ML19344A506
| ML19344A506 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/24/1980 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8008200522 | |
| Download: ML19344A506 (1) | |
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'o UNITED STATES s
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NUCLEAR REGULATORY COMMISSION E
REGION V
,g 1990 N. CALIFORNIA BOULEVARD o,%
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o SulTE 202. WALNUT CREEK PLAZA 44,,5 WALNUT CREEK, CALIFORNIA 94596 July 24, 1980 Docket No. 50-312 Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Attention: Mr. John J. Mattimoe Assistant General Manager Gentlemen:
The enclosed Ic Bulletin No. 80-18, is forwarded for action. A written response is required.
In order to assist the NRC in evaluating the value/ impact of each Bulletin on licensees, it would be helpful if you would provide an estimate of the manpower expended in conduct of the review and preparation of the report (s) required by the Bulletin.
Please estimate separately the manpower associated with correc-tive actions necessary following identification of problems through the Bulletin.
If you desire additional information regarding this matter, please contact this office.
Sincerely, R. H. Engelken Director
Enclosures:
1.
List of Bulletins Recently Issued cc w/ enclosures:
R. J. Rodriguez, SMUD L. G. Schwieger, SMUD 8008200 7 8 1
SSINS No.: 6820 Accession No.:
UNITED STATES
.J005050062 NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555 July 24, 1980 IE Bulletin No. 80-18 MAINTENANCE OF ADE0VATE MINIMUM FLOW 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 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 durir.g 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 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 J
letter (enclosure) be performed to determine if adequate minimum flow is assured under all conditions. If adequate minimum flow is not assured, Westinghouse recomends 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:
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 conditions by the calculations in 1:
a.
Make modifications to equipment and/or procedures, such as those 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 adequatePinimum flow for any transient or accident requiring SI can and will be accomplished in the time necessary.
c.
Verify that any manipulations required (valve opening or closing, along with the instrumentation necessary to indicate need for the action or accomplishment of the action, etc.) can be accomplished without offsite power available.
d.
Justify that flow available from the CCPs with the modifications in place will be sufficient to justify continued applicability of any safety related analyses which take credit for flow from the CCPs (LOCA,HELB,etc.).
e.
Justify that all Technical Specifications based on the Item 2.d analyses remain valid.
3.
Provide the results of calculations performed under Item 1, and describe any nodifications 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:
1.
In a quantitative manner similar to 1 above, determine whether or not 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 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.
IE Bulletin No. 30-18 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 aporopirate 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, U to V. Stello, IE dtd t/5'/80 i
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- Those considered to be "nearino licensing" are: North Anna 2, Diablo Canyon 1, McGuire, Salen 2, and Sequoyah.
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IE Bulletin No. 80-18 Enclosure
' 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 to 80-17 Testing Subsequent to facilities holding OLs 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 Discharge Volume Capability OL 80-13 Cracking In Core Spray 5/12/80 All BWR's with an Spargers OL 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 Monradioactive System and facilities with an Resulting Potential for OL or CP Unmonitored, Uncontrolled Release to Environment 80-09 Hydramotor Actuator 4/17/80 All power reactor Deficiencies operating facilities and holders of power reactor construction permits
V Westinghouse Water Reactor w kmanoMem ElectricCorporation DMslons em... r
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May 8, 1980 NS-TMA-2245 Mr. V. Stello, Director Office of Inspection and Enforcement FO 4 d'O*8 U. 5. Nuclear Regulatory Commission 1717 H Street Washington, D. C.
20555
Subject:
Centrifugal Charging Pump Operation Following Secondary Side liigh Energy Line Rupture
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) Tennination 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 termination critoria are satisfied and CCP operation tonninated.
Such consequential damage may adversely impact long-term recovery operations for the initiating event and is not permitted by design criteria. This 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 initiation. Attachment A identifies plants potentially subject to this concern. A summary of the concern and recommendations follow.
Following a set.ondary side high energy line rupture and associated reactor trip, Reactor Coolant System (RCS) pressure and temperature initially decrease.
Safe,ty injection is actuated and the CCPs start to increase RCS inventory.
