ML20198G509

Provides Corrections to Attachment 4 of Re AFW Design Info.Background Discussion of Errors Found Along W/Description of Each of Specific Correction Provided

ML20198G509
Person / Time
Site:	Prairie Island
Issue date:	08/29/1997
From:	Sorensen J NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9709040123
Download: ML20198G509 (55)

Text

{{#Wiki_filter:Northein States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minresota 55089 August 29,1997 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Corrections to Attachment 4 of February 4,1981 Letter Related to Auxiliary Feedwater Design Information By letter dated February 4,1981, Northern States Power transmitted information related to the Prairie Island Auxiliary Feedwater System. It has recently been discovered that Attachment 4 to that letter contained several errors related to the design of the Auxiliary Feedwater System. This letter provides, for the information of the NRC Staff, a corrected version of Attachment 4 to the February 4,1981 letter along with an explanation of the corrections. provides a background discussion of the errors found along with description of each of the specific corrections. Enclosure 2 is a copy of Attachment 4 to the February 4,1981 letter marked up to show the corrections. Enclosures 3,4 and 5 are copies of related correspondence being provided for background information. = In this letter we have made no new Nuclear Regulatory Commission commitments: Please contact Gene Ecknolt (612-388-1121) if you have any questions related to the information provided by this letter. s [}' OO k Joel P Sorensen 9 Plant Manager Prairie Island Nuclear Generating Plant ~ @@{${$!$i[l!$fN,$}$,kk 9709040123 970829 fDR ADOCK 05000282 L PDR jf waiu u m

- USNRC NORTHERN STATES POWER COMPANY August 29,1997 Page 2 c: Regional Administrator - Region 111, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg

Enclosures:

1. Discussion of Corrections to Attachment 4 of February 4,1981 Letter

2. Attachment 4 to Letter Dated February 4,1981, " Basis for Auxiliary Feedwater System Flow Requirements", With Corrections Noted

3. Westinghouse Letter PlW-P-540, dated October 1,1969, "AuxilNry Feedwater System Sizing."

4. Westinghouse Letter PlW-P-519, dated September 5,1969, " Auxiliary Feedwater System Criteria."

5. Pioneer Service and Engineering Letter PIP-N-353, dated October 6,1969,

" Auxiliary Feedwater System - Feedwater Line Rupture Accident." 9

ENCLOSURE 1 Discuss!on of Corrections to Attachment 4 of February 4,1981 Letter Discrepancies Related to Main Feedline Break Analyses Following the event at Three Mile Island, the NRC requested that licensees perform several actions related to the Auxiliary Feedwater (AFW) System. One of these actions was to provide information regarding Auxiliary Feedwater System design basis flow requirements for different transient and accident scenarios. NSP contracted Westinghouse to assist with responding to this action. Northern States Power then transmitted the Westinghouse prepared document to the NRC as Attachment 4 to a letter dated February 4,1981 (Reference 1). In Attachment 4 to Reference 1 NSP refers to an auxiliary feedwater flow requrement of 400 gpm to a single steam generator 10 minutes after the initiation of a main feedline break event. The inputs for this analysis were shown in Table 2-1 of Attachment 4 to Reference

1. Key points in Table 2-1 are that the reactor trip following a main feedline break is shown to occur at 20 seconds into the event (plus 2 seconds for control rod movement) resulting from low steam generator pressure. The reference document for this information is Westinghouse letter PlW-P-540, dated October 1,1969 (Reference 2) which is specifically referenced on page 4-11 of Attachment 4 of Reference 1.

A review of Westinghouse letter PlW-P-540 (Reference 2) indicates that the analysis of the main feedline break were performed to confirm the adequacy of auxiliary feedwater flow rates of 200, 250, and 300 gpm to the intact steam generator 10 minutes after a main feedline break accident. There is no mention of an analysis at 400 gpm in Reference 2. The analyses discussed in Reference 2 were based on a reactor trip occurring at 10 seconds (plus 2 seconds for control rod movement) due to a high containment pressure signal. Thus, the information provided in Attachment 4 to Reference 1 does not agree with the cited reference for this information. Further investigation of this discrepancy between the information in Attachment 4 to Reference 1 and Reference 2 was performed. During this investigation an additional Westinghouse letter was recovered (PlW-P-519, dated September 5,1969, Reference 3). - This letter is referred to in Reference 2 as having the same analysis basis. The analyses in Reference 3 indicate that 400 gpm is required to be delivered to the intact steam generator within ten minutes of the reactor trip, with the trip occurring at 20 seconds due to low steam generator pressure. The analyses documented by Reference 3 appears to be the actual basis for the information provided in Attachment 4 to the February 4,1981 letter. Reference 3 documented a main feedline break analysis performed by Westinghouse which assumed the reactor trip occurs due to low steam header pressure at 20 seconds into the event (plus 2 seconds for control rod movement). The timing of the reactor trip is u.

