ML19305E780
| ML19305E780 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/30/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19305E777 | List: |
| References | |
| NUDOCS 8005200413 | |
| Download: ML19305E780 (18) | |
Text
8005200'//3 p
mU UNITED STATES
+
8 NUCLEAR REGULATORY COMMISSION o
WASHINGTON. D. C. 20555 5
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 56 TO PROVISIONAL OPERATING LICENSE NO. DPR20 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255
1.0 INTRODUCTION
By letter dated April ll,1980, Consumers Power Company (the licensee) requested an amendment to the Provisional Operating License No. DPR-20 for the Palisades Plant to pemit perfomance of a test for water hammer.
Additional infomation was submitted by letter dated April 16, 1980.
2.0 DISCUSSION AND BACKGROUND The purpose of this test is to detemine whether water hammer will be induced by the introduction of cold auxiliary feedwater into a steam filled feedwater line at flow rates greater than 150 gallons per minute (gpm). The licensee has been limiting auxiliary feedwater flow administratively to 150 gpm as a precaution against steam generator water hamer. However, the licensee now wishes to raise this limit to provide a greater margin in the inventory of cooling water in the steam generators after a plant trip. With the present instrumentation and flow rate limit, when a plant trip occurs, the water level drops below the range of the level instrumentation and the level remains unknown until sufficient auxiliary feedwater is pumped in to recover the water level indication. By increasing the flow of auxiliary feedwater, the level will be recovered more quickly; and, thus, the period of " blind" operation will be reduced. The results of this test will be used to detemine the new auxiliary feedwater flow limit that will be employed.
The Palisades Plant Review Committee reviewed procedure T-130 for this test and concluded that an unreviewed safety question could result from the performance of this test, i.e., there is a greater probability of a water hamer occurring, as a result of the test, than there would be if no test were performed. This in turn might increase the proba-bility of damage to the feedwater piping.
However, the Committee evaluated the possible consequences of a water hammer and concluded that this operation does not involve a significant hazards consideration with respect to perfomance of the test or potential impact on the health and safety of the public.
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8 f 9 3.0 EVALUATION We have reviewed the licensee's request for license amendment dated April 11, 1980, the attached water hammer test procedure T-130 Revision Zero and the additional information submitted by letter dated April 16, 1980.
The test will be perfor.ned in two major steps; one at 200 pounds per square inch (psig) pressure in the steam generator and the second at 900 psig. Auxiliary feedwater flow rates of 150 gpm and 300 gpm will be introduced at 200 psig and flow rates of 200 gpm through 400 gpm may be tested at 900 psig.
By performing the test in this manner the probability of a feedwater line break will not be significantly increased. Although the probability of a water hammer occurring at the first test point, 200 psig, may be greater than at 900 psig, the magnitude would be lower because the differential pressure (DP) driving the water slug would be only about 200 psid instead of 800 to 900 psid. With this decreased DP, the probability of a pipe break 'due to water hammer would be significantly reduced. Furthermore, if the test at 200 psig results in water hammer, the test will not be continued without further evaluation.
Even if a water hammer were to occur at full pressure, the consequences would not be likely to cause a feedwater line break. There has been only one instance in the United States, Indian Point, Unit 2, where a feedwater line was broken by steam generator water hammer. The line broke as a result of repeated water hammer and would not have broken from a single blow.
If a water hammer were to occur during the full pressure test, the same test conditions would not be repeated to allow additional water hammer.
Although we have concluded that the probability of a feedwater line break will not be significantly increased by the performance of this test, we have considered the consequences of such a break during the performance of this test.
In terms of potential radiation doses to the public and operating personnel, the consequences would be far less than those previously considered in the Final Safety Analysis Report. Because the plant has been shutdown since September 7, 1979, the level of radioactivity in the secondary coolant is only 10-7 pCi/gm or less. This is at least a factor of one million below the authorized limit of 0.1 uCi/gm given in the Technical Specifications for the operating license. Potential radiation doses are, therefore, significantly reduced. Furthennore, the primary coolant activity is only 2x10-4 pCi/gm. This is a factor of ten thousand below that allowed by the Technical Specifications and provides an additional margin of safety in the event that some steam generator tube leakage might occur.
O i8 i In the unlikely event of a feedwater line break, it is our judgement that the operating crew will be prepared for such an event and no one would be injured. The situation would be well understood and completely controlled.
Although it is not very likely, we have also considered the consequences of the disablement of the auxiliary feedwater system during this test.
Because the plant has been shut down since Septenber, the reactor core decay heat level is so low that auxiliary feedwater would not be needed to cool the core or cooldown the plant. The core can be cooled by any one of the following systems:
(a) nomal makeup and letdown, (b) water inventory from the steam generators, (c) high pressure safety injection and power operated relief valves, and (d) low pressure safety injection.
Therefore, even if the auxiliary feedwater system were completely disabled during this test, there would still be adequate and redundant means available for cooling the core.
