ML20002D990
| ML20002D990 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/29/1971 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML20002D991 | List: |
| References | |
| NUDOCS 8101230752 | |
| Download: ML20002D990 (11) | |
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g Dr. Peter A. Morris, Director Re:
Docket 50-155 Division of Reactor Licensing License DPR-6 United States Atomic Energy Commission Washington, DC 20545
Dear Dr. Morris:
Your letter of July 20, 1971, requested reporting of interim improvements incorporated in our Big Rock Point Plant equipment and operating techniques with regard to emergency core cooling system per-formance.
This letter is to provide the required report and to provide further information on core spray system modifications performed in-1970 and 1971.
With regard to leak detection systems, we reported by letter, dated September 11, 1970, four methods of leak detection employed at the Big Rock Point Plant. In an effort to provide a more quantitative tea-sure of leakage from the primary coolant pressure boundary at Big Rock Point, we have since then installed an additional leak detection system and modified another. Running time meters have been installed on the enclosure dirty and clean sump pumps. Operators will record the individual operating times of these pumps and compare the recorded data to previous data obtained in periods of operation with no unexplainable leaks. This system will provide a sensitive indication of increasing leakage from the primary coolant pressure boundary. We have also relocated the dew cell from e ventilation recirculation duct that took a suction on the steam drum cavity to a ventilation exhaust duct from the recirculating pump room.
This relocation will provide indication of leakage from a broader spectrum of primary coolant pressure boundary system leaks.
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Dr. Pet;r A. Moru c 2
September 29, 1971 In addition, we are procuring additional equipment to eliminate the necassity of operator interpretation of diurnal dew point temperature changes. This system will incorporate a dew cell that monitors the dew point temperature of the air supplied to the containment vessel and a comparator which computes the moisture added to the air in the recircu-
. lating pump room. The readout will be moisture added in the recirculating pump room. We also have on ceder a continuous air radiation monitor that vill sample the air being discharged from the containment vessel. An in-crease in activity of air discharged will represent a possible increase in leakage from the primary coolant system pressure boundary. The modifica-tions discussed in this paragraph are scheduled to be completed by the end of 1971.
Our staff has reviewed Big Rock Point operating procedures and philosophy with regard to loss-of-coolant accidents. This review has re-sulted in changes in Big Rock Point operating techniques that will provide more prompt operator attention to an instantaneous small pipe break occur-rence and action that will enhance the depressurization rate in the unlikely event that a small pipe break were to occur. Our concern was that a break might well be within the makeup capability of the condensate and feed-water -
system allowing a depletion of makeup system supplies to the primary system while the plant continued in operation. We have eliminated this possibility by raising the low-level alarm in the condensate storage tank so that it will provide the operator warning of abnormally large primary system makeup
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rates.
In addition, plant operating philosophy and instructions have been changed to provide more rapid depressurization of the primary system if a rupture is suspected through manual scram, manual initiation of the emergency condenser, use of the main steam system to blow down to the condenser and maintaining feed-water flow. These instructions attempt to take advantage 4
of all possible installed means of depressurization prior to the closure of the containment isolation valves. They' do not include reopening of these valves. The results of our previously submitted reanalysis of the emer-gency core cooling system performance (submitted February 9,1970) show we do not meet the new criteria for both very large and very small breaks below the core midplane. A study of the results of the previous reanalysis in light of the new models and limits indicates that there is a good chance
6 Dr. P;ttr A. Morris 3
September 29, 1971-that presently installed systems at Big Rock Point will meet the new criteria for the big break loss-of-coolant accidents below the core midplane. Modi-fications to provide a greater supply of makeup water to the feed-water system to enhance depressurization rates under small break conditions have been planned and are discussed later in this letter. Our in-house analysis indi-cates that this modification will prevent fuel meltir.3 under previous ECCS system criteria but it is questionable whether the limits of the new ECCS system criteria vill be met. On this basis, plans have been made to reduce the inspection interval defined by Paragraph IS-241 of Section XI of the ASME Boiler and Pressure Vessel Code from ten years to three years for Items 1.1 through 4.6 of Table IS-261 of Section XI covered in our program located below the core midplane.
I am attaching a copy of our " Big Rock Point In-Service Inspection Program" for your informatica.
We are not proposing changes to the technical specifications with regard to allowable leak rates. The present technical specificatione for Big Rock Point do not address themselves to allowable leak rates, probably because of the difficulty in obtaining a quantitative measure of actual primary system leakage from a boiling water reactor with a containment con-figuration similar to ours. Our emphasis on leak detection system improve-ments is being focused on providing additional detection systems or codifying present systems such that we eliminate to the greatest degree possible re-liance on operator's senses and judgment to determine if e. leak exists.