Enclosura 1 Aupst 29,1997 Page 2 l important as this determines the amount of stored energy in the RCS. Using the low steam header pressure signal was intended to bound both inside and outside main feedline break scenarios. That is, for the main feedline break inside containment, the reactor trip would actually occur due to high containment pressura prior to the low steam header pressure signal; which would reduce the severity of the trancient. Subsequent discussions occurred between Westinghouse and Pioneer Service and Engineering (Prairie Island Architect Engineer) regarding Auxiliary Feedwater System capabilities, and new main feedline break sensitivity analyses were oerformed. The results of those analyses are documented in Reference 2. Reference 2 documents the results of an analysis specifically performed for the inside containment main feedline break. In this event, the reactor trip occurs on high containment pressure (conservatively assumed to occur at 10 seconds) with an additional 2 second time delay for control rod movement. With the exception of the reactor trip initiator, the basis for this analysis is the same as that in Reference 3. The conclusion of the Reference 2 analysis is that 200 gpm is sufficient to preclude over pressurization of the primary system. Following receipt of the Westinghouse analysis (Reference 2), Pioneer transmitted the analysis with h corresponding evaluation (Reference 4). This evaluation provides confirmation of the adequacy of the Auxiliary Feedwater System design (i.e., two pumps, 200 gpm each). Furthermore, Reference 4 provides discussion of conservatisms that are present in the Westinghouse analysis. Based on these conservatisms, Pioneer considered that there was margin in the Auxiliary Feedwater System, and that it was not necessary to increase the system capability to provide any additional margin. In order to provide an independent validation of the Westinghouse analysis in Reference 2, the NSP Nuclear Analysis Department (NAD) performed a main feedline break analysiF. In the NAD analysis, an auxiliary feedwater flow of 200 gpm at 110"F was assumed to be delivered to the intact steam generator starting ten minutes after the reactor trip. The reactor trip was assumed to occur at 10 seconds after the main feedline break plus 2 seconds for control rod movement. The analysis showed that a reactor trip would occur in less than 10 seconds due to containment pressure increase greater than 4 psi. The results of the analysis showed that reactor coolant system pressure remained within acceptance criteria and the core did not uncover. This analysis confirmed the results in the Westinghouse analysis documented in Reference 2. Therefore, for the main feedline break inside of containment, e uxiliary feedwater flow value of 200 gpm is acceptable. An auxiliary feedwater flow of 200 gpm is also adequate for a main feedline break outside containment because during such an event an inside containment check valve would close, isolating the steam generator and auxiliary feedwater flow from the break location. Isolating the steam generator from the break would preclude the main feedline break from effecting the Reactor Coolant System. At worst, the event would be the equivalent of a loss of normal feedwater event which only requires 200 gpm of auxiliary feedwater flow. ~

-~ Enclosuro 1 August 29,1997 Paps 3 Because of the discrepancies described above, the following corrections have been made to Attachment 4 of Reference 1, as shown in Enclosure 2:

1. On Page 4-11 the statement that the reactor trip occurs at 20 seconds due to low steam main pressure has been changed to state that the trip occurs at 10 seconds due to high containment pressure. In addition, the statement that the analysis allows for 400 gpm of auxiliary feedwater flow has been changed to state that the analysis allows for 200 gpm of auxiliary feedwater flow.

2. The main feedline break assumptions shown in Table 2-1 on page 4-13 have been changed to show the reactor trip occurs at 10 seconds and that a 200 gpm auxiliary

^ feedwater flow is required.

3. Figure 2-1 on page 4-15, which shows steam generator level versus time for a main feedline break, has been replaced with a figure from Reference 2 which shows steam generator level versus time for a main feedline break where the reactor trip is assumed to occur at 10 seconds with 200 gpm auxiliary feedwater flow.

Other Discrepancies In addition to the discrepancies with respect to the inain feedline break analysis other discrepancies with respect to the auxiliary feedwater design basis were also identified in Attachment 4 to Reference 1. As a result of those discrepancies, the following changes have been made to Attachment 4 of Reference 1, as shown in Enc':.sure 2:

1. The discussion of design conaitions on page 4-3 of Attachment 4 has been modified to correctly reflect which transients imposed safety-related performance requirements during the original design of the system and as described in the 4

Prairie Island Final Safety Analysis Report.

2. A discussion from Section 6.6.3 of the Prairie Island Final Safety Analysis Report, which discusses the auxiliary feedwater flow required to prevent thermal cycling of the tube sheet, has been added to the design conditions discussion on page 4-3 of. This information has been incorporated because it is pertinent auxiliary feedwater design information that should have been included in of Reference 1 when it was originally submitted.

3. A note has been added to Table 1B-1 on page 4-8 of Attachment 4 to clarify that the feedline rupture transient was not included in the Prairie Island Final Safety Analysis Report.

