ML19344E275

Forwards IE Bulletin 80-18, Maint of Adequate Min Flow Through Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture. No Written Response Required

ML19344E275
Person / Time
Site:	Wolf Creek
Issue date:	07/24/1980
From:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Koester G KANSAS GAS & ELECTRIC CO.
References
NUDOCS 8008280297
Download: ML19344E275 (1)

Text

THIS DOCUMENT CONTAINS *

1. E0h QUAUTY PAGES UNITED STATES o,

N JCL%R REGULATORY COMMISSION REGloN IV I

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LPDR July 24, 1980 TIC s NSIC In Reply Refer To:

CENTRAL FILES RIV Docket No.

STN 50-482/IE Bulletin No.i80-18 Kansas Gas & Electric Co.

Atm:

Mr. Glenn L. Koester Vice President-0perations Post Office Box 208 Wichita, Kansas 67201 Gentlen.en :

The enclosed IE Bulletin No. 80-18, is forwarded 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 infor.tation regarding this matter, please contact this office.

Sincerely,

[

,6 c.=-ps Karl V. Seyfrit,//

Director

'/

/

Enclosures:

1.

IE Bulletin No. 80-18 2.

List of Bulletins Recently Issued cc: w/ enclosures Messrs. Nicholas A. Petrick, SNUPPS D. T. McPhee, Kansas City Power and Light Company Gerald Charnoff, Shaw, Pittman, Potts & Trowbridge E. W. Creel, Kansas Gas and Electric Company 8008280 2f [

0

SSINS No.: 6820 Accession No.:

UNITED STATES 8005050062 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 IE Bulletin No. 80-18 Cate: July 24, 1980 Page 1 of 3 MAINTENANCE OF ADEQUATE 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 during SI following a recondary 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 utessurizer 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 letter (enclosure) be performed to determine if adequate minimum flow is l

assured under all conditions. If adequate ninimum 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.

l l

IE Bulletin No. 80-18 Date: 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 adequate minimum 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 j

analyses remain valid.

i 3.

Provide the results of calculations performed under Item 1, and describe 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:

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 l

l

• 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.

I

IE Bulletin No. 80-18 Date: July 24, 1980 Page 3 of 3 termination criteria.

If a " minimum flow bypass" line is present whica 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 specifical1y for identified generic problems.

Enclosure:

Ltr from T. M. Anderson, W to V. Stello, IE (w/ attachments) dtd 5/8/80 i

• Those considered to be " nearing licensing" are: North Anna 2, Diablo Canyon 1, McGuire, Salem 2, and Sequoyah.

l

1 BectricCorporation DMalons a

anumnruiernem nem May 8, 1980 NS-TMA-2245 Mr. V. Stallo, Director Office of Inspection and Enforcement TO ~' J d ' O ' 8 U. S. Nuclear Regulatory Connission 1717 H Street Washington, D. C.

20555 Sabject:

CentriYugal Charging Pump Operation Following Secondary Side High Energy Line Rupture

Dear Mr. Stallo:

This-letter is to confim 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-quantial damage of one or more centrifugal charging pumps (CCPs) before the SI temination criteria are satisfied and CCP operation teminated, 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 l

System (ECCS) pumps, where the CCps are automatically started, and where the CCP :ainiflow isolation valves are automatically isolated upon safety injection i

initiation.

Attachment A identifies plants potentially subject to this concern. A sunnary of the concern and reconnendations follow.

Following a secondary side high energy line 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 to the loss of secendary inventory, steamlir.e 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 ifmit RCS pressure to 2350 psia. Although these valves are normally available, they are not designed as safety-related equip-ment.

It can be postulated that, due to either loss of offsite power,

Mh V. Stello May 8, 1980 NS-TMA-2245 adverse environment inside containment, the pressurizer PORY in manual mode, or the PORY block valve in a closed position, due to PORY leakage, the-pressurizar 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 approximatev 200 seconds-and remain at that pressure until transient " turnaround." Transient " turn-around" can occur between 1800 and 4200 seconds depending on operator action and available equipment.

During the initial portion of this transient, the SI tennination criteria may not be satisfied.

Consequently, the RCS pressure can reach the pressurizer safety valve relief pressure before CCP operation i

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 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 uncerttinties associated with in-plant verification testing. The analyses-for two CMP operation, the best estimate condition, is similar to the analysis for one CCP operation except that the flowrate used I

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 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 recommends that operating plants perform a plant specific evaluation to assess this concern.

