ML18082A955

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Forwards IE Bulletin 80-18, Maint of Adequate Min Flow Through Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture. Action Required
ML18082A955
Person / Time
Site: Salem  PSEG icon.png
Issue date: 07/24/1980
From: Grier B
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Schneider F
Public Service Enterprise Group
References
NUDOCS 8008220077
Download: ML18082A955 (15)
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Text

{{#Wiki_filter:1-- 1 UNITED STATES e-NUCLEAR REGULATORY COMMISSION REGION I* 631 PARK' AVENUE KING OF PRUSSIA, PENNSYLVANIA 19406 Docket No. 50-272 50-311 Public Service Electric and Gas Company ATTN: Mr. F. W. Schneider Vice President - Production 80 Park Place Newark, New Jersey 07101 Gentlemen: July 24, 1980 The enclosed IE Bul*l eti n No. 80... 18, 11Mai ntenance of Adequate Minimum Fl ow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture, 11 is forwarded to you 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 i,n conduct of the review and preparation of the report(s) required by the Bulletin. Please estimate separately the manpower associated with corrective actions necessary following identification of problems through the Bulletin. If you desire additional information regarding this matter, please contact this office. Sincerely, ~tJ-Y11. rle!ti~.

Enclosures:

1.

IE Bulletin No*. 80-18 and Enclosure with 2 Attachments

2.

List of Recently Issued IE Bulletins CONTACT: E. G. Greenman (215-337-5267) cc w/encls: F. P. Librizzi, General Manager - Electric Production E. N. Schwalje, Manager - Quality Assurance R.* L. Mittl, G~neral Manager - Licensing and Environment H. J. Midura, Manager - Salem Generating Station 8008220 077 UNITED STATES NUCLEAR REGULATORY COMMlSSION O~FICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 SSINS No.: 6820 Accession No.: .8005050062 IE Bulletin No. 80-18 Date: July 24, 1980 MAINTENANCE OF ADEQUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING PUMPS FOLLOWING SECONDARY.SIDE HIGH ENERGY LINE RUPTURE Desc~ipti~n 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 11Part 21 11 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 res.ult 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 be adequate to insure pump cooling, and.. resulting pump dama'ge 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 minimum flow. The recommended modifications assure availability of the necessary minimum flow by assuring that the mini-flow bypass li.ne 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.

IE Bullet in No.. 80-18 July 24, 1980 Page 2 of 3. Actions to be taken by PWR licensees listed in the enclosure as 11operating plants; 11 and those listed as 11non-operating plants" which are nearing licensing* are listed below:

1.

Perform the calculations, outlined in the enclosure, for your plant.

2.

Ir availability of minimum cooling flOw for the CCPs is not assured for all conditions by the calculations. i.n 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 con-ditions. If modifications are made as. descri.bed 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 lor 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 act ion, 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 . (L-OCA, HELB, etc.).

e.

Justify that all Technical Specifications based on the ltem 2.d analyses remain,valid.

3.

Provide the results of calculations performed under ltem l, 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

  • Those listed in the enclosure considered to be "nearing licensin~' 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. 80-18 July 24, 1980 Page 3 of 3 terminatiOn criteria. If a "minimum flow bypass" line is present which remains open duri.ng high pressure injection, and if that line guarantees that minimum cooling flow will be provided to the pumps under such condi-tio.ns, 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 appropriate NRC Regional Office and a copy forwarded to the Director, NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, O. C. 20555. Approved by GAO, 8280225 (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 5/8/80 and Enclosure with 2 Attachments

  • Those considered to be 11nearing licensing 11 are: North Anna 2, Diablo Canyon 1, McGuire, Salem 2, and Sequoyah.
--~ *. *,

__ C--------~ *, Enclosure to IE Bulletin 80-18

.
  • Westinghouse Electric Corporation Water Reactor DMsions Nuciear Tectmol0£Y Olvlsloo Bo~ 355 Mr. *v. Ste 11 o. Di rector Office of Inspection and Enforcement U. S. Nuclea,.. Regulatory Commission 1717 H Street Washington, D. C.