4. A sentence at the top of page 4-10 of Attachment 4 states that analyses have been performed for the limiting transients which define the auxiliary feedwater performance requirements. The limiting transients listed include loss of main feedwater, rupture of a main feedwater pipe and rupture of a main steam pipe inside

~ Enclosura 1 August 29,1997 P:ga 4 containment. There is then a sentence that states that these analysos have been-provided for review and have been approved in the applicant's FSAR. Because we are not able to determine definitively what was submitted for review, and because the main feedline break transient was not discussed in the Prairie Island Final Safety Analysis Report, the subject sentence has been deleted. References i

1. NSP to NRC Letter, dated February 4,1981," Auxiliary Feedwater System Information."

2. Westinghouse Letter PlW-P-540, dated October 1,1969, " Auxiliary Feedwater System Sizing."

3. Westinghouse Letter PlW-P-519, dated September 5,1969, " Auxiliary Feedwater System Criteria."

4. Pioneer Service and Engineering Letter PIP-N-353, dated October 6,1969,

" Auxiliary Feedwater System - Feedwater Line Rupture Accident."

ENCLOSURE 2 to Letter Dated February 4,1981 ' " Basis for Auxiliary Feedwater System Flow Requirements" With Corrections Noted 9 t

Director of NRR Tobruary 4,1981 - l d PRAIRIE ISLAND NUCLEAR GENERATING PLANT i Basis for Auxiliarv Veedveter System Flow Requirements (Response to Enclosure 2 of NRC letter of 10/16/79)- i j 1 4 Prepared by Westinghouse Electric Corporation. for Northern States Power Company -l 4-1

L, l Question ~1' i =! a.- -Identify the plant transient and accident conditions considered in [ establishing AFWS flow requirements, including the following events. I

1) Loss of Main Feed (LMFW)-

1 4 i 2 LMFW w/ loss of offsite AC power - 3 LMFW w/ loss of onsite and offsite!AC power i '4 Plant cooldown - 5 Turbine trip with and without bypass p 4 i 6 Main-steam isolation valve closure I i 7 Main feed line break i L 8 ; Main steam line break- ) 9 Small. break LOCA l

10) Other transient or accident conditions not listed above.

' b.- Describe the plant protection acceotance criteria and corresponding- ] technical bases-used for each initiating event identified above. The acceptance criteria should address plant limits such as:- F

1) Maximum RCS pressure (PORV or safety valve actuation)

2) Fuel temperature or damage limits (DNB, PCT, maximum fuel-n central temperature) 3 RCS cooling rate limit to avoid excessive coolant shrinkage Minimum steam generator level to assure sufficient steam gen-l-

4 erator heat transfer surface to remove decay heat and/or.. cool down the primary system. Response to 1.a l The Auxiliary Feedwater System serves as a backup system for supplying feedwater to the secondary side of the steam generators at times when - the feedwater system is not.available, thereby maintaining the heat sink L capabilities of the steam generator. As an Engineered Sareguards Sys-l . tem, the Auxiliary Feedwater System is directly relied upon to prevent core damage and system overpressurization in the event of transients suchLas a loss cf normal feedwater or a secondary system pipe rupture, _ and. to provide a means for plant cooldown following any plant transient. Following a reactor. trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated steam either to the condensers through the steam dump or'to the atmoschere througn the steam Steam generator safety valves or tne power-operated relief valves.

generator water inventory must be maintained at a level sufficient to ensure adequate -heat transfer and continuation of the decay heat removal l-The water level is maintained under these circumstances by tne process.

Auxiliary Feedwater System which delivers an emergency water supply to [

~

-the steam generators.- The-Auxiliary Feedwater System must be capable;of l functioning for extended perieds, allowing: time either to restore normal L feedwater flow or to ' proceed with an orderly cooldown of the-plant to i 4 i ). 4-2 i. ev = amemamamm se e e. e w. r +- 3 = v - = ' '-'== -

the reactor coolant temoerature where the Residual Heat Removal System can assume the burden of decay heat removal. The Auxiliary Fetdwater System flow and the emergency watar sucoly capacity must be sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown. The Auxiliary Feedwater System can also be-used to maintain the steam generator water levels above the tuces fol-lowing a LOCA. In the latter function, the water head in the steam generators serves-as -a barrier to prevent leakage of fission products from the Reactor Coolant Systam into the secondary plant. DESIGN CONDITICNS The reactor plant conditions which impose nf;n r;ht:d performance requirements on the design of-the Auxiliary Feedwater System are as follows for the Prairie Island phnts. Loss of Main Feedwater Transient Loss of main feedwater with offsite power availabi Station blackout (i.e., loss of main feedwater wi+hout offsite power available) Secondary System Pipe Ruptures Feedlinerupturek Steamline ruptur Loss of all AC Power @ Loss of Coolant Accident (LOCA INSERT 1 Cooldown Loss of Main Feedwater Transients The design loss of main feedwater transients are those caused by: Interruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system Loss of offsite power or blackout with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a' rapid reduction in steam generator water levels which results in a reactor trip, a tur-bine trip, and auxiliary feedwater actuation by the protection system logic. Folicwing reactor trip frem hign power, the power quickly falls to decay heat levels. The water levels continue to decrease, progres-sively uncovering the steam generator tubes as decay. heat is transferred and discharged in the form of steam either through the steam duma valves to the condenser or through the steam generator safety or power-operated-relief valves to the atmosphere. The reactor coolant temperature

• Impose safety related performance requirements.

