ML19344A505
| ML19344A505 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 07/24/1980 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Reed C COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8008200521 | |
| Download: ML19344A505 (1) | |
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e UNITED STATES 8 ' 3.,., }
NUCLEAR REGULATORY COMMISSION
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.e REGION lli 799 ROOSEVELT ROAD e
GLEN ELLYN. ILLINOIS 60137 o
Jul. 2 4 S80 Docket Nos. 50-454, 50-455; 50-456, 50-457 Commonwealth Edison Company ATTN:
Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:
The enclosed IE Bulletin 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, p.s. A/
[ Kmes G. Keppler Director
Enclosure:
cc w/ encl:
)
Mr. D. L. Peoples, Director Director, NRR/ DOR l
of Nuclear Licensing PDR Mr. Gunner Sorensen, Site Local PDR Project Superintendent NSIC Mr. R. Cosaro, Project TIC Superintendent Mr. Dean Hansell, Office of Central Files Assistant Attorney General l
Director, NRR/DPM Myron M. Cherry, Chicago l
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80082006g{
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SSINS No.: 6820
'd Accession No.:
UNITED STATES 8005050062 NUCLEAR REGULATORY COM11SSION
'3 T'
c, OFFICE OF INSPECTION AND ENFORCEMENT
- 2 J WASHINGTON, D.C. 20555 r
July 24, 1980 IE Bulletin No. 80-18 MAINTENANCE OF ADEQUATE MINIMUM FLOW THRU CENTRIFUGAL CHARG 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 plat *s under construction (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 certain 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 may not 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 minimum 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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IE Bulletin No. 80-18 July 24, 1980 Page 2 of 3 Actions 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:
s
-1.
Perform the calculations, outlined in the enclosure, for your plant.
2.
If availability of minimum cooling flow for the CCPs is not assured for all conv. Lions by the calculations in 1:
Make modifications to equipment and/or procedures, such as those e.
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.
b.
Justify that any manual actions necessary to assure adequate minimum 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 bstrumentation necessary to indicate need for the action or accompilshment of the action, etc.) can be accomplished without offsite 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 to 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 the enclosure considered to be " nearing licensing" are:
North Anna 2, Diablo Canyon 1, McGuire 1, Salem 2, and Sequoyah. These plants Other non-licensed plants must respond in writing within the specified time.
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 temination criteria.
If a " minimum flow bypass" line is present which remains open during high pressure injection, and if that line g'uarantees 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:
Ltr from T. M. Anderson, W to V. Stello, IE dtd 5/5/80 l
l "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 No.
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 facilities holding OLs to 80-17 to Insert During a Scram at a BWR 80-17 Failure of Control Rods 7/3/80 All BWR power reactor facilities holding OLs to Insert During a Scram at a BWR 80-16 Potential Misapplication of 6/27/80 All Power Reactor Rosemount Inc., Models 1151 Facilities with an OL or a CP and 1152 Pressure Transmitters 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 holdfrg 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 I
Spargers 80-12 Decay Heat Removal System 5/9/80 Each PWR with an OL Openbility 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 facilities with an Nonradioactive System and OL or CP Resulting Potential for Unmonitored, Uncontrolled Release to Environment 80-09 Hydramotor Actuctor 4/17/80 All power resctor operating facilities and Deficiencies holders of power reactor construction pemits
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- o esser Terreurv ower, Wecinch00se Water Reactor Be:!fic Corporat!On DMslons a
Prsuran revernwee nzic i
May 8, 1980 MS-TMA-2245 Mr. V. 5tallo, Director E0-A M*0*8 Office of Inspection and Enforcement U. S. Nuclear Regulatory Comission 1717 H Street Washington, D. C.
20555 Centrifugal Charging Pump Operation Following Secondary Side
Subject:
High Energy Line Rupture
Dear Mr. Stello:
This letter is to confinn 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 l
Title 10 CFR Part; 21.
A review of the Westinghouse Safety Injection (SI) Ter uination Criteria following a. secondary side high energy line rupture (feedline or steaiuline rupture at high initial power icvels) has revealed a potential for conse-quential damage of one or more centrifugal charging pumps (CCPs)inated.
before the SI temination criteria are satisfied and CCP operation term Such consequential damage may adversely impact long-term recovery operations This for the initiating event and is not permitted by design criteria.
concern exists for plants which utilize the CCPs as Emergency Core Cooling System (ECCS) pumps, where the CCPs are automatically started, and where th CCP miniflow isolation valves are autee.atically isolated upon safety injection Attachment A identifies plants potentially subject to this initiation.A sumary of the concern and recomendations follow.
concern.