20555 PfttshlB'itl ~ l~ May a, 1980 HS-TMA-2245

Subject:

Centrifugal Charging Pump Operation Following Secondary Side High Energy Line Rupture

Dear Mr. Stello:

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. *.. \\. This letter is to confirm the telephone conversation of May 8, 1980 be~*een Westinghouse and Mr. Ed Blackwood of Divi*sion of Reactor Operations Inspec~ion. *J:-,..

    • ~:*

Office of Inspection and Enforcement) regarding notificiltion made pursuant to* Title 10 CFR Part 21. A r~v1aw ~f tha* West*inghouse Safety Injection (SI) Termination Criteria., following a secondary s1de high energy 1ine rupture (feedline or steaml1ne* rupture at high initial power levels) has revealed a potential for conse..

  • quentia1 damage of one or more centrifugal charging pumps (CCPs) before the SI te:rmination criteria are satisfied and CCP operation tenninated.

Such consequential damage may adversely impact long-term recovery operations ".*;,_::;.. :,;_~i.*:::.::*~:i :: for the initiating event and is not permitted by design criteria._ This -*

  • concern exists for:* pl ants which.utilize* the CCPs as Emergency Core Cooling - * *

>::~~~~-::<*). S1stem (ECCS) pum?s. whero*the CCPs are automatically, started, and where the* ~;-~~~*~-.'.-:~- CCP 1ninif1ow isolation valves are automatically isolated upon safety injection

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  • ":.i;. initiation. Attachment A identifies plants potentially subject. to this

~~/~i;\\~t.& concern.~~~ surrrnary of the concern and recorrmendations follow.

  • /;~~F{:<:*~~: *F*~:#fitt;g* a.secondary sid.e hi~h-energy line rupt~re"and associ.ated reac:tor-~

'.}(:.(.'.: ::,_,*trip~ '.Reactor Coolant System (RCS) pressure and temperature initially decr;ea~e *. ='~.-{TT...,-'\\.=~ S~fety injection is actuated and the CCPs start to increase RCS inventory._* _ ~?{::+;~'~.~:,.~:fifiactor Cool ant System pressure and temperature subsequently increase due.. {:-..* * =t~J~~;,,-~'<' to the loss of secondary ;nventory, steamline and feedline isolation, RCSJ'.c.,~* 2*.'.ii/:*.. :::.'1\\ inventory addition and reactor core decay heat generation. The accident <<~ .~:' ' .:i~~:~L':;;.::~**" scenario may vary with *rupture size and specific -plant design, but it will\\'* *..

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,;~I~;::i',*::\\~:,':,:,* develop into a RCS heatup transient with accompanying increase in RCS pressurt;~*

~)}}~:~:~::fJ:.. *As RCS pressure increases, the pressurizer power--operated relief valves>;:.:;:*.'*:.

.. ~::-i:~,;0~::::: (PORVs) ar-e designed to 11mit RCS pressure to 2350 psia. Although these* ~.L' ~ J~.~*::r_\\~!(_*va1ves are nonnally available, they are not ~esigned a! safe~y-relate.d equ.~_p..;*.. :** ,>< 0;K}*:~';..... ment.. rt can be postulated that:. due to e1tner loss 01 offs1te power,*:.~***'.~:. I.. . *I ":I ~ *: .:~

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  • Mr. V. Stello

-2* May 8, 1980 NS-TMA-2245 adverse environment 1nside containment, the pressurizer PORV in manual mode, or the ?ORV block valve in a closed position, due to PORV leakage~ the pressurizer PORVs may not be operable. As a l"'esu1t of the RCS hea:tup and inventory increase, the RCS pr~ssure could rise to the pressurizer safety valve setpoint of 2500 psia within approximately 200 seconds-arid remain at that pressure until transient "turnaround. 11 Transient turn-around0 can occur between 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 tenninated. During this period, the minimum f1ow required for CCP opera-tion must be satisfied by flow to the RCS since the CCP miniflow isolation.* valves are automatically c*losed on safety injection init1afion:, - This requires that the CCPs be able to de11ver their minimum required flew to the RCS at' the safety valve setpoint pressure. './.. To evaluate this concern, Westinghouse has developed a calculational method and hes reviewed typical CCP head versus flow perfonnance curves and other representative plant parameters. The calcu1ationa1 method considers the ::*. effects of safety va 1 ve re 1 i ef *setpoi rt a.ccuracy ~ RCS piping resistance, ECCS piping resistance. number of CCPs operat1ng~ technical specification allowable CCP head degradation, and uncertainties associated with in-plant verification

  • testing. The analyses for two CCP operation:. the best estimate condition,... is similar to the 6n3lysis for one CCP operation except that the flowrate us~d to detennine ECCS piping 1 ine loss must ensure the minimum flow through e~fch pump.