Because of the lack of operating history with the more quantitative systems, we are reluctant to propose firm operating limits that may be either mean-ingless because they are too high or too restrictive because they are too low. We are incorporating into the Big Rock Point operating procedures re-quirements that suspected primary system pressure boundary leaks be imme-diately investigated to determine the nature of the leak. If the leak is found to be due to cracking of the primary coolant pressure boundary, the plant will be immediately shut down, the leak repaired and the cause investigated.
By letter dated February 2, 1971, we requested changes to the Technical Specifications of License DPR-6, Docket No 50-155, to accommodate operations with a modified post-incident spray system designed to provide redundancy with respect to single failure criteria and the vessel jump
Dr. ' Pater A. ' Morris 4
Septtmber 29, 1971 accident. In corrersondence, TWXs and discussions' subsequent to the sub-
-mission of this request for a change to the Technical Specifications, we
.t agreed to report on the completion of the preoperational testing program for these modified systems. The remainder of this letter is to provide such report.
Section IIB of the Request for Change to the Technical Specifi-cations detailed preoperational test procedures that were employed to demonstrate the proper functioning of the modified post-incident systems.
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Testing was conducted in accordance with these procedures. All test results were satisfactory. Several minor discrepancies arose during the testing which were procedure problems rather than system problems. With regard to Step 5 B and C of the Preoperational Twat Procedures,-it was discovered that 1
when M07051 and -M07061 were remote manually shut, they would reopen as de-signed, if the.N operation was not blocked remote manually. The procedure should have been written " remote manually shut and block MOT 051 and M07061" or 'H07070 and M07071," depending on which step you are considering. A similar problem existed with Step 6 B and C.
These valves should always open upon receipt of proper signal unless they are deliberately. blocked from operation. The problem with the preoperational test procedures was failure to recognize this design feature.
During the flushing of the backup core spray system piping, flow data were obtained to confirm design data. This data were in close agree-meet with predicted flows from engineering calculations.
The plant was returned to service on March 13, 1971, with the modified post-incident systems in operation. On April 9, 1971, by Twx, we agreed to change the actuation schemes of the core spray system and backup core spray system motor-operated valves.
The change in these actua-tion schemes is such that both the core spray system and backup core spray system motor-operated valves shall be automatically actuated by tripping.
of the " low reactor water level sensor" along with the " low reactor pres-sure device." There is no time delay in the opening of the backup core spray system valves as was described in the original submittal dated j
February 2, 1971. These changen were completed and tested satisfactorily during our early September 1971 reactor shutdown.
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i Dr. Pstsr A. Morris 5
Septembar 29, 1971.
In Appendix II of the February 2,1971 submittal, we stated
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that further testing would be performed during the February 1971 refueling outage to determine the adequacy of stored water transfer rates to the con-densate and feed-water systems. In addition, feed-water addition rate versus pressure decay and leak rate curves were being developed for the limiting small-primary system break cases at Big Rock Point. The pressure decay and leak rate curves are contained in the attached letter, R. L. Theis to R. B.
Sewell, dated April 8, 1971. -Tusting performed at the plant during the re-fueling outage revealed that the transfer rate from the condensate storage tank to the hot well was approximately 300 gallons per minute. These tests were conducted with the condenser at atmospheric pressure and the condensate storage tank level approximately 55%. The transfer rate could be enhanced by maintaining condenser vacuum or an increased level in the condensate storage tank.
Based upon the review of the dapressurization data and the test results, calculations were made for four break sizes, 0.053 ft sq and smaller, of the times required for the water level to reach the core mid-plane and the times required to achieve effective core spray system flow.
These calculations reveal that for the.05 ft sq break size the core mid-plane is uncovered 190 seconds after instantaneously severing the pipe, assuming feedwater continues to be added at the rate of 965 gallons per minute. At this time, fuel heatup is assumed to begin and first fuel melting would occur 290 seconds after water level reaches the core midplane. From data submitted in Attachment A to our February 9,1970 letter, Analysis of Loss-of-Coolant Accident, and the April 8,1971 letter, it is concluded that core spray flow will be adequate 150 to 240 seconds after water level reaches core midplane. Adequate flow is assumed to be 250 gallons per minute at 85 psia reactor pressure. Actual core spray flov would start at appioximately 140 psia reactor pressure and increase with decreasing reactor pressure. It is concluded that, if sufficient water supply is guaranteed to maintain one feed pump in operation, no fuel melting will occur for the.05 ft sq break size below the core midplane. The range in times required for adequate flow in the core spray system is due to differ-ences in the models used in the October 23, 1967 report and the data sub-l mitted April 8, 1971. This difference is primarily thac the 1967 report l
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Dr. Petsr A. Mo ria 6
Sept mber 29, 1971 assumes two phase blowdown regardless of the location of the break, while the 1971 data assume a more conservative single phase blowdown.
