ML19344A504
| ML19344A504 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 07/24/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8008200518 | |
| Download: ML19344A504 (1) | |
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- p ug'sg UNITED STATES 8"
( n NUCLEAR REGULATORY COMMISSION fJC 3
E REGION 111 "o
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799 ROOSEVELT ROAD GLEN ELLYN. ILLINOIS 60137
'JUL 2 4 880 Docket No. 50-329 Docket No. 50-330 Consumers Power Company
. ATTN:
Mr. Stephen H. Howell Vice President 1945 West Parnall Road Jackson, MI 49201 Gentlemen:
The enclosed IE Ialletin No. 80-18 is forwarded to you for your information. Although no written response is required at this time, these concerns will be addressed as part of the licensing process for your plant.
If you desire additional information regarding this matter, please contact this office.
Sincerely, bf f James G. Keppler Director
Enclosure:
IE Bulletin No. 80-18 cc w/ encl:
Central Files Director, NRR/DPM Director, NRR/ DOR PDR l
TIC Ronald Callen, Michigan Public
- Service Commission Myron M. Cherry, Chicago 1
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SSINS No.: 6820 Accession No.:
UNITED STATES 8005050062 NUCLEAR REGULATORY COM415510N
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OFFICE OF INSPECTION AND ENFORCEMENT g d
uI WASHINGTON, D.C. 20555 July 24, 1980 IE Bulletin No. 80-18 MAINTENANCE OF ADEQUATE MINIMUM FLOV THRU CENTRIFUGAL CHARGING PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUPTURE Description of Circumstances:
Letters similar to the May 8, 1980 notification made pursuant to Title 10 CFR Part 21 (enclosure) were sent from Westinghouse to a number of operating plants and plants under constructicn (list, within enclosure) in early May, 1980.
The letters and the enclosed "Part 21" letter contain a complete description of the potential problem summarized below. The letters indicated that under certa;n conditions the centrifugal charging pumps (CCPs) could be damaged due to lack of minimum flow before presently applicable safety injection (SI) termination criteria are met. The particular circumstances that could result in damage vary somewhat from plant to plant, but involve unavail-ability of the pressurizer power operated relief valves (PORVs), with operation of one or more CCPs repressurizing the reactor during SI following a secondary system high energy line break. Since the SI signal automatically isolates the CCP mini-flow return line, the flow through the CCPs is determined by the individual pump characteristic head vs. flow curve, the pressurizer safety valve setpoint, and the flow resistances and pressure losses in the piping and in the reactor core. That minimum flow say not be adequate to insure pump cooling, and resulting pump damage could violate design criteria before current SI termination criteria are met.
Westinghouse recommends that plant specific calculations outlined in the letter (enclosure) be performed to determine if adequate minimum flow is assured under all conditions. If adequate minimum flow is not assured, Westinghouse recommends specific equipment and procedure modifications which will result in adequate minimus flow. The recommended modifications assure availability of the necessary minimum flow by assuring that the mini-flow bypass line will be open when needed, but will be closed at lower pressures when the extra flow resulting from bypass line closure might be necessary for core cooling.
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1 IE Bulletin No. 80-18 July 24, 1980 Page 2 of 3 Actiors to be taken by PWR licensees listed in the enclosure as " operating plants," and those listed as "non-operating plants" which are nearing licensing
- are listed below:
1.
Perform the calculations, outlined in the enclosure, for yodr plant.
2.
If availability of minimum cooling flow for the CCPs is not assured for all conditions by the calculations in 1:
Make modifications to equipment and/or procedures, such as those a.
suggested in the enclosure, to insure availability of adequate minimum flow under all conditions.
If modifications are made as described in the attachment for interim modification II, verify that the Volume Control Tank Relief Valve is operable and will actuate at its design setpoint.
Justify that any manual actions necessary to assure adequate minimum b.
flow for any transient or accident requiring SI can and will be accomplished in the time necessary.
Verify that any manipulations required (valve opening or closing, c.
along with the instrumentation necessary to indicate need for the action or accomplishment of the action, etc.) can be accomplished without 9ffsite power available.
Justify that flow available from the CCPs with the modifications in d.
place will be sufficient to justify continued applicability of any safety related analyses which take credit for flow from the CCPs (LOCA,HELB,etc.).