For example. at a specific required head, the pump with the higher.**-~ developed head may be required to deliver greater than the minimum fl ow in:'.',? . order to penuit the lower *head pump to meet the minimum flow requirement. A-- >:: - *:,:'.. This generic evaluation indicates that. sufficient_ flow *tc sathfy CCP minimum .~\\~'.-,~~:,*:: flow requirements to avoid_ pump degradation_ may not be ensured for a secondary i1t1,t~**.. ~~~~:: :1! :~i:~~i;i~~:;~~iEf ~~i:if ;;~:iii'.~!~;ii;~~::;f i~~~i;~~ e~:s ~:\\'.:;..'.**... - ** *can be used in this evaluation. Based on Westinghouse genetic review, satis-

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  • factpry results may not be obtained. Should a plant specific concern.be.. r:

~.~i.t~~-\\~;*.-* ~~mH~:~~"~~~~r~~~ 0 ii~~~mTI~~~m~:~~~:~~*~~hr!~:¥~~!~ c~~~i~F~f. ~~i~ alignment and operating procedure changes to provide backup to the pressurlzP.r [~c:f ::~ *.. ~~~~:~!~~:~~~:~ ~~~~~~:~m~~iH~~~*~:!~J~:~~~;e~1~K~~~~:~ !~~~~~.~ !t ~;~!~~c- ****** ,,::-,_,_;:~::~;.- -~*_:: to limit challenges to the pressu r-i zer safety va 1 ves, and -{ b) review the ;~;*

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)3r.~ maintenance operations and technical specifications for the backup {;.e.,.:third)

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  • charging pump to maximize it~ availability.for l?ng-ier'!l re~overy from a.~.~

t;~.:d.:-;,..c.:-,:/: secondary side _rupture. The$~ reccmnenda~ions) in cQmb1nat1on with the in,;er1m §l~Jtt*/.* *_.*. . - :c __...

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  • Mr. V. Stello

' May 8, 1980 NS-TMA-2245 m-ociifications described below~ are considered sufficient to address th1s con-cern in the interim until necessary. design modifications can be implemented. t-* Interim Modification I This interim modification is preferred and requires that component cool1ng* water be supplied to the seal water heat exchanger following safety 1njection initiation in order to provide cooling for CCP miniflow.

1. Verify that CCP minifiow return is aligned d1rectly to tht! CCP suction*.

during nonnal operation with the alternate return path to. the volume control tank isolated (lock closed). _ ? ~ Remove the safety injection initiation automatic closu..-*e signal from.";~ the CCP miniflow isolation valves. ,~~- Modify plant emergency operating procedures to instruct the operator,\\o:

a.
b.

Close the CCP miniflow isolation valves when the actual RCS pressu~e drops to the calculated pressure for manual reactor coo 1 ant pump trip. Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2000 psig * .. ~:;!~ *.. . ~*; .::-.~.

    • 4-
  • ' _.,.. :2 Intarim Modification II

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This modification.is'an alternative for plants in which component coo1ing-;~-.-~-- water is not supplied to th~ seal water heat exchanger following safety ,~;{_ injection initiation. Since miniflow cooling is not provided~ this a1tern'!;. tive directs miniflow to the volume control tank to permit ;the CCP minimu~/. 1 ~** f1ow requirements to be satisfied with cool uncirculated water. lhe volume.: control tank acts as a surge tank to co11ect minif1ow following safety .* -.-_:tf_::_**-***;.- __ ** injection initiation with excess flow directed-to a holdup tank via the .:z;.:.:, .. \\:.:~:;-'<!.:~_-i volu~ control tank relief valve.. p -:~~j-:

1.

