature less than or equal to 95 degrees Fahrenheit. If these limiting conditions for operation (LCO) cannot be met in operational MODES 1,2, and 3, the plant must be immediately shutdown.
The basis for Technical Specification 3/4.8.C reads:
"The canals provide an ultimate heat sink with suflicient cooling capacity to either provide normal cooldown of the units or to mitigate the effects of accident conditions within acceptable limits for one unit while conducting a normal cooldown on the other unit."
The Safety Evaluation by the Oflice of Nuclear Reactor Regulation related to Amendments No.144 and No.138 states in Section 3.3, TS 3/4.8.C: Ultimate lieat Sink that the UllS is described in UFSAR Section 9.2.5. Therefore, it is concluded that the NRC found TS 3/4.8.C acceptable based on the description of the UliS in UFSAR Section 9.2.5. It is also conservatively assumed that the description referred to in the SER includes the coping scenario described in Section 9.2.5.3.2 and not just the physical description of the canals.
Therefore, since the cop;ng scenario described in UFSAR Section 9.2.5 may be viewed as an implied margin of safety for a Technical Specification, Comed will maintain the description of the coping scenario in UFSAR Section 9.2.5.3.2 and ensure that future plant changes do not reduce the margin of safety. Ilowever, the coping scenario is still considered beyond the design basis of Dresden Units 2 & 3 and design basis calculations will not be generated.
Measures to Ensure a Margin of Safety Is Maintained Commensurate With the Coping Scenario Described in UFSAR Section 9.2.5.3.2 In addition to Tech Spec 3.8.C for the UliS, Dresden Station complies with Dresden Administrative Technical Requirement (DATR) 3/4.19.2 "Dresden Lock and Dam" The LCO for DATR 3.19.2 requires that the Dresden Lock and Dam shall be capable of maintaining level in the CCSW pump suction bays greater than or equal to 501 ft 6 in. If Page 5 of 9 JMIILTR: #98-0085
this LCO cannot be met, the CCSW systems must be declared inoperable per Technical Specification 3/4.8.A and appropriate actions taken.
How Dresden Station would respond to a failure of the Dresden Lock & Dam is detailed in procedure DOA 0010-01 "Dresden Lock and Dam Failure" This procedure provides instructions on how to refill the suction bay for the CCSW intake pipes by installing stop logs, opening CW pump bay drain valves to fill the refuse pump sump pit, aligning refuse pump discharge valves, and starting the refuse pumps. These measures would allow the operation of a CCSW pump.
The ability to achieve a CCSW pump flow of at least 3500 gpm using the method in DOA 0010-01 is tested every third refueling omge using procedure DOS 0010-01 "Dresden Dam Failure / Containment Cooling Water Functional Test" This functional test is required by Dresden Administrative Technical Requirement (DATR) 4.19.2.A.
Dresden Station will continue to maintain DATR 3/4.19.2, DOA 0010-01, and perform testing via DOS 0010-01.
Probabilistic Safety Assessment In a Stability Investigation that was performed by the U. S. Army Corps of Engineers in !
1973, the Dresden Lock and Dam was evaluated for earthquakes with a static equivalent horizontal acceleration of 0.05g for the dam and 0.10g for the lock and found to be acceptable. The safe shutdown earthquake (SSE) defined for Dresden Station has a magnitude of 0.2g zero period acceleration (ZPA) horizontal ground acceleration. For the Dresden site, the mean probability of an earthquake with a peak ground acceleration of j 0.05g is 4.6E-04/ year according to NUREG-1488, " Revised Livermore Seismic Hazard {
Estimates for 69 Nuclear Power Plant Sites East of the Rocky Mountains" However, i because of safety margins, the actual peak ground acceleration that would fail the dam would be greater than 0.05g and have a lower rate of occurrence. The Individual Plant Examination (IPE) for Dresden Units 2 and 3 used information from WASH-1400 indicating that the frequency (probability per unit of time) of a LOCA is 3.0E-04/ year for a large LOCA, 8.0E-04/ year for a medium LOCA, and 3.0E-03/ year for a small LOCA. For 1 the purposes of this evaluation, coincident is dermed as " occurring in the same month j (approximately a 30 day period)" and is based on the R.G.I.27 requirement that an UHS l be designed for a 30 day supply of water. Therefore, the frequency of an earthquake with l a peak ground acceleration of 0.05g coincident with a LOCA is 1.6E-07/ year {(3.0E-04/ year + 8.0E-04/ year + 3.0E-03/ year )/(12 months / year)* (4.6E-04/ year)/(12 i months / year)* 12 months / year }. The results are documented in Calculation DRE98-Page 6 of 9 JMHLTR: #98-0085
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- 0053, Rev.0. Guidance in Standard Review Plan (SRP) Section 2.2.3 " Evaluation of l
. Potential Accidents" indicates that accidents where the expected rate ofoccurrence is approximately 1.0E-07/ year or less need not be considered as a credible event.