4.0
SUMMARY
We have reviewed the test procedures that the licensee will use to perform a test for steam generator water hantner. We have considered the probability of causing a rupture in the feedwater line as a result of this testing and the consequences of such a rupture. We have found that the probability of a feedwater line rupture would not be significantly increased and that the consequences of such a rupture under the planned test conditions would be far less than those previously considered in the Final Safety Analysis Report.
Based on these findii., we have concluded that the performance of test T-130A, Revision Zero does not involve a significant hazards consideration and that the licensee may proceed with this test during startup testing at the beginning of Cycle 4.
We found it necessary to modify the licensee's proposed license amendment. We discussed these changes with the licensee and we have mutually agreed upon them.
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5.0 ENVIRONMENTAL CONSIDERATION
We have detennined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this detennination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment,
6.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's i
regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the pu blic.
Date: April 30,1980
7590-01 g.
UNITED STATES NUCLEAR REGULATORY C014ISSION DOCKET NO. 50-255 CONSUMERS POWER C04PANY NOTICE OF ISSUANCE OF AMENDMENT TO PROVISIONxL OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 56 to Provisional Operating License No. DPR-20, issued to Consumers Power Company (the licensee), which revised the license for operation of the Palisades Plant (the facility) located in Covert Township, Van Buren County, Michigan. The amendment is effective as of its date of issuance.
The amendment incorporates a new License Condition (Paragraph 3.G) into License DPR-20 to allow performance of a feedwater line water hammer test.
The application for the amendment canplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Prior public notice of this action was not required since the amendment does not involve a significant hazards consideration.
The Cannission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of this amendment.
7590-01
,. For further details with respect to this action, see (1) the app'.1 cation for amendment dated April 11, 1980, and a supplement thereto dat2a April 16, 1980, (2) Amendment No. 56 to License No. DPR-20, and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Comission's Public Document Room,1717 H Street, N. W., Washington, D. C. and at the Kalamazoo Public Library, 315 South Rose Street, Kalamazoo, Michigan 49006. A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, Attention: Director.
Division of Operating Reactors.
Dated at Bethesda, Maryland, this 30th day of April,1980.
FOR THE NUCLEAR REGULATORY COMMISSION Dennis M. Crutchfield Division of Operating Reactors
NUCLEAR REGULATORY COMMISSION SAFETY EVALUATION REPORT STEAM GENERATOR WATER HAMMER PALISADES PLANT e
DOCKET NO. 50-255
1.0 INTRODUCTION
Steam generator water hammer has occurred in certain nuclear power plants as a result of the rapid condensation of steam in a steam generator feedwater line and the consequent acceleration of a slug of water which upon impact within the piping system causes undue stresses in the piping and its support system.
The significance of these events varies from plant to plant.
Since a total loss of feedwater could affect the ability of the plant to cool down j
af ter a reactor shutdown, the NRC is concerned about these events occurring, even though an event with potentially serious consequences is unlikely to happen.
Because of the continuing occurrence of water harmer events, the NRC, in September 1977, informed all PWR licensees that water hammer events due to the rapid condensation of steam in the feed-water lines of steam generators represented a safety concern and that further actions by licensees for Westinghouse and Combustion Engineering designed nuclear steam supply systems are warranted to assure that an acceptably low risk to public safety due to such events is maintained. Accordingly, these licensees were requested to submit proposed hardware and/or procedural modifications, if any, which would be necessary to assure that the feedwater lines and feedrings remain filled with water during normal as well as transient operating conditions. At the same time, the NRC provided each PWR licensee with a copy of its consultant's report, "An Evaluation of PWR Steam Generator Water Hammer," NUREG-0291.
A steam generator water hammer. event has not occurred at the Palisades j
Plant. Operation of the plant began in 1971 and in January of 1977 precautionary procedures were adopted to avoid possible steam generator water hammer by limiting the flow of auxiliary feedwater to each steam generator to 100 gallons per minute (gpm). This limit was raised to 150 gpm per steam generator in October 1978. Because of the unique geometry of the feedwater piping at the Palisades Plant that provides the potential for a severe steam generator water hammer, it is consi-dered prudent to require this licensee to limit the flow of auxiliary feedwater to either those values previously experienced at the Palisades Plant during repeated normal operations or those values experienced during special steam generator water hammer tests to be performed at the Palisades Plant.
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.. 2.0 EVALUATION Our consultant, EG&G Idaho, Inc., prepared the attached evaluation of steam generator water hamer at the Palisades Plant as part of our technical assistance program.
(letter from J. A.
Dearien EG&G,
to R.E. Tiller, DOE, dated February 27,1980.) We have reviewed this report together with the licensee's submittals listed under item 4.0.
Our consultant concluded that based on its review of the Palisades Plant operating history and the present mode of operation (with limited auxiliary feedwater flow) the potential for damaging steam generator.
water hammer is sufficiently low to permit continued operation of this facility.
We concur with our consultant's conclusion.