Justify that all Technical Specifications based on the Item 2.d e.
analyses remain valid.
Provide the results of calculations performed under Item 1, and describe 3.
any modifications made as a result of Item 2 (include the justifications requested).
Actions to be taken by PWR licensees not listed in the enclosure are listed below:
In a quantitative manner similar ta 1 above, determine whether or not 1.
minimum cooling is provided to centrifugal pumps used for high pressure injection, for all conditions requiring SI, prior to satisfying SI "Those listed in t.he enclosure considered to be " nearing licensing" are:
North Anna 2, Diablo Canyon 1, McGuire 1, Salem 2, and Sequoyah. These plants must respond in writing within the specified time. Other non-licensed plants whether or not listed in the enclosure, are not required to submit a written response at this time.
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IE Bulletin No. 80-18 July 24, 1980 Page 3 of 3 termination criteria.
If a " minimum flow bypass" line is present which remains open during high pressure injection, and if that line. guarantees that minimum cooling flow will be provided to the pumps under. such condi-tions, then no further calculations are required if all safety related analyses (Item 2.d above) assumed presence of the open line. '
2.
Same as 2 above.
3.
Same as 3 above.
Licensees of all operating PWR power reactor facilities and those nearing licensing
- shall submit the information requested within 60 days of the date of this letter. Include in your response to this Bulletin, (a) your schedule for any changes proposed, (b) if reactor operation is to continue prior to completion of the proposed changes, include your justification for continued operation.
Reports shall be submitted to the Director of the appropirate NRC Regional Office and a copy forwarded to the Director, NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D. C. 20555.
Approved by GAO, B280225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems.
Enclosure:
W Ltr from T. M. Anderson,/5/80 to V. Stello, IE dtd 5 "Those considered to be " nearing licensing" are: North Anna 2, Diablo Canyon 1 McGuire, Salem 2, and Sequoyah.
Enclosure IE Bulletin No. 80-18 July 24, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To Mg.
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Supplement 2 Failures Revealed by 7/22/80 All BWR power reactor facilities holding OLs to 80-17 Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR Supplement 1 Failure of Control Rods 7/18/80 All BWR power reactor to 80-17 to Insert During a Scram facilities holding OLs at a BWR 80-17 Failure of Control Rods 7/3/80 All BWR power reactor to Insert During a Scram facilities holding OLs at a BWR 80-16 Potential Misapplication of 6/27/80 All Power Reactor Rosemount Inc., Models 1151 Facilities with an and 1152 Pressure Transmitters OL or a CP with Either "A" or "D" Output Codes 80-15 Possible Loss Of Hotline 6/18/80 All nuclear facilities With Loss Of Off-Site Power holding OLs 80-14 Degradation of Scram 6/12/80 All BWR's with an OL Discharge Volume Capability 80-13 Cracking In Core Spray 5/12/80 All BWR's with an OL f
Spargers 80-12 Decay Heat Removal System 5/9/80 Each PWR with an OL Operability 80-11 Masonry Wall Design 5/8/80 All power reactor facilities with an OL, except Trojan 80-10 Contamination of 5/6/80 All power reactor Nonradioactive System and facilities with an OL or CP Resulting Potential for Unmonitored, Uncontrolled Release to Environment 80-09 Hydramotor Actuator 4/17/80 All power reactor operating facilities and Deficiencies holders of power reactor construction permits l
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msmenervadarvomeurs Westinchouse Water Reactor sectric Corporation DMsions u,
maaremoverwensiszio May 8, 1980 NS TMA-2245 Mr. V. Stello, Director To-Ali'0*8 Office of Inspection and Enforcement U. S. Nuclear Regulatory Comission 1717 H Street Washington, D. C.
2555
Subject:
Centrifugal Charging Pump Operation Following Secondary Side High Energy Line Rupttere
Dear Mr. Stello:
This letter is to confirm the telephone conversation of May 8,1980 between Westinghouse and Mr. Ed Blackwood of Division of Reactor Operations Inspection.
Office of Inspection and Enforcement, regarding notification made pursuant to Title 10 CFR Part: 21.