Furthermore, it is acceptable to not postulate accidents where the expected rate of occurrence is approximately 1.0E-06/ year if, when combined with reasonable qualitative arguments, the realistic probability can be shown to be lower. Therefore, consistent with this guidance, the frequency of a coincident LOCA and earthquake is incredible and the two events satisfy the criteria in SRP Section 2.2.3 fe exclusion from design consideration.
Safe Shutdown Following an Earthquake and Dam Failure During Normal Operation UFSAR Section 9.2.5.3.1 describes how Dresden Station Units 2 & 3 would achieve safe shutdown using the isolation condenser system following an earthquake during normal operation which causes coincident failures of the dam, all Class II systems, and offsite electrical power. This section is based on Commonwealth Edison's response to Question I.F.2 in Amendment 9/10 of Dresden's original FSAR.
Vulnerabilities in using the isolation condenser system following an earthquake and dam failure were identified in " Report of AEC Regulatory Staff, Adequacy of the Structural Design for Dresden Nuclear Power Station Units 2 & 3" by N.M Newmark and W.J. Hall dated November 17,1969. Specifically, safe shutdown depends the ability of the fire protection system to provide make-up water to the shell of the isolation condenser.
However, fire protection is a Class 11 system (non-safety-related and non-seismically qualified). This report was provided as Appendix D to the Unit 3 SER. However, no additional requirements were imposed in the Unit 3 SER.
The vulnerability of the fire protection system to an earthquake was also identified during the SEP. FRC's evaluation of Comed's response to Question I.F in the FSAR is contained in Section 3.4.3.1 " Vulnerability of the UHS to Failure of the Dresden Lock and Dam" of the TER. Regarding make-up to the isolation condenser by either the diesel-driven fire pump or a local city fire tmck, FRC concludes on p. 63 that:
"taking into account the limitations that may be imposed on freedom of movement following the occurrence of a severe canhquake, it can be concluded that the time available is not suflicient to isolate failed pans of the fire system or to rely on the use of a fire tmck."
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- This conclusion indicates that make-up to the isolation condenser through the fire protection system immediately following a seismic event was not used as a basis for approval.
The FRC based their approval of using the isolation condenser for safe shutdown following a dam failure on the ability of the isolation condenser to function following an earthquake.
"In Reference 49 (NUREG/CR-0891 " Seismic Review of Dresden Nuclear Power Station - Unit 2 for the Systematic Evaluation Program", Lawrence Livermore Laboratory, April 1980), a seismic review team concluded that, in the case of Dresden Unit 2, there is strong reason to believe that the systems required for safe shutdown will remain functional under the design hazard (i.e., a SSE of 0.2 g).
This conclusion was predicated upon the redundancy of safety systems and components within safety systems and on the premise that a comprehensive equipment maintenance program has been carried out. The seismic review team concluded that the isolation condenser would withstand the 0.2 g SSE without loss offunction."
The SER for Topic II-C.3 did not impose any additional requirements.
As described in UFSAR Section 5.4.6, makeup water for the isolation condenser can be supplied from several sources. The preferred source is from the clean demineralized water storage tank via two diesel driven isolation condenser makeup water pumps, located in the Isolation Condenser Pumphouse. Alternately, water can be supplied from the clean demineralized water storage tank via two clean demineralized water transfer pumps. If clean demineralized water is unavailable, the fire protection system is the preferred source.
The 6re protection system has access to an inexhaustible supply of river water supplied either by the sersice water pumps or the diesel-driven fire pump. Make-up water is also available from the condensate storage tanks via the condensate transfer system. Although the condensate transfer and fire protection systems were designed to the State ofIllinois Code which accounts for earthquake loadings, none of these sources were designed to i Class I seismic requirements. I Using Problem Identification Form D1998 00455, Comed personnel recently identified that the suction of the Unit 2/3 fire pump would become uncovered following a dam failure. This is contrary to the description in UFSAR Section 9.2.5.3.1 which indicates the fire pump suction would still be submerged following a dam failure. However, the fire pump suction bay would be reflooded within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a dam failure through the Page 8 of 9 JMHLTR.: #98-0085
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l actions required by DOA 0010-01. Since the TER did not credit the fire protection system with the ability to provide make-up water to the isolation condenser immediately following an earthquake, the safety consequences are insignificant.