The licensee in its letters of November 19, 1979, and January 21, 1980, indicated that planned modifications to the auxiliary feedwater system will provide for (1) the automatic initiation of auxiliary feedwater flow, and (2) the control of flow to each steam generator to within a predetermined amount using an automatic flow controller. The licensee is planning to perform tests for steam generator water hamer at the Palisades Plant.
It proposes to test with auxiliary feedwater flow rates up to 400 gpmin order to establish a safe range of allowable flow rates.
We have reviewed the licensee's plans for the automatic operation of the auxiliary feedwater system; and we have concluded that such opera-tion, including automatic initiation and flow limitation, could reduce the potential for the occurrence of steam generator water hammer.
However, increasing the auxiliary feedwater flow limit above 150 gpm might increase the potential for steam generator water hammer. We have reviewed the licensee's test procedures titled "Feedwater Line Water Hammer Test" T-130 Revision Zero and have found that the s.ccessful performance of this test would provide an adequate basis for establishing a safe range for the control of auxii'wv feedwater flow with regard to steam generator water hammer.
3.0 CONCLUSION
Based on our knowledge of water hamer phenomena, and our review of the licensee's responses and the enclosed evaluation report, we concur with our consultant's conclusion that the potential for damaging steam generatorwaterhammerissufficientlylowtopermjtcontinuedoperation of this facility.
However, even though steam generator water hammer is not likely to occur, the licensee should be vigilant and monitor for water hanmers that might impose significant stresses on the pipinq systems or their supports. We will continue to monit or r eport. from
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this licensee for indications of possible water hamer.
If such indica-tions appear in the future, this matter will be reexamined and may result in additional requirements to reduce the probability of steam generator water hammer at this facility.
Based on our review of the licensee's plans for automatic initiation and automatic limitation of auxiliary feedwater flow, we have found that these measures will tend to reduce the probability of occurrence of a steam generator wate r hammer. Our review of the procedures for a feedwater line water hammer test shows that the successful performance of this test would provide an adequate basis for establishing a safe range for the control of auxiliary feedwater flow. These modifica-tions and procedures are, therefore, acceptable to the staff for the purpose of reducing the potential for steam generator water hammer.
We, therefore, find that steam generator water hamer at Palisades Plant presents no undue risk to the health and safety of the public.
4.0 REFERENCES
4.1 W.E. Bennett, Waterhammer in Steam Generator Feedwater Lines, Westinghouse Technical Bulletin, NSD-TB-75-7, June 10,1975.
4.2 R.B. Sewell. (CPC) " Palisades Plant," letter to DRL (NRC),
July 16,1975.
4.3 J.B. Block, et al, An Evaluation of PWR Steam Generator Water ~ ~ ~ ~
Hamer, Creare, Inc., NUREGT231, December 1976.
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4.4 Nuclear Services Corporation Evaluation of Auxiliary Feedwater and Water Hamer Potential at Palisades Plant, CPC-OlT67, F,ebruary 24, 1977.
4.5 A. Schwencer (NRC) letter to D. Bixel (CPC), " Steam Generator Water Hamer" dated September 2,1977.
4.6 D.P. Hoffman, (CPC) "Feedwater Line Water Har.m r," letter to A. Schwencer, (NRC) January 24, 1978.
4.7 D.P. Hoffman, (CPC) " Palisades Plant - Feedwater Line Water Hammer," letter to D.L. Ziemann (NRC), August 9,1978.
4.8 D.L. Ziemann, (NRC), " Docket 50-255," letter to D. Bixel (CPC),
September 22, 1978.
4.9 R.B. DeWitt, (CPC), "Palisachis Plant. - Addi tional In forma t ion Pertaining to NRC List dated May 4,1979," letter to D.I.. /iemann (NRC),May 10, 1979.
N 4-Q 4.10 NRC Staff, Water Hammer in Nuclear Pdwer Plants, NUREG-0582 July 1979.
4.11 D.L. Ziemann (NRC) letter to D. Bixel (CPC) " Steam Generator Water Hammer" dated September 12, 1979.
4.12 S.R. Frost, (CPC), " Palisades Plant - Steam Generator Water Hamer Response," letter to D.L. Ziemann (NRC), November 19, 1979.
4.13 R.W. Huston (CPC) letter to Director, NRR, Attn:
D.L. Ziemann
" Automatic Initiation of Auxiliary Feedwater System dated January 21, 1980.
4.14 Feedwater Line Water Hamec Test T-130 Revision Zero Date: April 30,1980
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,u STEAM GENERATOR WATER HAMMER TECt%iCAL EVALUATION PALISADES PLANT February 1980 EG&G Idaho, Inc,
CONTE'NTS Page I. INTRODUCTION.........................
1 II. WATER HAMMER EXPERIENCE...................
2 III. MEANS TO REDUCE THE POTENTIAL FOR WATER HAMMER........
3 IV. FEEDWATER SYSTEM DESCRIPTION AND OPERATION.
5 5
1.
Main Feedwater System..
2.
Auxiliary Feedwater System.
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'3.