Calculations and recent data on the.02 ft sq break size indi-cate that the core is never uncovered prior to obtaining adequate core spray flow; therefore, no fuel melting occurs. This calculation revealed that the leak rate became equal to the makeup rate 235 seconds after the start of the accident. At this time, the water level inventory had de-creased to 1,957 cubic feet, well above the core midplane (957 cubic feet).
As the leak rate decreases as reactor pressure decreases, water inventory will start increasing after 235 seconds. The rated spray is obtained at 980 seconds after the start of the accident. The rated feedwater for one feed pump is assumed throughout the accident until reaching rated spray.
Calculations for the other two break sizes showed that the.05 ft sq break size uas the most limiting.
Consumers Power Company has requested General Electric Company to reevaluate the performance of the core spray systems in accordance with the " Interim Acceptance Criteria for Energency Core Cooling Syatems for Light Water Power Reactors." The reevaluation will include the effects of rated feed-water flow for one feed pump.
If this reevaluation yields ac-ceptable results in terms of the new criteria for the small break loss-of-coolant accident, it is planned to proceed with plant modifications as described in the subsequent paragraphs.
If at the conclusion of the first phase of the reevaluation the results for the small break loss-of-coolant accident are unacceptable, the modifications and schedule described below will be cancelled and new plans made in accordance with the interim criteria.
General Electric Company recommends that 19,300 gallons of water be available as feedwater. To achieve this water supply and transfer rate at least equivalent to supplying one feed pump operation, Consumers Power Company intends to tie the six-inch fire header outside the condenser area to the condenser. If a break of this nature were to occur, the water supply could be initiated from either the diesel fire pump or the electric fire pump, providing approximately 1,000 gallons per minute as makeup to the condenser, thus being available for injection into the primary system by a single feed pump. Assuming both feed pumps were in operation, at the time of the accident, it is most likely that the feed pu=ps would pump down the condenser to the low water level condensate pump trip set point and
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Dr. Petsr A. Morris 7
Septtmbar 29, 1971-
. both condensate pumps and feed pumps would trip. At this time, or prior to this time,;the fire water supply would be lined up and put into service.
Then a single condensate pump and single feed pump would be started and the feed-water addition rate controlled by remote manual feed-water control until the system was sufficiently depressurized to allow adequate core
. spray system flow.
It is intended to accomplish a modification of the nature described above, if it meets the interim criteria, such that it would be available for service following the currently planned March 1972 refueling outage, assuming all materials can be obtained prior to March 1972.
Yours very truly,
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RBS/ map Robert L. Haueter Electric Production CC:
BHGrier Superintendent - Nuclear Div of Comp (1) a i
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4 NUCLEAR ENEROY GENERAL ELECTRIC DIVISION ATOttlC POWER EQUIPMENT GENER# L ELECTRIC COMPANY, 175 CURTNER AVE., SAN JOSE, CALIF. 95125 D E PA11T M E N T Phone (408) 297 3000. TWX NO. 910-338-0116 April 8,1971 Regulatory R!e Cy.
Subject:
BRP System Capabilities under Small Break Loss of Coolant Conditions with Only One Feed Pump in Operation Ref:
Letter dated December 18, 1970 R. B. Sewell to R. L. Theis
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Mr. R. B. Sewell Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 We have examined the Big Rock Point Reactor Loss of Coolant accidents with only one feedwater pump available. Results of this investigation are attached.
Figure 1 shows the depressurization transient and reactor vessel leak rate for a. 05 ft2 break. The data used for this analysis is the same as supplied in Table II of my September 22, 1970 letter except the feed-water flow rate which is anisumed to be 965 gpm instead of 1930. No credit is taken for the emergency condenser. Figure 2 shows a similar 2
result for the. 02 ft break. Note that with only one feedwater pump available, water level is not restored before core spray reaches rated flow.
Our original recommendation for a water supply of 19300 gallons is siill adequate.
With only one feedwater pump in service, the rate of water usage is diminished so that the makeup rate from the condensate storage tank to the hotwell could theoretically be reduced compared to the recommendation shown in Figure 2 of my September 22, 1970 lette r.
However, because there is no way of assuring that only one feed pump will be running (short of tripping one of the pumps on an accident signal which is definitely not recommended) the transfer rate indicated for the full feedwater case should be maintained.
BE SURE TO INCLUDE MAIL CODE ON RETURN CORRESPONDENCE
2-I believe this completes the requests of your letter referenced above, however, if you have any further questions, please let me know.
5 R. L. - Theis Warranty Service Engineer Overseas Warranty Service Mail Code 130, Ext, 6547 sw 101.3.1 2
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