Because achieving safe shutdown following an earthquake and dam failure is important to ,
plant safety, Dresden Station will evaluate developing a seismically qualified or verified path, capable ofwithstanding the effects of a Safe Shutdown Earthquake, to obtain water from the UHS and deliver it to the shell of the isolation condenser of each unit.
Discrepancies in UFSAR Section 9.2.5.3.1 " Dam Failure During Normal Operation" will be resolved through the provisions of 10CFR50.59.
Whether postulating an earthquake that causes a dam failure during normal operation was part of the plant's original design basis is not clear. However, like the coping scenario for dam failure coincident with a LOCA in Section 9.2.5.3.2, the coping scenario for dam failure following an earthquake during normal operation in Section 9.2.5.3.1 will remain a part of the licensing basis for Dresden Units 2 & 3.
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D 2, 3 1, y.1 QUESTION I.T With regard to A.10., above, we understand that failure of the Dresden Island Lock and Dam could reduce the Kankakee River to a level below the intake of the main cooling water canal. If this failure were to occur as a result of an earthquake which disabled all Class Il systems, it appears that the availability of an ultimate heat sink for all units would be uncertain. On this basis a complete safety evaluation of the consequences of such a failure thould be pro-ided for our review. The evaluation should include eleva-tion drawings showing the dam, the intake structures, and the levels at which service water pumps would lose suction.
- 1. Provide an evaluation of the seismic design of the lock and dam indicating whether the structure is considered Class I,
- 2. Provide an evaluation of the ability of Units 2 and 3 to cope with the effects of an earthquake during normal operation which causes coincident failures of the dam, all Class II systems, and offsite electrical power. Ilow would the conse-quences be affected by the availability of offsite power ?
- 3. Repeat the evaluations assuming in addition that the earthquake results in the design basis loss-of-coolant accident in one of the two units.
ANS%TR
- 1. The Dresden Dam and Lock are maintained and operated by U. S. Army Corps of Engineers as part of the Blinois River Waterway. The design of the lock and dam was started following World War I by the State of Blinois and was to be the first of a series of dams and locks to permit navigation on the Illinois River. Construction was started in 1925, however the State's funds were soon depleted. The Federal Government assumed responsibility and construction was completed in 1932.
In discussions with the Corps of Engineers, the design of the dam and lock us not based on any seismic criteria as is used today. The dam and lock were designed to withstand the large forces due to the mass movement of ice flows from the Des Plaines and
, Kankakee Rivers, floods during periods of heavy runoff, and the impact forces of run-away tows. It should be remembered that the problems of the large movement of ice were of real concern in the 1920's prior to the industrialization of the waterway above the dam.
The dam consists of eleven heavily reinforced concrete piers 10 by 45 feet at the top and 10 by 60 feet at the bottom with the taper on the down stream side. Each pier is socketed 5 feet into bedrock and anchored. Between the piers are concrete gravity section roll-aways. Above this are Tainter Gates, which control poollevel, supported from the piers.
The dam is anchored to the rock rising to the Kankakee Bluffs at the north end and the lock structure on the south. The lock walls are 10 feet wide at the top, 20 feet wide at the bottom and 800 feet long. The lock width is 110 feet.
Since the dam and lock are of major importance to the Chicago land area, and its function is the responsibility of the Corps of Engineers, an evaluation of its adequacy to seismic activity could only be made by the Corps. Commonwealth Edison does not have access to the detailed drawings or design calculations.
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- g. . D 2, 3 1.F-2 4-2.A Introduction The Des plaines and Kankakee Rivers provide the normal heat sink for the disposal of unusable energy inherent in the thermodynamic cycle used for the three Dresden reactors.
For Dresden Units 1, 2, and 3, the rivers also provide the principal means for removal of the fission product decay heat of the reactor nuclear core following a unit shutdown.
Aside from the main condenser circulating water system, there are four systems on Units 2 and 3 which depend on the use of river water. ' These systems are 1) LPCI in the containment cooling mode, 2) service water system, 3) fire system, and 4) cooling water
. for diesels. For Unit 1, the containment cooling and fire system are combined as one system.