Feedwater Piping and Sparger Description....
7 V. OPERATING EXPERIEtiCE AND WATER HAMMER SUSCEPTIBILITY.....
10 VI. CONCLUSIONS.........................
13 VII. REFERENCES..........................
14 TA?l.ES
- 1. Palisojes Plant Trips Due to Loss-of-Feedwater....
15 II. Palisades Plant Trips Due to Loss of Off-site Power.....
17 III. Palisades Plant Trip History Resulting From Low Steam Generator Level on Loss of Feedwater (1972)......
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I.
INTRCDUCTION A review of the Palisades Plant feedwater system was pericrmed.
The purpose of this review was to assess the susceptibility of the feedwater system to' water hammer during operating transients that coulo result in conditions conducive to water hammer.
Steam-water slugging in the steam generator feedrings and adjacent feedwater piping, generally referred to as steam generator water hartner, was considered in this review.
The Palisades Plant is one of these facilities which has bottom discharge steam generator feedwater spargers. Further, there have been no reported incidents of steam generator water hammer. '
Unusual features related to feedwater piping include a horizontal run of 28 feet before entering the steam generator and a 2 feet downward ' jog' inside tha steam generator befcre teeing into the sparger.
The information for. this evaluation was obtained from:
1) discussions with the licensee, 2) licensee submittals to f>RC,,6,7,8, 3) the Palisades Plant " Final Safety Analysis Report"', 4) "An Evaluation of PWR Steam Generator Water Hammer" NUREG-0291, 5) " Westinghouse Technical Bulletin, NSD-TB-75-7"#
3 and 6) NRC correspondence and reports.
A review of Palisades Plant steam generator water harrmer experience is presented in Section II.
The means to reduce the potential for water hammer at this fac ility are presented in Section III.Section IV presents a description of the feedwater system including a description of the feedwater piping and sparger.
Section V presents a description and tabulation of available operating transients and situations that could result in conditions conducive to water nammer. Conclusions are presented in Section VI concerning the susceptibility of the Palisades Plant feedwater system to steam generator water hammer.
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II.
WATER HAMMER EXPERIENCE There have been no reported steam generator water hammer incidents at the Palisades Plant during its pe.riod of operation.1,2 Available information,,5 indicates that there were 33 events (28 loss-of-feeowater and 5 loss-of-offsite power events) over the operating period of the Palisades Plant when feedwater sparger uncovery was likely and conditions conducive to steam ge.3eratcr water hammer existed. Of the total of 33 incidents,13 events occurred prior to any administrative controls relative to feedwater admission and sparger recovery limitations and included 7 loss-of-feedwater events where sparger uncovery and substantial or complete drainage were definitely known to have occurred. Although such conoitions are normally considered conducive to steam generator water hammer, none occurred.
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III. MEANS TO REDUCE THE POTENTIAL FOR WATER HAMMER Based on past experience at other plants, steam generator water hammer would be most'likely to occur during startup,. shutdown, and low power situations when feedwater is under manual control and the flow' rates are insufficient to maintain unsavered feedrings full of water.
To avoid admission of steam and possibly slugging during these operational conditions, feedrings with bottom discharge must remain covered with water.
As a means to reduce'the potential for damaging steam generator water hammer at the Palisaces Plant, administrative controls were established and are currently in effect that require operators to maintain specified steam generator feedwater flow. rates, delay times, and la. vel control in the steam generator.
Initially, based on a consultant's report, administrative procedures were adopted to limit steam generator feeawater ficw rate to 100 gpm effective January 10, 1977.
In August 1978, in order to meet anticipated decay heat removal requirements,S an increase to 4
10 150 gpm was proposed and approved effective October 4, 1978.
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The consultant's recommendations were based on a series of 7 4
loss-of-feedwater events in 1972 that resulted in the feedwater spargers inside tne steam generators being uncovered (see Section V).
10 The administrative controls currently in force for the use of the 150-gpm feedwater flow limit are defined as follows:
"The sparger has been uncovered for at least 15 minutes.
j Tne trip occurred at greater than 85% power.
i The steam generator level is maintained at least two feet below the sparger center line.
i We interpret this to mean that a reflood rate no greater than 150 gpm.may be-employed while the steam generator level is more f
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than tw3 feet below the sparger center line; that a rato no greater tnan 100 gpm may be employed while the steam generator is within two feet below the center line of the sparger; and that 1
the reflooding rate would be unlimited provided the sparger is full of water."
The probability of steam generator water har ner was considered to be acceptably low fer'ncrmal operations and plant trip situations while operating the main feedwater and auxiliary faedwater systems in,the above manner.10 From an operational viewpoint and to minimize operator error, higher AFW flow rates and simplification of existing administrative controls are certainly desirable.
However, the extent of the available operat 1g experience, '
tne basis of the consultants report, the lack of any of the recommended features '
incorperated in the feedwater piping, and the lack of any special tests which establisnes the absence of water hammer preclude such considerations at this time. Baseo on operating experience available, it is concluded that existing administrative controls are an effective means to reduce the potential for steam generator water hacmer and should be retained.