u u-4-3

l l INSERT 1 The Auxiliary Feedwater System provides three essential functions durin0 abnormal conditions: a. prevents thermal cycle of the steam generator tube sheet upon loss of main feedwater pump; b. removes residual heat from the reactor coolant system until the RCS temperature drops below 300-350'F and the RHR system is capable of providing the necessary heat sink; c. maintains a head of water in the steam generator following a loss of coolant accident. Tne feedwater flow rate required to prevent thermal cycling of the tube sheet and for removing residual heat is the same and is about 160 gpm per Unit (or 80 gpm per steam generator). A 200-gpm flow is therefore sufficient to fulfill all the three functions stated above. However, since the Auxiliary Feedwater System is a safety features system, an additional 200-gpm pump is provided as a backup for the first for each Unit. In other 2-toop plants (R.E. Ginna, Point Beach, etc.) prior to Prairie Island, the inventory of secondary side water in the steam generator versus secondary sic'e water level was such that the pump size required to provent thermal cycling was larger than the pump size required for removing the residual heat in the core (almost twice as big). As a result of this difference in capacity requirement between the two functions, these plants used the turbine driven pump to meet the larger capacity since it was not a safeguards requirement and therefore did not require redundancy. Smaller motor driven pumps are used for the safeguards requirements of removing fission product decay heat so as not to unnecessarily increase the diesel-generator size. The recent designs use a large steam generator with a different dimensional configuration. This reduces the pump size required to prevent thermal cycling, such duty being about the same as for removing residual heat, namely,160 gpm per unit or 80 gpm per steam generator. (Only one steam generator is required to remove all decay heat in the core.)

'Pi . increases as' thel residual heat-in excess of that dissipated through the 1 stern generators _ is absorbed. With increased temperature, the volumetof- ~ reactor coolant expands:and begins filling the pressurizer. Without the addition of sufficient auxiliary.feedwater, further expansion will - result in water being discharged througn the pressurizer _ safety and. relieffvalves. If the.temoerature rise and the resulting volumetric expansion of the primary coolant are permitted to _ continue, then (1) pressurizer. safety valve capacities may be exceeded causing over-stem and/or (2) the continuing pressuritation of _the-Reactor Coolant Sy'ystem may result in bulk boiling loss of fluid from the primary coolant s -in the Reactor Coolant: System and eventually in core uncovering, loss of natural circulation, and core damage. If such a situation were ever to occur, the Emergency Core Cooling Sjstem would be ineffectual because the primary cociant system pressure exceeds the shutoff head of the safety -injection pumps, the nitrogen over-pressure. in the ' accumulator tanks, and the design pressure of the Residual Heat Removal Loop. Hence, the timely introduction of sufficient auxiliary feeddater is necessary-to arrest the decrease in the steam generator water levels, to reverse the rise in reactor coolant temperature, to prRvent the pres-surizer from filling to a water solid condition, and eventually to establish stable hot standby conditions. Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satis-factorily corrected. The blackout transient differs frc'm a simole loss of main feedwater in that emergency power sources must be relied upon.toEcperate vital equip-ment. The loss of power to the electric driven condenser circulating water pumps results in a loss of condenser vacuum and condenser duma valves. Hence, steam formed by decay heat is. relieved through the steam - generator safety valves or the power-operated relief valves. The calcu-lated transient would be similar for 'both _ the loss of main-feedwater and the blackout, except that reactor coolant pume heat input is not a consideration in the blackout transient following loss of power to the 1 reactor coolant pumo bus. The station blackout transient serves as the basis for the minimum ~ flow required for the snallest caoacity single auxiliary feedwater pumo for the Prairie Island plants. The pump is sized so that any single pump will provide sufficient flow against the steam generator safety valve set pressure (with 3% accumulation) to_ prevent water relief from the pressurizer. Secondary System pice Ruotures The feedwater line rupture accident not only rcsults in the loss of -feedwater flow to the steam generators but also results in the comclete bicwdown of one steam. generator within a short time if the rupture should occur downstream of the last nonreturn valve in the main or auxiliary feedwater piping to an individual steam generator. Another significant-result of a feedline rupture may be the spilling of aux-iliary feedwater out the break as a consequence of the fact that the 4-4 --h__.__________.m._