Align the CCP m1n1f1ow to the. volume control tank during normal opera~'].{~*. ~_,it~rl *; m~~:;!~*mt:~~f ! f~y ~~;~3~nh~~~~~~~~o~~~~~m~:ia ! rf~~: ~'.*... '. '

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- Same as* Interim Mod1fica:tion I_, Item 3~ ~+/-'.- * -;,z_.

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  • Ste 11 o May 8, 1980

. ~.. NS-TMA-2245 Based on the generic evaluation~ Westinghouse has initiated efforts to perform additional plant specific analyses for non-operating plants and to develop, design modifieations to resolve any identified concerns. The modifi,ations will be designed to safety-related standards and will be compatible with ~~ Westinghouse SI tennination criteria and standardized technical specifications

  • If you re.quire further information, please call Ray Sero (412-373-4189) of. my staff.
    -~ -

TMA/jaw Attachments ~-. .. : ~. ~... .. ' ~ "/Jtruly yours., Y2tli~-~------ T. M. Anderson~ Manag~r. Nuclear Safety Department . *:~,: ** ;.. ti. ~. . ~~; -. ;. ~':

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North Anna l & 2 Beaver* Valley 2 Far1ey 2 OPERATING PLANTS NON-OPERATING PLANTS Shearon Harris 1. 2, 3 & 4 Virgi1 Surrmer ~**..

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e. P.TTACHMENT B MINIMUM CENTRIFUGAL CHARGING PUMP FLOW DURING rwo PUMP PARALLEL SAFETY INJECTION OPERATION 1.:~-.... --...

~.* .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 a_ctual plant,: data and determines what actual CCP developed head at the min.:iflow flowrate must be available.. '~'-' ~ ~*-

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~.*.-**. Individually determine the dave1oped head of each CC.P at the mf~.1-

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flow flowrate of 60 gpm from field test data~ (two pumps for l

  • 4-loop plants o.nd three pumps for 3-loop plants)

. { Step 1: Sample: Maximum developed head pump 257*1. 4 psi d = 5940 ft. @ 50 gpm Minimum developed head pump 2554.1 psid = 5900 ft. @ 60. gpm l

e ATIACHMENT B )/* . *.r:

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,*_-* !~ .* ~. Step 3:

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-.~ Detennine total CCP flow. Construct a pump curve fo: the maxi-.-~-* mum head pump that _is parallel to the actual "as-bu"ilt 11 vendor<:.: pump curve and passes through the above detennined developed head at th~ miniflow flowrate which is the measured developed --~ .. head p 1 us the deterini ned measurement, accuracy. **. ment Figure 1.)

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(See ~ttach- \\~ Use this head versus flow curve to detent'line the flow delivered= ~!.I by the maximum* head* pump (strong pump) at the devc 1 oped head of{'. the minimum head pump (weak ~pump) at the miniflow flowrate --:-;~}~:: (i.e., 5842 ft. as detennined 1 n Step l)._

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Sample: As illustrated in Figure 1 ~ the de1ivered flow of the-;~:- strong pump at-5842 ft. is l5Q gpm. Therefore~ the:*<{~ ~--'- total flow from both CCPs which auarantees that the * ~\\i: 1.* - weak CCP will be delivering at least 60 gpm is 210 gpmf~. ~- (150 gpm + 60 gpm). c~:~. ,' ---;:.. t~-..

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--*. *~* .,.... -~- Step 4: Detennine Injection Piping Head Loss. The head loss due to _..,~: ~-'P~ fr~ction in the safety injectio_n/RCP sea1 1njec_t_ion pJping _is- ~-~--~ ~- detenni ned as f o 11 ows: _..... -::, :-:-:-- : :_J;*j-:,:_ _ -- _:-.<* _-:<**,.,:::. :-.. __ :. *,-~~:~'::~f-4:.". *. _

  • The _ Ahf is equa 1 to* th~ strong CCP developed head at. runout. :~<~lp/'-.' *... *
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This re.sistance_ is _:established during the CCP ficrri* -~ ~:-\\~~-'.~;: * -~-::~*- balance testing which limits CCP flow t_o the run_out l_1m1t. *.: <:?t./~,. -... The inject-ion pip.ing resistance (k) is equa~ ~o the *~ev171u*p~d- *:it:>: head of the stron~ ~CP at its runout flow ~1~1ded.by the ~, _-_.'~/:-_;:lf*_. _ _ (run out fl owra te} * - * *.