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IV. FEE 0 WATER SYSTEM DESCRIPTION AND OPERATION This section gives a brief description of both main and auxiliary feedwater systems and the attendant operation of each for the Palisades Plant. A description of the feedwater piping arrangement and tne feedwater sparger is included.
1.
MAIN TU DWATER SYSTEM For the main feedwater (MFW) system, condensate from the condenser hot well is pumped by two half-capacity, electric motor-driven condensate pumps through the air ejector and gland steam condensers, then through two parallel trains of five stages of low-pressure feed water heaters to the suction of two half-capacity, turbine-driven MFW pumps.
The MFW is next pumped tnrough two parallel sets of feedwater regulating valves and single stage high-pressure feedwater heaters and tnen to each of the steain generators.
The main feedwater line and sparger are filled with 419 F water at full load conditions.
Steam for the MFW pump turbine orivers is normally taken j
from the main turbine crossover piping. A cross-connection to the main steam system provides autortatic backup steam for plant low load operation.
Both MFW pumps are used to furnish the feedwater flow rate required at unit loads greater than 50%.
The system is designed to permit operation with one feed pump under all modes at reduced unit load.
Each pump is rated at 13,500 gpm with a total developed head j
(TOH) of 2640 feet at 5000 rpm. Each turbine driver and pump must be set up locally and brought up to speed before the driver can be controlled from the main control room.
If MFW pump suction pressure falls below a preset critical value, the pump will be automatically tripped.
The turbine drivers will also be tripped from thrust-bearing failure, low turbine exhaust vacuum, reverse rotation, excessive vibration, and loss of a condensate pump (trip of a preselected turbine driver).
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The MFW controls maintain steam g:nerator downcomer level within acceptable limits by positioning the feed water regulating valves supplying each steam generator. The speed of the turbine-driven main feedwater pumps will also be controlled by the feedwater controls.
MFW pump speed control is by autcmatic or manual means.
Automatic control of MFW is provided when the plant is above 15% power.
Steam flow, feedwater flow and downcomer level are used in a three-element controller on each steam generator to maintain preset water level during steady-state and transieot operation.
Manual control of MFW flow may be assumed by the operator at any time.
At normal operating conditions on receipt of a reactor / turbine trip signal, the turbine driver speed is ramped down from about 5000' rpm to about 3200 rpm within 60 seconds The attendant MFW pump head is thus reduced from about 2300 feet to about 1000 feet (350 psi). As a result of main steam isolation caused by the turbine trip, the steam generator pressure rises from 700 to 900 psi which is the set-point of the steam dump and turoine by-pass valves.
Since the MFW pump is pumping against a back pressure tnat is in excess of the developed pump head, no MFW flow enters the steam generator.
The MFW pumps are then manually tripped and AFW is promptly initiated typically within 1-1/2 minutes of the turbine trip.
The feedwater temperature gradually decreases to that of the AFW stared condensate in the condensate storage tank (CST).
In the event of a design basis accident (DSA), the main feed pumps will be tripped from low-condensate header pressure which will result from shedding of tne condensate pumps from their supply buses.
The motor-driven auxiliary feed pump will be available for service at the operator's discretion.
2.
AUXILIARY FEEDWATER SYSTEM The auxiliary feedwater (AFW) system provides unheated water to the steam generators during piant startup and shutdown operations and wnen the main feedwater (MFW) system pumps are isolated.
Two AFW 6
pumps are pr:vided (one of which is redundant). One electric motor-driven and one steam turbine-driven pumo take suction from the 125,000 gallon C3T.
In tne event of a loss or depletion of the CST water supply, the backup water supply from the fire system car be utilized by starting one of the fire pumcs.
Each AFW pump has a capacity of 415 gpm with a TOH of 2733 feet.
On start-up, the motor-driven AFW pump operates to provide feedwater to the steam generators until sufficient steam can be generated to operate the MFW pump turbine drivers.
The level in the steam generators is maintained by remotely adjusting the AFW control valves in each respective steam generator suxiliary feed header.
The AFW
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system also supplies water to the steam generators to remove decay heat during the initial phase of primary system cooldown.
For any condition during which MFW to the steam generators is interrupted and the reactor is tripped, sufficient feedwater flow is maintained by tne l
motor-driven or the turbine-driven AFW pumps to remove decay heat from the primary system.
i The condensate pumps may be used to pump water through the normal feedwater train to the steam generatcrs in the event of a failure of the AFW piping system.
The steam generator pressure may be relieved by the steam dump system to accommodate this mode of operation.
In the event of a steam line break, the main feedwater pumps are inoperative.
The turbine and motor driven AFW pumps are available to be used to maintain shutdown cooling flow to one intact loop steam generator.
The AFW needed for the various, plant operating conditions is such that one pump can supply all of the necessary water requirements.
1 3.