auxiliary fetteater branch line may be connected to the main feefdater line i_n the region of the postulated break. Such situations can result in the spilling of a disproportionately large fraction of the total auxiliary feetsater flow because the system preferentially pumps water to the lown.st pressure region in the f aulted loop rather than to the effecti.t steam generator whien is at a relatively hign pressure The system design must allow for terminating, limiting, or minimizin that fractie of auxiliary feedwater flow which is delivered to a f au ted loop or spilled-through a break in order to ensure that sufficient flow will be delivered to the remaining effective steam generator. The concerns are similar for the main feedwater line rupture as those i:xplained for the loss of main feedwater transients. Main steamline rupture accident conditions are characterized initially by plant cooldown and, for breaks inside conthinment, by increasing containment pressure and temperature. Auxiliary feedwater is not needed during the early phase of the transient but flow to the f aulted loop will contribute to the release of mass and energy to containment. Thus, steamline rupture conditions establish the upper limit en auxiliary feedwater flow delivered to a f aulted loop. Eventually, however, the Reactor Coolant System will heat up again and auxiliary feedwater flow will be required to be deliveref' to the unf aulted loop, but at somewhat lower rates than for the loss of feedwater transients described pre-viously. Provisions must be made in the design of the Auxiliary Feed-water System to allow limitation, control, or termination of the auxil-iary feedwater flow to the f aulted loop as necessary in order to prevent containment overpressurization folicwing a steamline break inside con-tainment, and to ensure the minimum ficw to the remaining unf aulted i loops. Loss of All AC Power The loss of all AC power is postulated as resulting frem accident con-ditions wherein not only onsite and offsite AC power is lost but also AC emergency power is lost as an assumed coc:non mode failure. Sattery-power for operation of protection circuits is assumed available. The impact on the Auxiliary Feedwater System is the necessity for providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown until AC power is restored. Loss-of-Coolant Accident (LOCA1 The loss of coolant accidents do not imoose on the auxiliary feedwater system any flow retuirements in addition to those required by tne other accidents addressed in this response. The following description of *.he small LOCA is provided here for the sake of completeness to exclain the role of the auxiliary feedwater system in this transient. Small LOCA's are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume. The principal con-tribution from the Auxiliary Feedwater System following such small LOCAs is basically the same as the system's function during hot shutdown or 4-5

. following spurious safety injection signal whien trips the' reactor. Maintaining La water level-inventory in the secondary side of the steam generators provides a heat sink for removing decay heat and establishes-the capability for providing a buoyancy head for natural circulation. The-auxiliary. feedwater system may. be utill:ed to assist in a system cooldown and deoressurization following a small LOCA while bringing the reactor to a cold shutdown condition. Cooldown' The cooldown function performed by-the Auxiliary Feedwater System is a partial one since the reactor coolant. system is reduced from normal zero -load temperatures to a hot leg temperature of accroximately 3500F.- The latter is the maximum temoerature recommended for placing the Resi-dual: Heat Removal System (RHRS) into service. The RHR system completes the cooldown to cold shuttiown conditions, cooldown may be required following expected transients,.following an - accident such as a main feedline break, or during a normal cooldown prior to refueling or performing reactor plant maintenance. If the reactor is-tripped following extended operation at rated power level. the AFWS is capable.of delivering sufficient AFW to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while main-taining the stemn. generator (SG) water level. Following transients or accidents, the recommended cooldown ratt is consistent with excected needs and at the same time does not impose additional requirements on-the capacities of the auxiliary feedwater pumps, considering a single f ailure. In any event, the process consists of being able to dissipate 4 i plant sensible heat in addition to the decay heat produced by the reac-tor core. 4-6

Resoonse to 1.b Table 13-1 sumnarizes the criteria which are the ger.eral design bases for each event, discussed in the response to Question 1.a, above. Specific assumptions used in the analyses to verify that the design bases are met are discussed in response to Question 2. - The primary function of the Auxiliary Feedwater System is to provide sufficient heat removal capability for-heatup accidents following reac-ter trip to remove the decay heat generated by the core and-prevent ' system overpressurization. Other plant protection systems are designed to meet short term or pre-trip fuel f ailure critaria. The effects of excessive coolant shrinkage-are bounded by the analysis of the rupture of a main steam pipe transient. The maximum flow requirements deter-r.ined by other bases are incorporated into this analysis, resulting in no additional flow requirements. I e e B e 9 0

• +

6 e ump 4-7

' TABLE 18-1 Criteria for Auxiliary feedwater System Design Basis Conditions Condition Additional Design or. -Transient Classification

• Criterla*

Criteria Loss of Main Feedwater Condition 11 Peak RCS pressure not to exceed design pressure. No consequential fuel failures l Station Blackout Condition !! (same as LHFW) Pressurizer does not fill,with ~ 1 single motor driven aux. feed pump feeding 1 SG, Feedline Rupture Condition IV 10 CFR 100 dose Ilmits. Core does not uncover Containment design pressure a r- /= not exceeded Steamline Rupture Condition IV 10 CFR 100 dose limits-Containment design pressure not exceeded. i Loss of all A/C Power N/A Note 1 - Same as blackout assuming turbine driven pump Loss of-Coolant Condition 111 10 CFR 100 dose 11mits 10 CFR 50 PCT Ilmits Condition IV. 10 CFR 100 dose limits 10 CFR 50 PCT limits 8 10C F/hr Cooldown N/A 5470F to 3500F' ANSI N18.2 (This information provided for those transients performed in the FSAR).