2.B Topography of Circulating Water System (Figure I.F.1)
The normal pool water level above the Dresden Island Lock and Dam is 505 feet 0 inch MSL (Mean Sea I,evel). The pool level can vary from a low of 503 feet 0 inch to a high of 506 feet 5 inches MSL. The pool level below the Dresden Dam is 483 feet 4 inches MSL.
The top of the next dam downstream, approximately 25 miles at Marseilles, is 486 feet 6 inches MSL.
Units 2 and 3 share a common intake canal of approximately 1800 feet long; Unit 1 has a separate intake canal of the same length. The high point on the floor of both intake canals is 495 feet 0 inch and is located 123 feet downstream of the floating booms which protect the e~ntrance to both canals from floating debris. The canal floors then decrease in elevation until a low point of 482 feet 6 inches is reached,at_igrebay of the crib houses, l
There are two discharge canals of approximately 2000 feet in length. One canal serves Unit I and the second serves both Units 2 and 3. The high point of 498 feet 0 inch on the floor of the discharge canals, is located near the discharge flume, the point where the canals join the river. Between this high point and the discharge head works. the floor of the canal decreases to an elevation of 489 feet 0 inch.
Connecting the discharge head works of Units 2 and 3 and the forebay of Units 2 and 3 crib house is an 8-foot diameter delcing line. The bottom of deicing line in the head works has an elevation of 495 feet 0 inch. A slide gate valve is used to isolate this line when not in use. Low point of deicing line in forebay is 489 feet 0 inch.
2.C Failure of Dam - Normal Plant Shutdown If catastrophic failure of the Dresden Dam and Lock were to take place. It is still possible to effect a safe shutdown on all three units. The Class I sections of Class II systems were only considered in bringing the plant to a safe condition. It is immaterial whether the shutdown is carried out on normal auxiliary power or on the diesel generators.
As an example, assume the case with all three units in operation with normal outflow of electrical power and each unit on its normal auxiliary power mode. At this point complete failure of the dam takes place with a rapid decrease in the poollevel. The first indica-tion of trouble, the operator in the control room would have, is a drop in power require-ments of the circulating water pumps and service water pumps. Vacuum on each unit I
D 2, 3 s e 1.F-3 condenser would decrease and the reactors would scram on condenser low vacuum.
With the loss of the main heat sink, reactor pressure would increase and the isolation condenser on each unit would go into service.
During this entire sequence of events, the actions required of the operator are to trip off electrically the power to the circulating water and service water pumps to protect them from damage. Also equiprnent would have to be removed from service which either added heat to the primary system or are cooled directly or indirectly by river water.
Following the reactor scram on Units 2 and 3, the relief valves from the primary system to the suppression chamber would open to maintain a fixed pressure. Level in the reactor would be maintained by reactor feed pumps, control rod drive pump, or in the case of los.
. auxiliary power, the HPCI. With the initiation of the isolation condenser, depressurization of the primary system would start.
Each of the three reactors could now be depressurized at a controlled rate by use of its isolation condenser. By using the isolatton condenser, the primary system temperature could be reduced to 212'F in 8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and held at this point. The temperature could not be reduced below this point since the system depends on steam flow to remove the core decay heat.
Normal makeup water to the isolation condensers is supplied by etMensate transfer pumps which take water from one of the condensate storage tanks. There is apprcximately one million gallons of water available from these tanks. However, since they are Class II '
structures, they would not be available. For the case of Units 2 and 3, river water would be pumped to the isolation condensers by use of the diesel-driven fire pumps or by using a local city fire truck taking suction from the area in front of the Unit 1 intake structure and pumping into the fire system. The fire system is considered a Class II system, how-ever parts of this system can meet the requirernents of a Class I system. By use of existing valves, it is possible to sectionalize the system to isolate the failed parts.
Due to the topography of the circulating water canals and piping, approximately 9,000.000 gallons of river water is trapped within the system. This is due to the high points in both the intake and discharge canals. As the Dresden pool level would fall, back flow from the discharge canals would stop at 498 feet 0 inch, and from the intake canals at 495 feet 0 inch.
Advantage can be taken of this impounded river water as a heat sink for the long term l
removal of decay heat from the reactors. The suctions of the service water pumps for Units 2 and 3 are below elevation 495 feet 0 inch, therefore a service water pump could be valved to supply cooling water to the reactor building closed cooling system which, in turn, could be valved to cool the reactor shutdown heat exchangers. The heated service water would be discharged to the discharge canal to dissipate its heat to the environs. The water l
in the discharge canal would then be recirculated back to the intake canal through the !
deicing line.