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k = devel coed head

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,'"'Step 6: .* -**.e ATTACHMENT a * ~* *. The resistance cf the injection piping (t.hf), at the total CC?.Jlow required to maintain 60 gpm through the weak CCP is: -~,* ~. -':.1 *..

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_.tL Detennine head loss through the Reactor Coolant System. '}~ Consider that the reactor co61ant pumps are operating, therefor~,* ~ J..... the pressure drop from the CCP cold leg injection nozzles throi1gh *~': ".~- the reactor vessel to the pressurizer surge 1 ine off the hot 1e;g:_ at full RCS flow ar£> to be inciuded. This pressure drop.is ti~. L approximately SQ psid (116 ft.) for 4-loop plants and 48 psid.ft:.'

  • (111 ft.) for' 3... 1aop plants. This pressure drop must be ~v~rto~

by the CC:Ps 1n order to. deliver flow to the *Res at tile *hot leg/:;~;~ pressurizer pressure. Detennine the elevational head between the RWST and the safety valves~ . e.. g. RWST elevation -:.<

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~- ~*J 1:?~ ATIACHMENT B Step 7: Calculate the pressurizar safety valve relief pressure. e.g. re1ief pressure "'= safety valve nominal relief pressure + 1% setting tole~ance -~: . ~-~

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~* relief pt"essure = 2485 psi.9 + 25 psig

  • 2510 psig (5798 *ft.)

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-.** ".'c. ~~~~;*. SteE...§i_ Determine the maximum RCS pressu.r{zer pressure at which 60 gpm ;;:* ~!:-. . *::It ."'f.*. minimum flow is maintai_n~d through the we*ak CCP. .*)S,.. . **~~~;*,_ .*-Maximum RCS pressure = i (injection piping head

tion head loss)

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(CCP developed head at total CCP f1owrat"e) tr.*. 1oss) - (head loss through RCS) - {eleva~-

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Maximum RCS pressure 5497 ft. ~ 2380 psig = 5842 ft. - 219 ft... 116 ft. = :B:~. 10 ft. l:.. ,f*",. ' -.~*~*;*. ~....... .... ~.. ', . -~ *.. -*~*~- ~ Comp~ring this pr-essure to the pressurizer safety valve* re1ief*l:.. 'pressure (Step 7) of 2510 psig, it is evident that the *so gpm .**~E. 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 No.

Subject Oate Issued Supplement 2 t-0 80-17 Supplement 1 to so-11 80-17 80-16 80-15 so~14 80-13 80-12

  • 80-11 80-10 F'ailures Revealed by T est*i ng Subsequent to Failure of Control Rods to *Insert During a* Scram at a BWR Failure of.Control Rods*

to Insert During a Scram . at a BWR. Failure of Control Rods to Insert During a Scram at a BWR 7/22/80 7/18/80 7/3/80 Potent i a 1 Mi sapp li cation of 6/27 /80 Rosemount Inc.,.Models 1151 and 1152 Pressure Transmitters with Either 11N' or 110 11 Output Codes Possible Loss Of Hotline 6/18/80 With Loss Of Off-Site Power Degradation of Scram Discharge yolume Capability Cracking In Core Spray Spargers Decay Heat Removal System Operability Masonry Wall Design Contamination of Nonradioactive System.and Resulting Potential for Unmonitored, Uncontrolled Release to Environment 6/12/80 5/12/80 5/9/80 5/8/80 5/6/80 !~sued To All holders.of BWR power reactor OLs All holders of a BWR OL or CP All holders of a BWR OL or CP All holders of a power reactor OL or CP All holders of a power reactor OL anci *fuel;*cycle licensees connected to the Emergency Notification System All holders of a BWR OL All holders of a BWR OL All holders of a PWR OL All holders of a power reactor OL, except Trojan All holders of a power reactor OL or CP}}

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