A review of the Westinghouse Safety Injection (SI) Temination Criteria following a secondary side high energy line rupture (feedline or steamline rupture at high initial power levels) has revealed a potential for conse-quential damage of one or more centrifugal charging pumps (CCPs) before the SI temination criteria are satisfied and CCP operation taminated.
Such consequential damage may adversely finpact long-tenn recovery operations
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This for the initiating event and is not pemitted by design criteria.
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concern exists for plants which utilize the CCPs as Emergency Core Cooling System (ECCS) pumps, where the CCPs are automatically started, and where the CCP miniflow isolation valves are automatically < isolated upon safety injection Attachment A identifies plants potentially subject to this initiation.A sumary of the concern and recomendations follow.
concern.
Following a secondary side high energy lir.e rupture and associated reactor trip,, Reactor Coolant System (RCS) pressure and. temperature initially decrease.
Safety injection is actuated and the CCPs start to increase RCS inventory.
Reactor Coolant System pressure and temperature. subsequently increase due l
to the loss of secondary inventory, steaml.ine.and feedline isolation, RCS The accident in~ventory addition and reactor core decay. heat generation.
scenario may vary with rupture size and specific plant design, but it will deirelop into a RCS heatup transient with; accompanying increase in RCS pressure.
As RCS pressure increases, the pressuri2er power-operated relief valves Although these (PORVs) are designed to limit RCS pressure to 2350 psia.
valves are normally available, they are not designed as safety-related equip-It can be postulated that, due to either. loss of offsite power.
ment.
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-Mr. V. Stello.
May 8, 1930 NS-TMA-2245 adverse environment inside containment, the pressurizer PORV in manual mode, or the PORY block valve in a closed position, due to PORV lekkage, the pressurizer PORVs may not be operable. As a result of ths RCS heatup and inventory increase, the RCS pressure could rise to the pressurizer safety valve setpoint of 2500 psia within approximately 200 seconds-and remain at that pressure until transient " turnaround." Transient " turn-around" can occur betwe'en 1800 and 4200 seconds depending on operator action and available equipment.
During the initial portion of this transient, the SI termination criteria may not be satisfied.
Consequently, the RCS pressure can reach the pressurizer safety valve relief pressure before CCP operation is terminated. During this period, the. mini'num flow required for CCP opera-tion must be satisfied by flow to the RCS since the CCP miniflow isolation valves are automatically closed on safety injection initiation.~ This requires that the CCPs be able to deliver their minimum required flow to the RCS at the safety valve setpoint pressure.
To evaluate this concern, Westinghouse has developed a calculational method and has reviesed typical CCP head versus flow performance curves and other representative plant parameters. The calculational method considers the effects of safety valve relief setpoint' accuracy, RCS piping resistance, ECCS piping resistance, number of CCPs operating, technical specification allowable CCP head degradation, and uncertainties associated with in-plant verification The analyses for two CCP operation, the best estimate condition, is testing.
similar to the analysis for one CCP operation except that the flowrate used to determine ECCS piping line loss must ensure the minimum flow through each For example, at a specific required head, the pump with the higher pump.
developed head may be required to deliver, greater than the minimum flow in order to permit the lower head pump to meet the minimum flow requirement.
This generic evaluation indicates that sufficient flow to satisfy CCP minimum flow requirements to avoid pump degradation. may not be ensured for a secondary system high energy line rupture under the conditions described above.
Based on the generic evaluation, Westinghouse reconnends that operating plants perform a plant specific evaluation to assess this concern.
provides the Westinghouse calculational method and a sample calculation which can be used in this evaluation. Based on Westinghouse generic review, satis-Should aiplant specific concern be factpry results mey not be obtained.
identified, the following recommendations have been developed and can be tailored to specific plant applications for the : interim until necessary design The interim modifications consist of system modjfications can be implemented.
alignment and operating procedure changes to provide backup to the pressurizer In conjunc-PORys in ensuring that CCP minimum flow requirements are sati the pressurizer PORY operations to maximize the availability of these valves to limit challenges to the pressurizer safety valves, and (b) review the maintenance operations and technical specifications for the backup (i.e., third) charging pump to maximize its availability for long-term recovery from a secondary side rupture.