FEEDWATER PIPING AND SPARGER DESCRIPTION 4
The MFW lines are IE-in. diameter and AFW lines are 6-in.
diameter..Both are Schedule 80, seamless carbon steel, ASTM A-106,-
Grade B, and 'are designed to meet the requirements of ANSI B31.1.
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The MFW check valves are loc 6ted en a horizental run at elevation 603 f t-0 in. approximately 11 feet outside of containment penetrations seven and eight. The AFW branch connections tee into the MFW lines juf,t downstrea'n of the check valves and outside of the containment.
The AFW check valves are immediately upstream of the branch c.cnnection (injection location).
The MFW lines run parallel, approximately 11 feet apart in an east-southeast direction through the contair. ment penetrations, and continue in the same direction for approximately 35 feet.
The lines then drop to elevation 601 f t-0 in..
The "A" line turns 90 degrees to the south-southwest and the "B" line turns 90 degrees to the north-northeast.
Each runs horizontally thrcvgh straightening vanes, and flow elements, for 3 total length of apprcxi.nately 46 ft.-6 in.
Both lines turn 90 degrees upward and rise to an elevation of 653 f t.-9 5/8 in.
They then both turn 90 degrees to the east-southeast and run horizontally for 20 feet.
The "A" line turns 90 degrees to the north-northeast end tne "S" line turns 90 degrees to the south-southwest, and they both run horizontally for approximately eight feet to tne steam generater nozzles. After passing through the steam generator walls, the pipe " jogs" two feet downward to connect to a ring-shaped feedwater sparger. The sparger feedring is 100 inches in diameter and has bottom discharge.
The spargers (one per steam generator) are 12-inch, Schedule 40, pipe and are located in the downtomer annulus at the appr0ximate level of the top of the tube bandle.
There are 68 bottom discharge orifices (on each sparger) which are formed by we! ding on 5-inch long,1-1/4-inch diameter, Schedule 40, pipe nipples.
As described, the feedwater oiping arrangement contains neither of the two recomenaations '
to prevent or abate water hammer influences, e.g., a pipe' loop seal nor a snort horizontal pipe run preceding the feedwater nozzle through the steam generator wall. The Palisades Plant has the longest known run of horizontal pipe length of any operating P'a'R f acility, i.e., 28 feet. As visualized in a horizontal plane or plan view, this section of.feedwater piping is "L" 8
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9 shaped with a 'short-leg' nearest the stgam generator 8 feet long and a 'long-leg' 20 feet in length before turning 90 degrees downwr.' in ii
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the upstream direction.
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V.
OPERATIriG EXPERIENCE Atto WATER HAMMER SUSCEPTIBILITY The conditions ccasidered most conducive to steam generator water hamer occur when the steam generator feedrings are uncovered ar.d steam enters the feedrings and attached horizontal feeowater piping.3 Steam-water slugging and subsequent water hanner may occur when incoming feedwater mixes with the steam in the piping and rapid condensation occurs.
The conditions can be avoided by keeping the feedrings and associated piping full of water.
This can be accomplished by 1) keeping the water levels in the steam generators above the feedrings or 2) supplying feedwater at a higher flow rate than the rate at which feedsater orains through the discharge holes on the bottom of uncovered feedrings. The drop or "shrickage" in steam generator water level and suosequent feedring wr.cc4ery is the respit of interrupted reactor power production causing the collapse of steem voids within the secor.dary side of the steam generators.
A similar situation would also be experienced during events such as loss of main feedwater, loss of effsite power, steam line break, or less-of-coolant accident.
The expected water level behavior immediately following a loss of feedsater or loss of cff-site power is not altereo by reccvery rates.
Auxiliary feedwater is manually initiated following the trip.
Actual f
steam generator level recovery is controlled by resconse time of the operator and feed flow - steam flow mismatch prior to the trip.
The normal steam generator water level in the downcomer is 72 inches above the sparger center iir.e (65% of narrow range).
The steam generator low water level setpoint is at the sparger center line (24.7% of narrew range). The zero percent narrow range tap is located 44 inches below the sparger centerline.
Slow recovery rates were realized after the imposition of the 100-gpm fee 6ater flow limit and as such reflect conditions realized fcr prevention of steam generator water ha:aner.
Later, the 100-gpm recovery flow rate was found to be marginally acceptable to remove :ecay heat and reestablish water level, and 150-pm recovery rates as given in Section III were instituted.
10
Tables I and II identify situations concerning loss of fcodwater which have resulted during the operational period since and including the year 1972.2 These situations reflect conditions in which the sparger was uncovered in one or both steam generators and significant Such or complete drainage occurred as typified in Table III.
conditions are normally expected and considered conducive to steam generator water hammer prior to tne establishment of administrative controls relative to feedwater admission beginning January 10, 1977.
Those events thereafter basically reflect the effectiveness of the existing administrative controls.
It is apparent from the review of feedring uncovery events that it is not possible to avoid drainage of the feedrings and adjacent piping by keeping the feecrings covered with water.
Tne uncovery time not' only varies among the events but between the A and B steam generators.
