ML20062K056

Application to Amend License DPR-77,to Resolve Concerns Addressed in IE Bulletin 80-18, Maint of Adequate Min Flow Thru Centrifugal Charging Pumps Following Secondary Side High Energy Line Rupture. Response & Justification Encl

ML20062K056
Person / Time
Site:	Sequoyah
Issue date:	11/18/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20027A635	List: ML20027A634 ML20062K056
References
IEB-80-18, NUDOCS 8011210276
Download: ML20062K056 (14)

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ENCLOSURE 1 i

SEQUOYAH NUCLEAR PLANT UNIT 1 PROPOSED LICENSE AMENDMENT  ;

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In order to complete the modifications recommended by IE Bulletin 80-18 'on adequate minimum flow through centrifugal charging pumps, we request that the following paragraph be added to license DPR-77.  ;

I l In conformance with IE Bulletin 80-18, TVA shall complete interim modifications identified in L. M. Mills' letter to J. P. O'Reilly dated October 16, 1980, to ensure adequate '

!- minimum flow through the centrifugal charging pumps.

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ENCLOSURE 2 SEQUOYAH NUCLEAR PLANT UNIT ,1 TVA'S OCTOBER 16, 1980, RESPONSE TO IE BULLETIN 80-18 ADEQUATE MINIMUM FLOW TO CENTRIFUGAL CHARGING PUMPS Response to Item 1 of the Bulletin TVA has completed calculations to determine if the Sequoyah Nuclear Plant unit 1 charging system would maintain adequate pump flow during parallel safety injection operation cnd determined that adequate flow would not be -

maintained. The detailed calculations outlined by the Westinghouse Electric Corporation letter (NS-TMA-2245) are included as Attachment 1.

Response to Item 2 of the Bulletin

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a. Modifications tre planned for Sequoyah unit 1 as described under Interim Modification I of the Westinghouse letter attached to the bulletin. These modifications include:

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

(2) Removing the safety injection initiation automatic closure signal from the CCP miniflow isolation valves.

(3) Modifying plant emergency operating procedures to instruct the operator to:

(a) Close the CCP miniflow isolation valves when the actual RCS ___ _

pressure drops to the calculated pressure for manual reactor 4

coolant pump trip.

(b) Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2,000 ,

psig,

b. As indicated in the Westinghouse Electric Corporation safety evaluation

( Attachment 2), if manual operator action is taken to close the CCP miniflow valves when the RCS pressure drops to the calculated pressure for manual reactor coolant pump trip (1,500 psig), no significant _

change in peak clad temperature (PCT) would be observed. Since tripping of the reactor coolant pumps is itself a manual operator action, it is our opinion that the additional requirement of closing the CCP miniflow valves (two handswitches) will not burden the operator and can be accomplished in the time necessary.

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c. The CCP miniflow valves are supplied with shutdown power via the diesel generators. The same post-accident monitoring instrumentation (powered by batteries and/or diesel generators) used to determine the reactor coolant pump trip pressure will be utilized to determine the need for opening or closing the CCP miniflow valves.

d. As indicated in the Westinghouse safety evaluation, the flow available from the CCP's with the modification in place, along with the operatec action' indicated in item 2.b above, will have a negligible effect on the safety-related analysis (note that part of Attachment 2 is dedicated to UHI plants).

e. Sire:e the results of the safety-related analyses evaluated in item 2.d indicate the insignificant effects of the interim modification and pro-cedure change, all technical specifications based on these remain valid.

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ATTACHMENT 1 SEQUOYAH NUCLEAR PLANT UNIT 1 '

MINIMUM CENTRlFUCAL CHARGING PUMP FLOW DURING '1WO PUMP PARALLEL SAFETY INJECTION CALCULATION FOR NRC IE BULLETIN NO. 80-18 4

Purpose Check capability to provide minimum pump flow during parallel safety injection with two centrifugal charging pumps (CCP's).

References

1. NRC IE Bulletin No. 80-18. -

2. Letter from T. M. Anders6n, Westinghouse Water Reactor Division, to V. Stello, NRC, dated Fby 8,1980, No. NS-TMA-2245.