1.
l
D 2. 3 1. F-4 Operation of the Units 2 and 3 diesel generators is assured since the suction for their cooling water pun.ss are at 487 feet 8 inches. The diesel fire pump of Units 2 and 3 has its suctica at 492 feet 0 inch.
Ins riimpun6ed ylver water, due to evaporation, could be made up by use of portable low h t high volume, engine-driven pumps. Commonwealth Edison has six 1500 gpm engirar-ariven pumps on standby at various fossil fuel generating stations. These pumps cou!d'A moved io Dresden within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Pumps are also available from large Con-
' tractors in the northern Illinois area.
Depending upon the season of the year, by use of the above procedure it will be possible to maintain the reactor primary system well below 212* F.
- 3. If, at the time of the catastrophic fa!!ure of t.he dam, any of the three units were to have a loss of coolant accident, it is still possible to handle the LOCA and safely shutdown the remaining two units. This is still possible with coincident failure of off-site electrical power and Class II systems.
If it is assumed that Unit 2 were to have a loss of coolant accident, Units 1 and 3 would be shutdown and depressurized as outlined in I-F.2.
For Unit 2 depressurization would be to its suppression chamber. None of the core cooling systems would be affected by the loss of river water except the containment cooling mode of the LPCI. This is due to the fact that the suction piping, in the Units
'I and 3 intake structure, for the containment cooling service water pumps are at 499 feet 0 inch and 498 feet 0 inch. As indicated in Figure 1.F.1, this places the suction piping above the level of water in the intake canal, which would be at 495 feet 0 inch.
In order to reduce containment pressure and cool the water in the suppression chamber.
it is necessary to raise the water level in the area of these suction pipes. With the operation of all emergency core cooling mode, the drywell pressure and teinperature =
willincrease due to the transfer of core decay energy to the containment system. In about 1/2 hour, containment system pressure will start to increase and, after a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, cooling of the suppression chamber is needed. During this period of time, measures would be undertaken in the intake structure to restore suction to the containment cooling service water pumps.
Reference should be made to the SAR Vol. II, Figure 12.1.36. The suction lines for the containment cooling service water pumps take their suction from a com;nrtment between column row B and C with its center line on column line 4. The diesel fire pump also takes its suction from this compartment. River water enters this compartment through two screened openings to the left and right of column line 4. The floor of this compart-ment is at elevation 493 feet 8 inches and the ceiling at 509 feet 6 inches. The two openings extend between these two elevations.
The wire mesh screens from each of these openings would be lifted out of place and replaced with wooden stop logs. De-watering valves, located at elevation 480 feet 0 inch,
D 2, 3 1. p.5
. s would be opened to permit river water to flow from the compartments under the circu.
lating water pumps and intake piping to the trash rake refuse pit located between column row C and D and colurnn line 7 and 8. The floor of this pit is at 477 feet 0 inch elevation.
Thus, the water in this pit will rise to 495 feet, the level in the intake canal.
Two refuse purnps take suction from this pit. The pumps are located in a cotopartment adjacent to the pit with their suction at 479 feet 0 inch. Each pump has a discharge capacity of 2400 gpm. A light weight flanged spool piece of pipe is bolted in place in a permanent pipe line between the discharge line of these pumps and the cornpartment with the containment cooling water service pumps. By proper electrical switching, the refuse pumps can be operated off the diesel generator. I One containment cooling water service pump, capacity of 3500 gpm, is placed in service which discharges to the containment cooling heat exchanger of Unit 2 and then to the dis-charge canal. River water in the discharge canal would be recirculated by way of the deicing line back to the crib house forebay. In this manner, containment pressure will be reduced and the suppression chamber water cooled.
If it is necessary to completely flood the Unit 2 containment, this is possible by use of the service uter pumps. As it was indicated in 1.T.2, the suction of these pumps are below the surface of the water impounded in the circulating water system. Flooding of the containment would be through the feedwater system.
In discussions with the Corps of Engineers, they have indicated that a number of measures would be undertaken on their part if damage were to occur to the dam. The use of increased diversion from Like Michigan, with approval of the U. S. Supreme Court, would be one method if the dam were partially damaged. This would be done to maintain pool level. Another measure would be the sinking of stone-loaded barges directly above the dam as a base for a temporary rock-filled dam. Both of these would
, help to hold levels in the Dresden canals to permit maintaining the Dresden Units in a safe condition.
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