Attachment B provides the Westinghouse calculational method and a sample calculation which can be used in thi:. valuation.

Based on Westinghouse generic review, satis-factpry results may not be obtained.

Should a plant specific concern be identified, the following recorrrnendations have been developed and can be tailored to specific plant applications for the interim until necessary design modifications can be implemented.

The interim modifications consist of system alignment and. operating procedure changes to provide backup to the pressurizer PORVs in ensuring that CCP minimum flow requirements are satisfied.

In conjunc-tion with the interim modifications, it is recommended that plants, (a) review the. pressurizer PORY operations to maximize the availability of these valves to limit challenges to the pressurizer safety valves, and.(b) review t.M maintenance operations and technical specifications for the backup (i.e., third) l charging pump to maximize its availability for long-term recovery from a secondary side rupture.

These recomendations, in combination with the interim i

Mr. V. Stallo May 8, 1980

~

NS-TMA-2245 modifications described below, are considered sufficient to address this con-cern in the interim until necessary design modifications can be implemented.

Interim Modification I l

This interim modification is preferred and. requires that component cooling water be supplied to the seal water heat exchanger following safety injection initiation in order to provide cooling for CCP miniflow.

1.

Verify that CCP miniflow return is aligned directly to the CCP suction during normal operation with the alternate return path to the volume control tank isolated (lock closed).

2.

Remove the safety injection initiation automatic closure signal from the CCP miniflow isolation valves.

3.

Modify plant. emergency operating procedures to instruct the operator to:

z.

Close the CCP miniflow isolation valves when the actual RCS pressure drops to. the calculated pressure for manual reactor coolant pump trip.

b.

Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2000 psig.

Interim Modification II This modification is an alternative for plants in which component cooling we.ter is not supplied to the seal water heat exchanger following safety l

injection initiation.

Since miniflow cooling is not provided, this alterna-tive directs miniflow to the volume control tank to permit the CCP minimum flow requirements to be satisfied with cool uncirculated water. The volume 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 reitef valve.

i 1.

' Align the CCP miniflow to the volume control tank during normal cpera-

~

tion with the miniflow return path direct: to the CCP suction isolated (lock closed).

Verify that the volume control tank relief valve and 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.

l

Mr. V. Stello May 8, 1980 NS-TNA-2245 i

Sased on the generic evaluation, Westinghouse has initiated efforts to perform 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 temination criteria and standardized technical specifications.

If you require further information, please call Ray Sero (412-373-4189) of my staff.

Very truly yours, 1

i

~

T. M. Anderson, Manager Nuclear Safety Department TMA/ jaw Attachments

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ATTAC1NENT A OP9tATING PLANTS 3-1.oon 4-1. coo Beaver Valley 1

.A cok 1 & 2

.. j Farley 1

-Salem 1 & 2 ~

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Surry 1 & 2 Trojan North Anna 1 & 2 Zion 1 & 2 Sequoyah 1 MON-OPERATING PLANTS Be' aver Valley 2

'Braidwood 1 & 2 Farley 2 Byron 1 & 2 Shearon Harris 1, 2,. 3 & 4 Calloway 1 & 2 Virgil Stamer Catawba 1 & 2 Comanche Peak 1 & 2 Diablo Canyen 1 & 2 Jamesport 1 & 2 Haven Marble Hill 1 & 2 McGuire 1 & 2 Mi11stene 3 Seabrook 1 & 2 Sequoyah 2 Sterling Vogtle 1 & 2

'datts Sar 1 & 2 Tyrone Wolf Creek et e

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ATTAC}9 TENT S MINIMUM CENTRIFUGAL CHARGING PUMP FLOW DURING' TWO "41;? 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 910w a sample calculation utilizing actual plant data. and detemines whP actual CCP developed head at the miniflow flowrate must be available.

Step 1:- Individually determine the developed head of each CCP at the mini-flow flowcate 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 Stap 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 i

head and subtract this error.from the miniinum developed head.

l Sample:

Pressure instrument accuracy of + 0.5 perc5nt x span of measuring instrument of 3000 psig = 15 psi (35 ft. of head), plus 10 psi (23 ft.) reading accuracy = 58 ft.