3. Sequoyah Nuclear Plant Unit 1 Preoperational Test WG.1C data.

Calculations I

Following the format suggested in Reference 2, using data from Reference 3. .. ,

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. Step 1: Maximum developed head pump flow = 2,600 psid = 6,006 ft. @ .

73.1 gpm (pump 1A-1A) f 1

Minimum developed head pump flow = 2,470 psid = 5,705.7 ft. @ l 72.3 gpm ~(pump 1B-1B) i s ,

Step 2: Correction for testing error.

Test gauge accuracy = .257. x 3,000 psig = 7.5 psi (17.25 f t.)

+ 10 psi (23 ft.) reading accuracy = 40.25 ft. -

Maximum pump = 6,046.25 ft. @ 73.1 gpm Minimum pump = 5,665.45 ft. @ 72.3 gpm Step 3: From construction of pump flow curves, attached, minimum pump =

5,670 ft. @ 60 gpm

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Projection of weak pump head point on strong pump operating curve shows flow of 224 gpm.

Total flow from both CCP's gua'ranteeing 60 gpm to tie weak pump is 224 gpm + 60 gpm = 284 gpm Step 4: Determination of injection piping head loss.

From Reference 3, runout head of pump 1A-1A = 480 psi runout flow of pump 1A-1A = 490 gpm

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ATTACHMENT 2 .I ' ' . . . ~ 3 . 'p, .. WESTINGHOUSE ELECTRIC CORPORATION SAFETY EVALUATION - CEftTRlFUGAL CHARGiliG PUMP OPERAT10!! FOLLDWIl{G SEC0!!DARY SIDE HIGH EtiERGY LIllE RUPTURE .. l , i Reference 1: NS-TMA-2245,5/8/80 ,. . Reference i notified the itRC of a concern for consequential damage of one or more centrifugal charging pumps (CCP) following a secondary system high energy line rupture. Reference 1 included a calculational method and sample calculation to permit evaluation of this concern on a plant specific basis. Should a plant specific problem be identified, Westinghouse provided several recommendations for the interim.until necessary design j mod,ifications can be implemented to resolve the problem. These recommenda-tions inc1'uded two ' proposed interim modifications which included:

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

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

a. Close the CCP miniflow isolation. valves when the actual RCS pressure drops to the calculated pressure for manual reactor .-

- coolant pu:qp trip. , ~

b. Reopen the CCP miniflow isolation valves should the wide range ~

RCS pressure subsequently rise to greater than 2000 psig. g Prior to making this recommendation, Westinghouse evaluated the impact of the recommended operating procedure modifications on the results of the. various accidents which initiate safety injection and are sensit'ive to CCP flow delivery. The accidents evaluated in detail include secondary system ' ruptures and th'e spectrum'of small loss of coolant accidents. The analytical results for steam generator tube rupture and large loss of coolant accident are' not sensitive t'o a reduction in CCP flow pf'the magnitude'that results , from the recc= ended modifications. This letter functions to supplement Reference l'and identify the sensitivity of the accident analyses to the recemended modifications. This evaluation,is generic in nature. 1 _2 F - t Secondary System Ruoture_ . g ~ Sensitivity analyses have been performed for secondary high energy line ' ruptures to evaluate the impact of reduced safety injection flow due to - These analyses indicate an normally open miniflow isolation valves.  ! insignificant effect on the plant transient response. A. Feedline Rupture , ~ Following a feedline rupture, the reactor coolant ' pressure will, reach the pressurizer safety valve setpoint within approximately 100 seconds assuming maximum safeguards with the power-operated relief valves inoperable. With minimum safeguards, the reactor coolant pressure will ' not reach the pressurizer safety valve setpoint until approximately The time that the reactor coolant system pressure remains 300 seconds. at the pressurizer safety valve setpoint is a function of the auxiliary ' ] ~ feedwater flow injected into the non-faulted steam generators and the' ~ time at which the operator is assumed to take action. With the mini-flow isolation valves open,'the peak reactor coolant system pressure and the water discharged via the pressurizer safety valves are insignifi-cantly changed from the FSAR results.. B. Steamline Rupture ' . The effects of maintaining the miniflow isolation valves in a 'normally - , open position was also investigated following a main steamline rupture. For~ the condition II " credible" steamline rupture, the results of_the - transient with the miniflow valves open showed that the' licensing [ criterion (noreturntocriticalityafterreactortrip)continuesto ' The conditiori III and IV main steamline ruptures were also b.c met. , reanalyzed' assuming the miniflow valves wdre open. The results of / . th,e analysis showed that, even with reduced safety injection fl'ow ,in.to the core, no O!G occurred for any rupture. ~ 9