Reactor Coolant System pressure and temperature subsequently increase due to the loss of secondary inventory, steamline and feedline isolation, RCS inventory addition and reactor core decay heat generation. The accident scenario may vary with rupture size and specific plant design, but it will develop into a RCS heatup transient with accompanying increase in RCS pressure.
As RCS pressure increases, the pressurizer power-operated relief valves (PORVs) are designed to limit RCS pressure to 2350 psia. Although these valves are normally available, they are not designed as safety-related equip-
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It can be postulated that, due to either loss of offsite power,
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Mr. V. Stello Mty 8, 1980 NS-TMA-2245 adverse environment inside containment, the pressurizer PORV in manual mode, or the PORV block valve in a closed position, due to PORV leakage, the pressurizer PORVs may not be operable. As a result of the 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 entil transient " turnaround." Transient " turn-around" can occur betwe'en 1800 and 4200 seconds depending on operator action and tvailable 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 injection initiation.- This requires that the CCPs be able to deliver their minimum required flow to the RCS at s
the safety valve setpoint pressure.
To evaluate this concern, Westinghouse has developed a calculational method and has reviewed 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 testing. The ana'yses for two CCP operation, the best estimate condition, is similar to the analysic for one CCP operation except that the flowrate used to detemine ECCS piping line loss must ensure the minimum flow through each pump. 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 order to pemit 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 recomends that operating plants perform a plant specific evaluation to assess this concern. Attachment 8 provides the Westinghouse calculational method and a sample calculation which can be used in this evaluation.
Based on Westinghouse generic review, satis-factpry results may not be obtained.
Should a plant specific concern be identified, the following recommendations have been developed and can be tailored to specific plant applications for the interim until necessary design modificaticns can be implemented.
The interim modifications consist of system alignment and operating procedure changes to provide backup to the pressurizer PORVs in eosuring that CCP minimum flow requirements are satisfied.
In conjunc-tion with the interin modifications, it is recomended that plants, (a) review the pressurizer PORV 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 These recomendations, in combination with the interim secondary side rupture.
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May 8,1980 r. V. Stello NS-TMA-2245 d sufficient to address this con-can be implemented.
nodifications described below, are considere i cern in the interim until necessary design modificat ons Interim Modification _I_
ponent cooling
_This interim modification is preferred and requires that comfollowing safet water be supplied to the seal water heat exchangerinitiation the CCP suction N
Verify that CCP miniflow return is aligned directly tod lume t
1.
control tank isole.ted (lock closed).
i al from Remove the safety injection initiation automatic closure s gn the CCP miniflow isolation valves.
2.
h perator to:
Modify plant emergency operating procedures to instruct t e o S
Close the CCP miniflow isolation ~ valves when the 3.
l reactor pressure drops to the calculated pressure for manua a.
coolant pump trip.
ide range Reopen the CCP miniflow isolation valves should b.
Interim Modification II_
n nt cooling This modification is an alternative for plants in which compc e r following safety h
water is not supplied to the seal water heat exc angeSince miniflow cool mit the CCP minimum tive directs miniflow to the volume control tank to perflo injection initiation.
t The volume j
following safety control tank acts as a surge tank to collect miniflowin tank via the volume control tank relief valvo.
' Align the CCP miniflow to the volume control tank during n tion isolated tion with the miniflow return path direct to the CCP sucVerif 1.
t f all CCPs discharge line capacity exceeds the miniflow requiremen s o (lock closed).
plus the reactor coolant pump seal return flow.
Same as Interim Modification I, Item 2.
2.
Same as Interim Modification I, Item 3.
3.
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t May 8, 1980. V. Stello NS-TMA-2245 ased on the generic evaluation, Westinghouse has initiated efforts to perform dditional plant specific analyses for non-operating plants and to develop The modifications
- csign modifications to resolve any identified concerns.and will be compatible with d d 1111 be designed to saftty-related stan ar s lestinghouse SI termination criteria and standardized technical specifications.
If you require further information, please call Ray Sero (412-373-4189) of my Itaff.
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Very truly yours,
/[dh t, M. Anderson, Manager Nuclear Safety Department IMA/ jaw Attachments g~.