Complete drainage of the feedrings and adjacent piping has likely occurred.
The time required for complete drainage is unknown, but is probably on the oroer of a few minutes. As shown in Table III, one or both feedrings have oeen uncovered on four occasions for periods of 21 to 113 minutes which should be ample time for drainage.
As mentioned at the beginning of this section, the alternative to continuous coverage of the feedrings with water is to maintain sufficient feedwater flow through uncovered feedrings to keep the feedrings full of water.
To do this, the main feedwater flow required for Palisades Plant is roughly calculated to be 2000 gpm per steam generator. As such, based on the design main feedwater flow of about 13,500 gpm per steam generator, tne flow required to keep the feedrings full of water is about 15% of design flow.
The feedrings, if uncovered, would not be kept full of water below power levels of about 15% of full power, and the AFW system is not capable of keeping the piping and sparger full.
11
Auxiliary feedwater flow is administratively limited to less than 150 gpm or 100 gpm per stream generator as explained in Section !!I.
Plant experience during a number of complete loss of main feedwater events in Tables I, II and III have shown that this flow is adequate for reactor coolant pump and decay heat removal provided both steam generators are in service.
Although s team generator narrow range level indication has, at times, been lost following a complete loss of main feedwater event, adequate removal of heat from the primary system was always maintained while feeding at the administrative limit.
The influence of the 2-foot " jog" downward of the feedwater piping inside the steam generators (before teeing into the sparger) is not firmly establishdd.
A consultants' report indicated no apparent influence either for prevention or inducement of steam generator water hammer.
As indicated previously, water hammer would be most likely during startup, shutdown, and low power operation since feedring uncovery events are frequent due to manual feedwater control and feedwater flow requirements are insufficient to keep the feedrings full of water.
However, the experience at this facility indicates an apparently low susceptibility to water hammer under conditions normally considered conducive to steam generatcr water hammer.
Prior to instituting administrative controls on feedwater admission, there were 13 loss-of-feedwater events that covered a broad spectrum of plant operating conditions from low power (start-up and hot stand-by) to high power levels (93%).
The reason for the Palisades Plant low susceptibility is presently unknown and undetermined.
1 S
l 12
VI.
CONCLUSIONS We have reviewed the operating history of the Palisades plant pertinent to steam generator water hammer and the related operational and procedural characteristics of the feedwater system.
The review has shown that conditions conducive to steam generator water hammer have occurred at the Palisades plant but no water hammer events have been observed.
The conditions are encountered during normal operating transients and during startup and shutdown operations.
Such ccnditions would also be expected in the future during the normal and accident operating situations addressed in the review.
Based on this review and the present mode of plant operation, we have concluded that the potential for damaging steam generator water hammer is sufficiently low to permit continued operation of this facility.
J 13 l
VII. REFERENCES 1.
R. B. Sewell, (CPC) " Palisades Plant," letter to DRL (NRC),
July 16, 1975.
2.
S. R. Frost, (CPC), " Palisades Plant - Steam Generator Water Hammer Response," ltr. to D. L. Zieman (NRC), Nove-ber 19, 1979.
3.
J. B. Block, et al, An Evaluation of PWR Steam Generator Water Hammer, Creare, Inc., NUREG-0291, December 1976.
4.
W. E. Bennett, Waterhammer in Steam Generator Feedwater Lines, Westingnouse Technical Bulletin, NSD-TB-75-7, June 10,1975.
5.
Nuclear Services Ccrporation, Evaluation of Auxiliary Feedwater and Water Hammer Potential at Palisades Plant, CPC-01-07, February 24, 1977.
6.
D. P. Hoffman, (CPC) "FeedWater Line Water Hammce," letter to A. Schwencer (NRC), January 24, 1978.
7.
D. P. Hoffman, (CPC), " Palisades Plant, - Feedwater Line Water Hammer," letter to D. L. Zieman (NRC), August 9,1978.
8.
R. B. DeWitt, (CPC), " Palisades Plant, - Additional Information Pertaining to NRC List Dated May 4, 1979," letter to D. L. Zieman (NRC), May 10, 1979.
9.
Final Safety Ana',ysis Report, Consumers Power Company, Palisades Plant, Docket 60-255.
10.
D. L. Zieman, (NRC), " Docket 50-255," letter to D. Bixel (CPC),
September 22, 1978.
11.
NRC Staff, Water Hammer in Nuclear Power Plants,.*iUREG-0582, July 1979.
i 4
14
TABLE I f
PALISADES PLANT TRIPS DUE TO LOSS OF FEEDh'ATER Trip Pcwer
- acer Date level Cause
~2-01 1-11-72 5%
Feedwater pump trip due to sharp increase in desand.
Operator manually opened the rag valve fully.
(See Event 72-04.)
~2-03 1-12-72 5%
Feedwater pump trip due to oscillations in suction pressure.
Low section pressure trip.
(See Event 72-04.)
t
'2-04 1-13-72 Hot Feedwater pump trip due to low suction pressure.