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Following a secondary side high energy lire rupture and associated reactor Safety injection is actuated and the CCPS (tart to increa Reactor Coolant System pressure and temperature subsequently increase due RCS
, to the loss of secondary inventory, steamline and feedline isolation, The accident inventory addition and reactor core decay heat generation. sc develop into a RCS heatup transient with' accompanying increase in RC5 pressure As RCS pressure increases, the pressuri2er power-operated relief valves l
Although these (PORVs) are designed to limit RCS pressure to 2350 psia.
l valves are normally available, they are not designed as safety-related equip-i It can be postulated that, due to either. loss of offsite power,
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Mr. V. Stello 2-.
May 8, 1980 NS-TMA-2245 adverse environment inside containment, the pressurizer PORY in manual mode, or the p0RV block valve in a closed position, due to PORV leakage, the pressurizer PORVs may not be operable. As a result of ths RCS haatup 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. minimum 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 infection initiation.' Thfs 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 perfonnance curves and other Npresentative plant parameters. The calculational authod considers the eHects 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 developed head may be required to deliver. greater than the minimum flow in pump.
order to pemit the lower head pump to meet the minimum flow requirement.
This generic evaluation indicates that'suffici6nt 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, nasca vi, the generic evaluation, Westinghouse reconnends that operating plants perform a plant specific evaluation to assess this concern.
provides the Westinghouse calculational'methodiand a sample calculation which can be used in this evaluation. Based on Westinghouse generic review, satts-Should 'aiplant specific concern be factpry results may not be obtained. identified, the following reconsnendation tailored to specific plant applications for the : interim until necessary design The interim modifications consist of system modifications can be implemented.
alignment a-d operating procedure changes to provide backup to the pressurizer In conjunc-PORVs in ensuring that CCP minimum flow requirements are satisfied.
tion with the interim modifications, it is recocynended that plants. (a) review the pressurizer PORV operations to maximize the availability of these valves 7
to limit challenges to the pressurizer. safety valves, and (b) review the l
maintenance operations and technical specifications for the backup (i.e., third) charging pump to maximize its availability for long-tenn recovery from secondary side rupture.
_-.. _ -, ~. _ c-
. May 8,1980 Mr. V. Stello MS-TMA-2245 modifications described below, are considered sufficient to address this con.
cern in the interim until necessary design modifications can be implemented.
In_terim 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 nomal 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 operating procedures to instruct the operator to:
3.
Close the CCP miniflow isolation valves when the actual RCS pressure drops to the calculated pressure for manual reactor a.
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 modification 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-injection initiation.
tive directs miniflow to the volume control tank to pemit 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 the volume control tank relief valve.
l
- 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 (lock closed). Verify that the volume. control tant relief valve and f all CCPs
. discharge line capacity exceeds the miniflow requirements o plus the reactor coolant pump seal return flow.
2.
Same as Interim Modification I, Item 2.
Same as Interim Modification I, Item 3.
3.
. May 8, 190
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Mr. V. Stello hS-TMA-2245 Sesed on the generic evaluation, Westinghouse has initiated efforts to perfom additional plant specific analyses for ncn-eperating plants and to develc; The sodificatiens design redifications te resolve any identified concerns.
will be designed to safety-related standards and will be competitle with Westinghouse SI temination criteria and standardized technical spe:ificatiens.
If you re:ptre further infomation, please. call Ray Sero (412-373-4189) of g statf.
Very truly yours, hm i
e T. M. Mderson, Manager L: lear Safecy Departrent TRA/ jaw Attachments o
ATTACWlhT A OPERATIPi3 PLAMS 3-toop 4 toop Beaver Valley 1
,, Cock 1 & 2
, s Farley 1
, Salem 1 & 2
.surry 1 & 2 Trojan North Anna 1 & 2 2 ion 1 & 2 Sequoyah 1 NON-0*ERATIN3 PLANTS Beaver Valley 2 Braidwood 1 & 2 Farley 2 Byron 1 & 2 Shearon Harris 1, 2, 3 & 4 Calloway 1 & 2 Cata#a 1 & 2 Virstl Sterwr Corsanche Peak 1 & 2 Diablo Canyon 1 & 2 Jamesport 1 & 2 Haven Ma61e Hill 1 & 2 McGuire 1 & 2 Millstone 3 Seabrook 1 & 2 Sequoyah 2 Steriing Vogtle 1 & 2 Watts Bar 1 & 2 Tyrone Wolf Creek e
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ATTAcit(ENT B MINIM'JM CENTRIFUGAL CHARGING. PUMP FLOW DURING TWO PUMP PARALLEL SAFETY _ INJECTION OPERATION In order to ensure that minimum pump flow is maintained during parallel safety injection 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 pump 2571.4 psid M 940 ft. 0.60 gpm Minimum' developed head pump 2554.1 psid.hiS900 ft. 9.60 gpm e
Step 2: Correct the pump head for testihg error.. Add the appropriate error in detennining the above-measured developed head, i.e.,
I
' instrument error plus reading, error.cto the maximum developed head and subtract this' error.from the miniinum developed head.