• Ref:

Although this transient establishes the basis for AFW pump powered by a diverse power source, this is not Note 1 evaluated relative to typical criteria since multiple failures must be assianed to postulate lhls 'translent. " Not included in Prairie Island Final Safety Analysis Report =. - g

Question 9 Describe the analyses and assumptions and corresponding technical justi-fication used with eacn plant conditicn considered in 1.a above includ-ing: a. Maximum reactor power (including instrument error allowance) at the , time of the initiating transient or accident.

b. ' Time delay from initiating event to reactor trip.

c. Plant parameter (s) which initiates AFWS flow and time delay between initiating event and introduction af AFWS flow into steam genera-tor (s). d. Miniinum stes. generator water level when initiating event occurs.

e.. Initial steam generator water inventory and depletion rate before and after AFWS flow comences -- identify reactor decay heat rate used.

f. Maximum pressure at which steam is released from steam generator (s) and against which the AF4 pump must develop sufficient head. g. Minimum number of steam generators that must receive AFW flow; e.g.,- 1 out of 27 2 out of 47 h. RC flow condition -- continued: operation of RC pumps or natural circul ation,

i. Maximum AFW inlet temperature.

s ,1. Folicwing a postulated steam or feed line break, time delay assumed to isolate break and direct AF4 flow to intact steam generator (s). AFW pump ficw capacity allowance to accomodate the time delay and maintain minimum steam generator water level. Also identify credit taken for primary system heat removal due to blowdewn. k. Volume and maximu:n temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line. 1. Operating condition of steam generator ner.nal blowdown folicwing initiating event. Primary and secondary system water and metal sensible heat used for m. cooldown and AF4 flew sizing. Time at hot standby and time to cooldown RCS to RHR system cut in n. temperature to size AF4 water source inventory. 4-9 .l 1 ..n

• v Resconse to 2 Analyses have been performed for the limiting transients wnich define -

the AFWS performance requirements. Th::: = d y::: 5;: t;= ;r;;id;d f;r r;;i;w =d hr;: t;:n ;;;r:v; in th; A;;li; =t', ~20 Specif1- -cally, they include: Loss of _ Main Feedwater (Station Blackcut) Rupture of a Main Feedwater-Pipe Rupture of a Main Steam Pipe Inside Containment In addition to the above analyses, calculations have been performed specifically for the Prairie Island plants to determine the plant cool-

1. wn flow (storage capacity) requirements. The Loss of All AC Power is evaluated via a comparison to the transient results of a Blackout, assuming an available auxiliary pump having a diverse (non-AC) power supply. The LOCA analysis, as discussed in response 1.b, incorporates the system flows requirements 45 defined by other transients, and there-fore is not performed for the purpose of specifying AFWS ficw require-ments. Each of the aalyses listed above are explained in further detail in the fo1' awing sections of this response.

Loss of Main Feedwater (Slackouti A loss of f eedwater, assuming a loss of power to the reactor coolant pumps, was performed in FSAR Section 14.1.10 for the purpose of showing _ that for a station blackout transient, a single motor driven auxiliary feedwter pump delivering flow to one steam generator does not result in fill. J the pressurizer. Furthermore, the peak RCS pressure remains

elow the criterion for Condition II transients and no fuel failures occur (refer to Table 1E-1). Table 2-1 sumarizes the assumptions used in this analysis. The transient analysis begins at the time of reactor trip. This can be done because the trip occurs on a steam generator level signal, hence the core power, temperatures and steam generator level at time of re

ctor trip do not depend on the event sequence prior to trip. Although the time frcm the loss of feedwater until the reactor trip occurs cannot be determined frem this analysis, this delay is expected to be 20-30 seconds. The analysis assumes that the plant is initially operating at-102% (calorimetric error) of the Engineered Safeguards design (E50) rating shovn on the table, a very conservative assumption in definiitg decay hat end stored energy in the RCS. The reactor is assumed to be tripped on low-law steam generator level.

Steam generator level at the time of reactor trip was assumed to be 0% NRS for additional conservatism; to that, allowance for level uncertainty was also accounted for. The FSAR shows that there is a considerable margin with respect to filling the pressurizer. This. analysis establishes the escacity of the smallest single pump and also establishes train association of equipment so that this analysis remains valid assuming the most 1_imiting single failure. 4-10 h

Ruoture of Main Feedwater piee high c n nment The double ended ructure of a main feedwat e pipe downstream of the main feedwater line enect valve is not analyzed in the FSAR. Tacle 2-1 sum-marizes the assumotions used in the analyses performed for Prairie Island (Ref erence 1). Reactor trip is assamed to occur as a result of a 10 safety injection signal basec on ic: ::: # - m pressure in either loop at seconds into the transient. This conservative assumption maxi-mizes tne stored heat prior to reactor trip and minimizes the acility of the steem generator to remove heat fran the RCS following reactor trip due to a conservatively small total steam generator inventory. As in the loss of nomal feedwater analysis, the initial power rating was 200 assumed to be 102'.' of the ESD rating. The analysis allows for <Me gym auxiliary feedwater celivered to the intact loop within 10 minutes of the reactor trip (10 minutes fcr operator action to reroute ficw catns and to start the auxiliary feecdter pumps). The criteria listec in Table 13-1 are met. This analysis may establish the capacity of single pumps, establis' es n requirements f or layout to preclude indefinite loss of auxiliary feed-water to the postulated break, and establishes train association requirements f cr equioment so that the AFWS can deliver the minimum flow required in 10 minutes folicwing operator actions assuming tne worst single failure. Primary system heat removal due to blowdown is included in our analytical code model and is correctly simulated during the feec-line ructure analysis. Ruoture of a Main Steam pice ?nside Containment Because the steamline break transient is a cooldcwn, the AFWS is not needed to remove heat in the short term Furthermore, addition of excessive auxiliary f eedwater to the f aultt:d steam generator will affect the peak containment pressure following a steamline break inside con-tainment. This transient is performed for several break sizes. Aux-liiary feedwater is assumed to be initiated at the time of the break, independent of system actuation signals to provide the most conservative analysis with respect to containment pressure. Taole 2-1 sumarizes the assumptions used in this analysis. The criteria stated in Table 13-1 are met. This transient estaolishes auxiliarv feecwater flow rate to a single f aulted steam generator assuming one pur'p operational and estaclisnes layout requirements so tnat the flow requirements may be met considering the worst single f ailure. Primary system heat removal due to blowdown is included in our analytical coce model and is correctly simulateo during the steamline rupture analysis. Reference 1: "idP/ iRP Auxiliary Feecwater System Sizing." PIW-P-540/KW-P-571. Octocer, 1969. 4-11 l