4*
. V. Stello May 8. 1980 MS-TMA-2245
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modifications described below, are considered sufficient to address this con-cerft in the interim until necessary design modifications can be implemented.
Interim Modification I This interim modification is preferred and requires that component cool'ing water be supplied to the seal water heat exchanger following safety injection initiation in order to provide cooling for-CCP miniflow.
Verify that CCP miniflow return is aligned directly to the CCP suction 1.
during nonnal operation with the alternate return path to the volume control tank isolated (lock closed).
Remove the safety injection initiation; automatic closure signal from 2.
the CCP miniflow isolation valves.
Modify plant emergency cperating procedures.to instruct the operator to:
3.
Close the CCP miniflow isolation valves when the actual RCS a.
pressure drops to the calculated pressure for manual reactor coolant pump trip.
Reopen the CCP miniflow isolation valves should the wide range b.
RCS pressure subsequently rise to greater than 2000 psig.
Interim Modification II This mdification is an alternative for plants in which component cooling water is not supplied to the seal water: heat exchanger following safety Since miniflow cooling is not provided, this alterna-1r.jection initiation.
'ive directs miniflow to the volume control tank to permit the CCP minimum The volume flow requirements to be satisfied with cool uncirculated water.
control tank acts as a surge tank to collect miniflow following safety injection initiation with excess flow directed to a holdup tank via tho i
volume control tank relief valve.
- 1. ' Align the CCP miniflow to the volume control tank during normal opera-tion with the miniflow return path direct to the CCP suction isolated Verify that the volume.contnol tank relief valve and
- (lock closed).
' discharge line capacity exceeds the miniflow requirements of all CCPs
-plus the reactor. coolant pump seal-return flow.
2.
Same as Interim Modification I, Item 2.
3.
Same as Interim Modification I, Item 3.
Mr. V. Stello May 8, 1980 NS-TMA-2245 9
Based on the generic evaluation, Westinghouse has initiated efforts to perfonn additional plant specific analyses for non-operating plants and to develop design modifications to resolve any identified concerns. The modifications will be designed to safety-related standards and will be compatible with Westinghouse SI tennination criteria and standardized technical specifications.
If you require further information, please, call Ray Sero (412-373-4189) of g staff.
Very truly yours.
I e
T. M. Anderson, Manager Nuclear Safety Department TMA/ jaw Attachments
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ATTACitiENT A J
l OPERATING PLANTS-3-Loop 4-Loop Beaver Valley 1
.A ook 1 & 2
, Salem 1 & 2
, s Farley 1 5urry 1 -5 2 Trvjan North Anna.1 & 2 Zion 1 & 2 Sequoyah 1 NON-OPERATING PLANTS 5e' aver Valley 2 Braidwood 1 & 2 Farley 2 Syron 1 & 2 Shearon Harris 1, 2, 3 & 4 Calloway 1 & 2 Catawba 1 & 2 Virgil Sumer Comanche Peak 1 & 2 Diablo Canyon 1 & 2 Jamesport 1 & 2 Haven Marble Hill 1 & 2 McGuire 1 & 2 Millstone 3, Seabrook 1 & 2 Sequoyah 2 Starling Vogtle 1 & 2 Watts Bar 1 & 2 Tyrone Wolf Creek l
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ATTACMENT B
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i MIhMJM CENTRIFUGAL CHARGING. PUMP FLOW DURING TWO PUMP PARALLEL SAFETY _ INJECTION OPERATION In order to ensure that minimum pump flow is maintained during parallel safety infection operation of two centrifugal charging pumps (CCPs),
Westinghouse provides below a sample calculation utilizing actual plant data and determines what actual CCP developed head at the mirliflow flowrate must be available.
Step 1:
Individually determine the developed head of each CCP at the mini-flow flowrate of 60 gpm from field test data.
(two pumps for 4-loop plants and three pumps' for 3-1 bop plants)
Sample: Maximum developed head puch 2571.4 psid v 5940 ft. 9.60 gpm Minimum' developed head pump 2554.1 psidfiS900 ft. 9.60 gpm Step 2: Correct the pump head foi-testih'g error.. Add the appropriate error in detemining the above measured-developed head, i.e.,
instrument error plus reading, error,c to the maximum developed head and subtract this' error.from the minibum developed head.