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I Small loss of Coolant Accidents .I Sensitivity analyses have been performed to evaliat'e the impact of red'uced safety injection flow on small break loss of coolant accidents (LOCAs). These analyses indicated that miniflow isolation can be delayed, but it must occur at some time into the small break LOCA transient in order to ~ limit the' peak clad temperature (PCT) penalty. The proposed modification delays miniflow isolation and reduces SI flow , delivered by approximately 45 gpm at 1250 psia during the delay time period. The impact of this modification was evaluated based on two isolation times:

1) The tim'e equivalent to the RCP trip time, and 2) approximately 10 minutes in the transient, or just prior to system drain to the break for the worst i

The second time was evaluated to determine the impact small break sizes. if the operator does not isolate miniflow within the proposed pre:,cribed time . The spectrum of sh.all break sizes are considered to encompass all possible small break sce'narios. Only cold leg break locations are considired ~ - since they will continue to be limiting in terms of PCT. . A. Very small breaks that do not drain the RCS or uncover the core..and ___ maintain RCS pressure above secondary pres,sure (< s2" diameter). l For these break sizes, it is quite possible that the operator may never isolate the miniflow line, since the pressure setpoint will ';. . not be reached, and continued pumped SI degradation will persist. . However, this'will have no adverse consequences in terms of core uncovery and PCT. flo core uncovery will be expected for the degraded SI case, similarly to the base comparison case with ful.1 SI. The only effect would be a slightly lower equilibration pressure for a , given break' size. 'B. Si5511 breaks that drain the RCS and resul.t in the maximum cladding C Umperatures(2"< diameter <6"). . This range of break sizes represents the worst'small break size for - - . - - - - .- -- . , - - _ __ } . . ~ - _4 .. most plants as determined utilizing the currently approved October 1975 Evaluation Model version, as shown in WCAP-8970-P-A. IfmiYlificwis isolated at the RCP trip setpoint rather than the "S" signal, a reduc-tion in safety injection flow of less than 45 gpm results, averaged for the approximately 50 second period of time separating the two events. This reduction in RCS liquid inventory "results in core uncovery less than one second earlier, and has a negligible impact on PCT. If mini-flow is isolated at the time of core uncovery, or approximately 10 minutes for break sizes in this range, a greater reduction in RCS liquid inventory results in a core uncovery 10 seconds earlier in the transients ' resulting in less than a 10*F PCT penalty for the worst size small break. This viould not result in any~ present FSAR small break analysis becoming more limiting than the corresponding large break i.0CA FSAR analysis. . I If minificw isolation does not occur at any time into the transient for this category of small LOCA, a PCT penalty of 200 F or more could occur. C. Small break sizes larger than the worst break thrcugh the intermediate break sizes (> 6" diameter)'. Break sizes in this range _have been determined to be non-limiting for small break utilizincj the currently approved October 1975 Evaluation Model, WCAP-8970-P-A. If miniflow isolation occurs at the RCP trip time for these break sizes, the negligible effect cm PCT presented , above also applies. Similarly, if isolation occurs prior to core uncovery, the'small (< 10 F) PCT penalty will resuras well. However, for these larger break sizes, the time of first core uncovery occurs prior to 10 minutes. If miniflow isolation is not prformed until , 1,0 minutes, reduced SI will be delivered during the ere uncovery time, which can have a greater impact on PCT. Studies indcate a potential PCT penalty of 40*F resulting for these non-limiting break sizes if miniflew is not isolated until 10 minutes. This is nut expected to shift the worst break size to larger bre'aks, since titese breaks are typically hundreds of degeses'1.s than snaller liastinc small becak:: analyzed with the currently approved Evaluation Madrii. b - w+.. . . - - .S... . It ., For all FSAR small LOCA analyses, one complete train ( failure is .acsum  ! 1; 4 is clear that two charging pumps without miniflow isolation provides more . ' The 16. pact presented in this flow than one pump with miniflow isolation. , evaluation maintainsIf both thepumps one were train failure operating, theand assumes no PCT results tion for the remaining pump. would be much lower than present FSAR calculations In this situation, eventhe if miniflow is ' tion is not assumed to occur for the two pump case. plant FSAR sma'll break calculations remain conservative. . These sensitivity studies form the basis for the recommended interim The accidents evalu - ' modifications to the emergency operating procedures. Further, ated are relatively insensitive to the recommended modifications. the accidents evaluated will give results that satisfy acceptance criteria ~ , as .long as the CCP miniflow is i'solated with'in 10 minutes of event in ' However, small LOCA sensitivity studies with one SI train operating confir i hin.10 minutes. .that small LOCA analyses require miniflow isolation w t To comply with the recommen'ded modifications, the operator can is flow at any point in the depressurization transient prior to RCS pressure Should a repressurization transient oct,ur, reaching the RCP trip setpoint., the operator can open CCP miniflow at any point between the RCP trip se Such operator actions will ensure that plant accidents point and 2000 psig. satisfy acceptance criteria and protect the CCPs from consequential damage during the repressurization transient that accompanies a secondary systcm high energy line rupture at high initial power levels. . * \  ;. . da e