~- n .(; Plant Cooldown Maximum and minimum flew recuirements from the previously discussed -. transients meet the flow requirements of plant cooldown. This opera-tion, however, defines the basis for tankage size, based on the recuired cooldown duration,~ maximum decay heat -input and maximum stored heet in the system. - As previously discussed in response 1A, the auxiliary feed. water zsystem partially cools the systes to the point where the RHRS may complete. the cooldewn, i.e., 350cF in the RCS. Table 2-1 shcws the-assumptions used to deter. sine the cooldown heat capacity of the auxil-iary feedwater system.. -The cooldown is assumed to comnence at the maximum rated power, and maximum t-ip delays and decay heat source terms are assumed when the reactor is -tripped. Primary metal, primary water, secondary system metal and secondary 'dS. See Table 2-2 for the items constituting the -system watet are ail in be removed by the AF sensible heat stored in the NSSS. This operation is analyzed to establish minimum tank size requirements for auxiliary feedwater fluid source which are normally aligned. ? ~. 9 + D s 4-12 k

+ s h h. ' 1A8tE 2-1 Samunary of Assumptions Maed in MW5 Sestge Verification Analyses. h la 5teamilee treak' i - 1ess of feeduator-(i etten blackout) Coeldnesa 8eale feedilee Greak kant ainmenil ll r t 1ranMe_nt a. Stas reacter remer 1025 of f58 rat . 1683 MWt 1025 of ESS rett 55 et 1658 left for first les sec. If thereafter - no (1925 ei 1722 last (1021 of 1722 Mut k b. ilme delay f rom - 1.5 sec 2 sec sec 1.5 sec event te Aa trip 2 sec addftlanal for red movement c. MW5 acteation sig-le-le M level N/A Sperater actlen/ Assamed lasediately," - mal / time delar for I minista 10 minutes af ter e sec (se adelay) free j Mus flow reacter trip - ISE design centalement pressure and lies nemleal steam flew signel ' d. M water level at 36.9 ft. N/A N/A-tevel'at het aere gamer eleiss I feet. 36.s75 ft. time of reacter trip 65 Ist Span (24.671 tR 5 pan). i

e. ' Initial SG levantary 61,000 16m/56 go 058 ths/58 92,500 lha/SS 154,000 lha/58

~ 0$l3.74F Rate of change betere see figure 2-2 N/A see flysre 2-1 N/* e 4 l & af ter Mit$ actesatten 7 decay heat see figure 2-3 N/A see figure 2-3 see figure 2-3 f. MW pump desige - see cooldsun 1833 pela see coeldsun see coeldsun pressure g. Miniman # of SGs I of 2 N/A 1 of 2 - 1 of 2 which must receive - j Mit flou h. RC pump status Tripped e reacter trip. Tripped Tripped 0 reacter trip All aperating. l. Maslaun MW -100eF leoer. gegeF leser temperature sinne assumed / N/A/ 18 min./ le ele./ J. Operator actlan/ ll/A II/A Vek Ves prSmary teat removal ese to Iloudseen 3 3 3 3 les it /404.6 Ste/tta 100 ft /414.4 Sta/the 100 f t /4le.4 Stu/he - 100 ft / k. If W purge volume / temp. ' 432.8er 1. storeal blundown mene assessed, none assessed none assemed none assumed m. Sensible heat see coeldense - Table 2-2 see coeldsun see coeldsun < n. Ylee at standby / time see coeldome 2 brf6 br see coeldous see coeldsun to coutdown to kHit e. MW flow rate 200 GPfl - constant - variable GMt at ten ele- - 20s spel (min, ressesireamt) wies fram reacter trip 59 %A

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i. . TABLE 2-2 9 Summary cf-Sensible Heat Scurces Primary 1Watar Sourcas-(initially at ratec power tamcerature cnc inventery). - ACS fluid. - Pressurizer fluid (liquid. anc vapor) 1 Primary Metal Scurcas (initially at rated power tamcerature) - Reacter ccolant piping, pumps and reactor-vassal - Pressurizar - Stasn generatar tuce metal and tuce sheet - Stesa generater metal belew-tuce snett - Reactor vessel-internals Secondary Water Scurces (initially at ratad pcwer tam;erature anc inventery) - Steam generstar fluid'(liquid and vaccr) - Main f etcwater purge fluid between s:can generater and AF45 piping. Secondary Metal Sources (initially at rated power tamcerature) - All staam generater metal accve tube-sheet, exclucing tuces. ~ e (_ \\ 4-14 I