Pressure instrument accuracy 6of + b.5 percent x Sample:
span of measuring instrument.cf 3000 psig = 15 psi (35ft.ofhead),plus10' psi (23ft.) read'ing
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accuracy = 58'ft.
The resultant CCP developed; heads at miniflow which I
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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 SM2 ft. for the minimum heaa pump.
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2 ATTACif1ENT B
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- r, Step 3:
Detemine total CCP flow.
Construct a. pump curve for the max 1-num head pump that is para 11el' to the actual "as-built" vendor pump curve and passes through the above detemined developed 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 pymp) at the devoloped head of the minimum head pump (weak: pump) at the miniflow flowrate (i.e. 5842 ft. as detemined in Step:1).
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Semple: As illustrated in Figure 1, the delivered ficw of the strong pump at 5842 ft. is 150 gpm. Therefore, the total flow from both CCPs which guarkntees that the weak CCP will be delivering at least 60 gpm is 210 gpm (150 gpm + 60 gpia).
Step 4: Deterrine Injection Piping Head Loss. The head loss due to friction in the safety. injectior}/RCP seal infection piping is detemined as follows:.
The ah is equal to the strong:CCP developed head at runout
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f flow. This resistance,.is established during the CCP flow balance testing which limits CCP flow to the tv-Ogi limit.
The injection piping resistance -(k) is equal. 'a J.e'cleveloped head of the stron CCP'at itsi runout. flen kW by the -
(runout flowrate)
(runoutflowrate)2=,g.,1500ft.2 g, develooed: head Q
(550 gpm)
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2 k = 4.96 x 10-h./ggn t
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ATTACW4ENT B The resistance of the injection piping (Ah ) at the total CCP flow f
required to maintain 60 gpm through the weak CCP is:
g ) (210 gpm)2 = 219 ft.
f = (4.96 x 10-3, t22 Ahf = k0 or ah Step 5: Determine head loss through the Reactor Coolant System.
Consider that the reactor coolant pumps are operating, therefore, the pressure drop from the CCP cold 1.eg injection nozzics through the reactor vessel to the pressurizer surge line off the Sot leg
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at full RCS flow are to be included! This pressure drop.it 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.
Step _6:_
Determine the elevational head between the RWST and the pressuriz'er safety valves.
160 ft.
RWST elevation 100 ft.
e.g.
CCP suction elevation RCS cold les injection nozzle elevation - 126 ft.
187 ft.
Pressurizer safety valve elevation 60 ft.
1 RWST to CCP suction minus CCP suction.to RCS
- (-26 ft.)
minus RCS to pressurizer safety. valves (61 ft. assuming a full pressurizer) corrected for density difference
- (-44 ft.)
-10 ft.
Thus, in this example the CCPs.must provide an additional 10 ft.
of elevational head.
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,. ATTACHMENT B Step 7: Calculate the pressurizer safety valve relief pressure.
e.g.
relief pressure = safety valve. nominal relief pressure
+ 1% setting tolerance relief pressure = 2485:psig + 25 psig = 2510 psig (5798 ft.)
Step 8: Determine the maximum RCS pressurizer pressure at which 60 gpm minimum flow is maintained through the' weak CCP.
Maximum RC5 pressure = (CCP developed head at total CCP flowrate) -
(injection piping head loss) - (head less through RCS) - (eleva-tion head less)
Maximum RC5 pressure = 5842 ft. - 219 ft. - 116 ft. - 10 ft. =
5497 ft. = 2380 psig Comparing this prer,sure to the pressurizer safety valve relief pressure (Step 7) of 2510 psig,-it it evident that the 60 gpm flow required for the weak CCP will not be maintained.
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