Standby Fine mesh start-up strainer had been lef t in the suction.
- -06 2-03-72 20%
Feeddater reg valve f ailure.
2-10 3-27-72 60%
Feedwater pump trip due to high vibration.
Oil filter change out procedure induced air into oil system.
'i-14 4-14-72 15%
Feedwater reg valve f ailure.
'?-20 7-31-72 18%
Defective feedwater pump vibration sensor caused a pump trip.
'?-23 12-21-72 32%
Inadvertent closing of feedwater re5 valve.
~5-04 6-30-75 20%
Feedwater pump trip, rapid demand in feed-water flow caused low puma suction pressure.
'5-06 7-29-75 Mot Feedw iter pump trip, low suction pressure.
Standby i-02 5-10-76 25%
Feedwater pump speef control ramped pump to minimum rpm thus insufficient discharge head.
' ' 03 1-17-77*
100%
Feedwater pump trip, low suction pressure, dump valve on moisture separator drain tank f ailed open.
04 1-18.77 35%
Feedwater pump trip, cause unknown.
(See Event 77-08.)
~ 07 3-24-77 90%
Feedwater pump trip, cause unknown.
(See Event 77-08.)
1 Beginning 1-10-77 auxiliary feedwater flow rate restricted to 100 gpm.
15
1 TABLE I (Contd)
+-
~_
~ ip Power r
m3er Date Level Cause
'7-08 3-27-77 82%
Feedwat'
? ump trip, faulty low-pressure switch
. condensate pump disch.
77-21 11-27-77 50%
Feedwater reg valve closed, coerator error whi' transferring from manual to auto.
73-02 4-21-78 50%
Feedwater pump trip, defective vibration sensor.
73-04 5-11-79 Hot Feedwater pump trip, low suction pressure, Standby condensate polisher strainer plugged with powder.
'3-09 6-07-78 23%
Feedwater reg valve failed to provide flow.
(See Event 78-10.)
'3-10 6-08-78 20%
Feedwater stop valve not opened during start-up sequence.
~3-1; 6-13-78 83%
Feedwater pump trip, severe transient induced by Technician error while maintaining feeowater flow recorder.
73-18 9-19-78 86%
Feedwater pump trip, turbine governor failure.
73-24 10-17-78**
84 %
Feedwater pump trip, axial position, operator failed to warm LP STF line prior to cutting it in.
'3-27 12-18-78 88%
Feedwater pump trip, cause unknown.
'?-03 3-03-79 100%
Feedwater pumps tripped, low suction pressure, moisture separator drain tank dump valve failed open.
3-04 4-07-79 100%
Feedwater pump trip, cause unknown.
Pump vibration trip removed from service.
'9-09 8-10-79 88%
Feedwater pumps tripped, high axial thrust incorrect main turbine valve test procedure.
79-10 8-24-79 91%
Feedwater pump tripped, low suction pressure, valving error while valving in the condensate demineralizers.
Seginning 10-04-78, increased recovery flow rate to 150 gpa while sparger is uncovered.
Th3LE II
~
PALISADES PLANT TRIPS DUE TO LOSS OF 0FF-SITE POWER Event Power Nem3er Date Level Cause 72-17 4-15-72 15%
Lightning strcxe to tie line.
76-05 7-20-76 93%
Lightning storm, unit ground relay operated.
77-17**
9-24-77*
85%
Lightning stroke caused "R" bus to clear.
77-20**
11-25-77 85%
"R" bus cleared cause unknown.
78-13 6-18-78 84%
Lightning stroke.
Beginning 1-10-77, auxiliary feedwater flow rate restricted to 100 gpm.
After 10-04-78 recovery flow rate increased to 150 gpm.
Only events 77-17 and 77-20 actually resulted in a loss of power.
The other three events caused a transient severe enough to cause a plant trip.
17 l
l
s o
TABLE 111
+
PALISADES PLANT TRIP lilSTORY RESilLTING FROM LOW STEAM GENERAIOR LEVEL l
ON LOSS-OF-FEEDWATER (1972)
Reflood Rate Time Below Minimum Through Sparger, Feed water Level During Trip Steam Steam Generator Level, Sparger Transient Number Cause of Trip Generator inches / minute (gpm)
(minutes)
(inches) 72-01 feedwater pump trip A
4.5 (677)
<5
-8 8
9*
(1354)*
<5
-8 72-04 Feedwater pump trip A
N/A B
1.39 (209) 21
-24 72-06 Feedwater regulating valve failure A
9*
(1354)*
<5
-16
+4 6 8
5 72 Feedwater pump trip A
0.32 (48) 60
> -48 0
0.49 (74) 113
> -4 8 72-14 Feedwater regulating valve failure A
9*
(1354)*
<5
-28
+38 B
72-20 Feedwater pump trip A
3.01 (453)
<5
-14
~
469 e
B 72-23 feedwater regulating valve closure A
0.76 (114) 69
-44
<5
-6 B
- Note:
The maximum detectable.
I