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Sample:
Pressure instrument accuracyiof i 0.5 percent x span of measuring instrument of 3000 psig - 15 psi (35 ft. of head), plus 10' psi (23 ft.) read'ing accuracy = 58 ft.
The resultant CCP developed; heads at miniflow which
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can be supported areia maximum developed head of 5998 ft. for the maximum head pump, and a minimum developed head of 5842 ft. for the minimum head pump.
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ATTACWlENT B
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Step 3:
Determine total CCP flow.
Construct a. pump curve fo,r the maxi-num head pump that is parallel' to the actual *as-built" vendor pump curve and passas through the above detemined develjoped head at the miniflow flowrate which is the measured developed head plus the detemined measurement accuracy.
(Seeattach-ment Figure 1.)
Use this head versus flow curve to detemine the flow delivered by the maximum head pump (strong pptp) at the developed head of the minimum head pump (weak: pump) at the miniflow flowrate (i.e., 5842 ft. as detemined in Step.1).
.1 Sample: As illustrated in Figure 1, the delivered flow of the strong pump at 5842 ft. is 150 gpm. Therefore, the total flow from both CCPs which guarantees that the weak CCP will be delivering at least 60 gpm is 210 gpm (150 gpm + 60 gpia).
Step 4: DeterT-ine Infection Piping Head Loss. The head loss due to friction in the safety injection /RCP seal infection piping is detemined as follows:
- Yhe Ah is equal to the strong:CCP developed head at runout f
flow. This resistance is established during the CCP flow balance testing which limits CCP flow to the runout limit.
The injection piping resistence -(k) is equal to the' developed head of the stron CCP'at its. runout. flow divided,by the -
(runoutflowrate) g, develooed head (runout flowrate)2 = T,1500 ft. 2 a h.,,
(S50 gpm)
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2 k = 4.96 x 10-Ift./gpm
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ATTACWiENT B i.
Theresistanceoftheinjection. piping (Ah),atthetotalCCPflow f
required to maintain 60 gpm through the weak CCP is:
f = (4.96 x 10-3 ft 2) (210 gpm)2 = 219,ft.
Ahf = kQ or Ah Step 5: Determine head loss through the Reactor Coolant System.
Consider thr.t the reactor coolant pumps are operating, therefore, the pressure drop from the CCP cold leg injection nozz1cs through the reactor vessel to the pressurizer surge Ifne off the $ot leg at full RCS flow are to be included! This pressure drop.is approximately 50 psid (116 ft..) for 4-loop plants and 48 psid (111 ft.) for 3-loop plants. This' pressure drop must be overcome by the CCPs in order to deliver' flow to the'RCS at the' hot leg /
pressurizer pressure.
Determine the elevational head between the RWST and the pressuriz'er Step 6:
safety valves, 160 ft.
e.g.
RWST elevation 100 ft.
CCP suction elevation RCS cold leg injection nozzle elevation - 126 ft.
187 ft.
Pressurizer safety valve. elevation 60 ft.
RWST to CCP suction minus CCP suction to RCS'
- (-26 ft.)
minus RCS to pressuffrer safety. valves (61 ft. assuming a full pressurizer)
- (-44 ft.)
corrected for density difference
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-10 ft.
Thus, in this example the CCPs.must provide an additional 10 ft.
of elevational head.
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. ATTACHMENT 8 Step 7: Calculate the pressurizer safety valve relief pressure.
relief pressure = safety valve nominal relief pressure
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+ 15 setting tolerance l
relief pressure = 24S5:psig + 25 psig = 2510 psig (5798 ft.)
Step 8: Determine the maxinum RCS pressurizer pressure at which 60 gpe.
minimum flow is maintained through the weak CCP.
Maxiem RCS pressure = (CCP developed head at total CCP flowrate) -
(injection piping head loss) - (head 1 css through RCS) - (eleva-tion head less)
Maximum RCS pressure = 5842 f t. - 21g f t. - 116 f t. - 10 ft. =
5497 ft. = 2380 psig Comparing this pressure to the pressurizer safety valve relief pressure (Step 7) of 2510 psig, it 15 evident tht the 60 c:e flov required for the weak CCP will not be maintained.
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