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ATTACHMENT A O_PERATING PLANTS 3-Loop 4-Loop Beaver Valley 1
-Cook 1 & 2
. > Farley 1
-Salem 1 & 2 5urry 1 & 2 Trojan North Anna 1 & 2 Zion 1 & 2 Sequoyah 1 NON-0PERATING PLANTS Beaver Valley 2 Braidwood 1 & 2 Farley 2 Byron 1 & 2 Shearon Harris 1, 2, 3 & 4 Calloway 1 & 2 Virgil Summer Catawba 1 & 2 Comanche Peak 1 & 2 Diablo Canyon 14 2 Jamesport 1 & 2 Haven Marble Hill 1 & 2 McGuire 1 & 2 Millstone 3 Seabrook 1 & 2 Sequoyah 2 Sterling Vogtle 1 & 2 Watts Bar 1 & 2 Tyrone Wolf Creek e
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ATTACitlENT B MINIMUM 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 detemines what actual CCP developed head at the mirliflow flowrate must be available.
Step 1:
Individu>.lly 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-loop plants)
Sample: Maximum developed head pump 2571.4 psid = 5940 ft. 9 60 gpm Minimum' developed head pump 2554.1 psid = 5900 ft. 9 60 gpm Step 2: Correct the pump head for testing error. Add the appropriate error in detemining the above measured. developed head, i.e.,
instrument error plus reading error, to the maximum developed head and subtract this, error.from the miniinum developed head.
Sample:
Pressure instrument accuracy of t,0.5 percent x span of measuring instrument of 3000 psig - 15 psi (35'ft. of head), plus 10 psi (23 ft.) reading
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accuracy = 58 ft.
The resultant CCP developed heads at miniflow which can be supported are a maximum developed head of 5998 ft. for the :aaximum head pump, and a minimum developed head of 5842 ft. for the minimum head pump.
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Determine total CCP flow. Construct a pump curve for the maxi-mum head pump that is parallel to the actual "as-built" vendor pump curve and passes through the above determined developed head at the miniflow flowrate which is the measured developed head plus the determined measurer.ent accuracy.
(Seeattach-ment Figure 1.)
- 3 Use this head versus flow curve to determine the flow delivered by the naximum head pump (strong pump) at the developed head of, the minimum head pump (weak pump) at the miniflow flowrate (i.e., 5842 ft. as detenmined in Step 1).
fample: 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 gpm).
Step 4:
Determine Injection Piping Head Loss.
The head loss due to friction in the safety injection /RCP seal injection piping is determinc ' as follows:.
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-The ah is equal to the strong CCP developed head at runout f
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This resistance is established during the CCP flow balance testing which limits CCP flow to the runout 1.imit.
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The injection piping resistance (k) is equal to the developed head of the strong CCP at its runout flow divided by the -
(runout flowrate)2 develoopd haad Ah 19nn ft.
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" (runout flowrate)
Q ~ (550 gpm) k = 4.96 x 10-3 ft./gpm2 I
- ATTACHMENT E The resistance of the injection piping (ah ), at the total CCP flow f
required to maintain 60 gpm through the weak CCP is:
2
= (4.96 x 10-3 ft 2) (210 gpm)2 = 219 ft.
Ahf = kQ or ahf gp 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 leg injection nozzles through the reactor vessel to the pressurizer surge line off the hot 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 CCPL 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 pressurizer safety valves.
160 ft.
e.g.
RWST elevation CCP suction elevation 100 ft.
RCS cold leg injection nozzle elevation - 126 ft.
187 ft.
Pressurizer safety valve elevation 60 ft.
(-26 ft.)
minus RCS to pressurizer safety valves l
(61 ft. assuming a full pressurizer)
(-44 ft.)
corrected for density difference 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 RCS pressure = (CCp developed head at total CCp flowrate) -
(injection piping head loss) - (head loss through RCS) - (eleva-tion head loss) i Maximum RCS pressure - 5842 ft. - 219 ft. - 116 ft. - 10 ft. =
5497 ft. = 2380 psig Comparing this pressure to the pressurizer safety valve relief pressure (Step 7) of 2510 psig, it is evident that the 60 gpm flow required for the weak CCP will not be maintained.
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