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CENTRIFUGAL CHARGING PUMP OPERATION FOLLOWING SEC0"DARY SIDE HIGH ENERGY LINE BREAK The small loss of coolant accident (LOCA) section of the' main report was ' generated primarily for plant applications which do not include upper . This supplement pro- ' head injection (UHI) as part of the ECCS design. vides additional small LOCA information for UHI plants and, together with the main report, assesses the impact of delayed miniflow isolation for small LOCAs for UHI plants. ' ' T he model utilized to determine the SI sensitivities and't'o identify the worst small break size discussed in the main report was the October This model 1975 Model 'WCAP-8970-P-A) version of the Evaluation Model. UHI small break analyses are is not yet approved for UHI plant analyses. However, sensi-performed with the December 1974 small break version. tivity studies performed to determine the effect of pumped SI on small ~ break LOCA PCTs utilizing the December model yielded nearly identical $ results as presented in the main report. This is expected since the model changes included in the October model do not affect the basic vessel inventory and core boiloff relationships that determine the impact of changes in pumpe1 safety injection on PCT. . An important difference in UHI plant small break analysis results as compared to similar non-UHI plant analysis results is the small break size resulting in the highest PCT. This break size is generally greater for UHI plants than for similar non-UHI plants because of the additional safety injection flow provid, ed by the UHI accumulator 'at relatively high RCS pressures. The worst small break size for UHI plants may be a six inch diameter break or larger. The main report identified breaks of this size and larger as non-limiting small break sizes. While this is-true for non-UHI plants, it is not accurate for typical UHI plant small break analyses. Therefore. the stated 40 F potential penalty for 9 j 4 l . . .