The-resultant CCP developed. heads at miniflow hich

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I can be supported are a 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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2 ATTACHMENT B Step 3:

Determine total CCF 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 i

head plus the determined measurement accuracy.

(See attach-ment Figure 1.)

Use this head versus flow curYe to detemine the flow delivered by the maximum head pump (strong pump) at the developed head of.

the minimum head pump (weak pump) at the miniflow flowrate (i.e., 5842 ft. as detennined in Step;1).

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Sample: As illustrated in Figure 1 the delivered flow of the strong pump at 5842 ft. is 150 gpm.

Therefore, the total flow frcm both CCPs which guarantees that the weak CCP will be delivering at least.60 gpm is 210 gpm (150 gpm + 60 gpm)..

Step 4:

Detamine Injection Piping Head Loss.

The head loss due to friction in the safety injection /RCP seal injection piping is determined as follows:

\\

l The ah is equal co the strong CCP developed head at runout f

flow.

This resistance is established during the CCP flew balance tes*.tg which limits CCP flow to the runout limit.

J The injection piping resistance (k) is equal to the' developed head of the strong CCF at its runout flow divided by the (runout flowrate)2

• dovetens,1 heari(runout flowrate)#

• Ah.,

• 1500 1't. 2

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(550 gpm) k = 4.96 x 10~3'ft./gpm2

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~ ATTAC) MENT B The resistance of the injection piping (Ah ), at the total' CCP flow:

f required to maintain 60 gpm through the weak CCP is:

2 Ahf = kQ or Ahf = (4.96 x 10~32) (210 gpm)2 = 219 ft.

Stso 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 Sot 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.

Thispressuredropmustbeoverdome by th= CCP= ir:- =rd=r t= deliver flew to the 'RCS at the hot leg /

pressurizer pressure.

Step 6:

Determine the elevational head between the RWST and the pressuriz~er safety valves.

e.g.

RWST elevation 160 ft.

CCP suction elevation

.- 100 ft.

l RCS cold leg injection nozzle elevation -

126 ft.

Pressurizar safety valve elevation 187 ft.

RWST to CCP suction 60 ft, 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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Sten 7:

Calculate the pressurizer safety valve relief prsssure.

e.g.

relief pressure = safety valve nominal relief pressure

+ 15 setting tolerance relief pressure = 2485 psig + 25 psig = 2510 psig (5798 ft.)

Step 8:

Determine the maximum RCS pressuri;:er pressure at which 60 gpm minimum flow.is-maintained through the weak. CCP..

Maximum RCS pressure = (CCP developed head at total CCP flowrate) -

i (injection piping head loss) - (head loss through RCS) - (eleva-tion head loss)

Maximum RCS pressure = 5842 ft. - 219 ft. - 116 ft. - 10 ft. =

5497 ft. = 2380 psig Comparing this pressure to the pressuri::er 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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IE Bulletin No. 80-18 July 24, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.

80-12 Decay Heat Removal System 5/9/80 Each PWR with an Operability Operating License (OL) or Construction Permit (CP) 80-13 Cracking in Core Spray 5/12/80 All BWRs with an Spargers Operating License (OL) or Construction Permit (CP) 80-14 Degradation of Scram 6/12/80 All BWRs with an Discharge Volume Operating License (OL)

Capability or Construction Permit (CP) 80-15 Possible Loss of Emergency 6/18/80 All nuclear facilities Notification System (ENS) holding Operating Licenses with Loss of Offsite Power (OLs) 80-16 Potential Misapplication 6/27/80 All Power React'r of Rosemount Inc., Models Facilities with an 1151 and 1152 Pressure Operating License (OL)

Transmitters with Either or a Construction Permit "A" or "D" Output Codes (CP) 80-17 Failure of 76 of 185 7/3/80 All BWR power reactor Control Rods to Fully facilities holding Insert During a Scram Operating Licenses (OLs) at a BWR or Construction Permit (CP)

Sup. 1 Failure of Control Rods 7/18/80 All BWR power reactor to 80-17 to Insert During a Scram facilities holding Operating at a BWR Licenses (OLs) or Con-struction Permits (cps)

Sup. 2 Failures Revealed by 7/22/80 All BWR power reactor to 30-17 Testing Subsequent to facilities holding Failure of Control Rods Operating Licenses (OLs) to Insert During a Scram at a BWR or Construction Permits (cps)

Enclosure a

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