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six inch breaks applies to the worst break for UHI plants fog the case where miniflow isolation is delayed until 10 minutes. It is Westinghouse's opinion, however, that the stated' penalty of 40*F i~s conservatively high and bounding for UHI plants, for the following reasons: a) The 40 F penalty was based on, sensitivity studies performed l assuming an approximate 20" reduction in total HPI flow. However, the anticipated 20" reduction actually applies only to the charging pumps. Intermediate head SI pumps are not affected. Therefore, total HPI for plants with intermediate head SI pumps, which includes all UHI plants, , will result.in less total degradation, and thus a smaller PCT penalty. The high pressure accumulator on UHI plants has a similar effect of , reducing' the total HPI degrada' t ion due to the delay in miniflow isolation. b) The UHI accumulator is a significant source of liquid mass inventory for breaks greater than or equal to six inches in diameter. This addi-tional mass delays the core uncovery time as ' compared to the same size break occuring on a similar non-UHI plant, since more liquid mass must exit.from the break prior to core uncovery. The ' delay in core uncovery' results in clad heatup at a lower power level caused "by the decay in residual core heat. Therefore, clad heatup rates are slower which also tends to reduce the sensitivity to changes in HPI delivery rate. __ In conclusion, the sensitivity provided for six inch diameter and larger break sizes in the main report represents the worst break size range for UHI plants. The stated 40 F PCT penalty for breaks of this . size ', resultant from a 10 minute delay in miniflow isolation is a conservatively . high and bounding value for liHI plants, for the reasons stated above. If miniflow is iso' lated at the time of RCP trip, the negligible impact on PCT discussed in the main report applies for UHI ,plar.ts as well. The <10 F penalty resultant if miniflow isolation occurs prior to core t ~ uncovery also applies to UHI plants, with the added benefit that this , event occurs later in a UHI plant transient than for a non-UHI plant i ' transient of the same break size, allowing more time for the operator , to act. , l . ~ .o r - *. ENCLOSURE 3 SEQUOYAH NUCLEAR PLANT UNIT 1 JUSTIFICATION FOR CENTRIFUGAL CHARG1NP PUMP MINIFLOW VALVE MODIFICATION i i The centrifugal charging pump miniflow valve modification has been deter- t mined to be an unreviewed safety question caused by uncertainties involved (' with the manual operator action required. In the Westinghouse Safety Evaluation, sensitivity studies were used to determine the effects of reduced ECCS flow (s45 gpm) on the design base accidents. Since Sequoyah [ is an upper head injection plant, the hypothesized peat clad temperature  ; (PCT) increase of 400F applies to the worst case small break if no j operator action occurs until 10 minutes after the break initiation. How- t ever, manual closure of the miniflow valves at the time the RCP's are i tripped would result in a negligible increase in peak clad temperature.  ; Since the PCT penalty is based on sensitivity studies done for non-UHI . plants, a number of assumptions made do not apply to UHI plants and  ; indicate the existing FSAR small break analysis may be bounding or close enough so to be insignificant. These assumptions include: (1) A 20-percent reduction in HPI flow - Since Sequoyah has both high head and intermediate head SI pumps, the 20-percent reduction in HPI flow applies only to the high head pumps so that the overall HPI flew reduction is on the order of 10 percent (or less for worst case small break). b (2) No upper head injection flow assumed - This additional flow would act to delay the core uncovery time resulting in clad heatup at a lower power level caused by the decay in residual decay heat and result in a lower peak clad temperature increase.  ; In addition to these assumptions, the small break analysis Westinghouse utilized assumed a core power peaking factor of 2.32. Sequoyah is presently limited by technical specification to a peaking factor of 2.237 adding an additional margin in PCT. i If, disregarding the above arguments, the bounding PCT increase of 400F is assumed for the Sequoyah worst case break (8"), the resulting PCT would go from the FSAR valve of s1490 F to s,1530 F. The PCT is still well below the worst case large break valve of 21900F. Position indication for the CCP miniflow valves is provided by lights on the handswitches and separate status lights on the main control panel. Since the proper position of this valve is dependent on system pressure, the valve position alarm system described in section 6.3 2.2 of the Sequoyah FSAR cannot be applied to these valves. It is our opinion that j the redundant valve position indication along with explicit instructions in l the Emergency Operating Instructions for manual operation of these valves are sufficient to ensure proper valve position in the event of a safety injec tion. The final modification will address the need for a permanent valve position alarm system. We believe that the proposed interim modification does not adversely effect the consequences of the accidents analyzed in the FSAR and does not con-stitute a significant hazard consideration. <