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ENCLOSURE 3 Westinghouse Letter PlW-P-540 Dated October 1,1969 " Auxiliary Feedwater System Sizing"

{ Westinghouse Electric Corporation. Power Systems msnmum eum PimtughPtnryma 15?30 j O g r 1, 1969 i PIW-P-5I.0 I l s l Hr. D. !!. Leppke, Manager FW-P-571 Nuclear & Advanced Energy Systems Pioneer Service & Engineering Cr any CE-CPA-975 2 Nosth Riverside Plaza Chicago, Illinoin 00606 e' ,iil Dear Mr. Leppke PRAIRIE ISLAND hTCLEAR CENERATING PLANT E-6197 ![. KEWAUNEE NUCLEAR POWER PLAh7 Auxiliary Feedwater System Sizing 'i j; Further to our recent telephone conversation with Hessrs. Hollinphaus and lyer of Pioneer Services, we have now conducted analyses of the } feedwater main rupture accident using high containment pressure to I 1 Initiate reactor trip. Detailed calculations using the COCO code indicate that the high containment pressure setpoint will Se reached in eight seconds. Our calculations used a conservative time of ten seconds for the pressure setpoint to be reached, with a further two second delay before the initiation of control rod movemer.t. Three cases were considered corresponding to flows of 200 gpm, 250 gpm and 300 gpm i delivered to We intact steam generator ten minutes af ter the initiation of the accident. Our conclusion is that a flow of 200 gpm will be sufficient to preclude over pressurization of che primary systen. The basis of the calculations was the same as those detailed in our letter PIW-P-519/KW-P-546 of September 6. The results are shown in the attached figures. 4 Do NOT Destroy PROJECT Fil.E N N'2'7 Date M-0 FDe #_8 #' #O

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~. :; 9 O 1 Nr. D. H. Leppke October 1, 1969 ) .l If Pioneer elects to proceed on the basis of their proposed system of two 200 gpm pumps, without adding any back-up capacity, the following j points should be borne in mind: l 1. Although Westinghouse has confidence in the defensibility of the current calculations, arror' margins are small, and will receive much attention during licensinC. 1 2. The system provides no diversity of energy sources. This in itself does not preclude the system, but may attract AEC criticism. 3. It may be necessary co add additional check valves in order to satisfy failure criteria. I 4 With only two pumps, the likelihood of drying out a steam generator j on a normal loss of feed (blackout) is greatly increased. 5. Maintenance of auxiliary feed pump will be more difficult. In particular, outage time will probably be limited by the technical specifications. We will naturally be pleased to answer any further questions which you may have. I Very truly yours. k A. C. Hall Control and Protection Analysis I APPROVED: (2/yto 4 /)l 844 /,vA.B. Cutter, Manager Prairie Island & Kewaunee Projects cc: D. M. Leppke, 10L V. H. Simon, IL C. D. Albrecht, 2L F. M. Sovis, IL i L. D. Bird, it l R. C. Straub, IL j R. J. Jensen, 2L l C. A. Lins, IL 8 + nf-wy. 5 ww..e _n ~ 3, ,r. w r, p mv n r., - ir m e a y n w =rtn r s. ; ; v .,,o r,

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ENCLOSURE 4 Westinghouse Letter PlW P 519 Dated September 5,1969 " Auxillary Feedwater System Criteria"

p g.x c y ~ s. +.c m ; my. .. - ~ ~ 3 2 3. 3 N / -ul Westingh0use Electric C0rp0ratl0n P0wer Systems resnwtan asm PittstupWna tt?30 September 5, 1969 PIW-P-519 KW-P-546 y AtW Hr. D. M. Leppke, Manager ' ( un. Nuclear & Advanced Energy Systems 8 gjyd // P ' Pionear Service & Engineering Company j' 2 North Riverside Plaza Chicago, Illinois 60606 PRAIRIE ISLAND NUCLEAR GENERATING PLANT E-6197 KEWAUNEE NUCLEAR POWER PLANT Auxiliary Feedwater System Criteria Dear Mr. Leppket At the request of Pioneer Services, we have examined in detail the re-quirements for Auxiliary Feedwater System performance in the event of a hypothetical rupture of a feedwater line. The conclusion of the analysis is that any installed system must meet, in spite of any assumed single active failure, one of the following conditions: 1. 400 gpm delivered to the intact steam generator within ten (10) minutes of reactor trip. 2. 300 gpm delivered to the intact steam generator within one (1) minute of reactor trip. 3, 200 gpm delivered to the intact steam generator within one (1) minute of reactor trip and a further 200 gpm within ten (10) minutes of reactor trip. (Note that the above capacities represent actual pump delivery. Pioneer should add any margin, to these capacities, that they believe is prudent.) This analysis of the auxiliary feed system was based on preventing over-pressurization of the primary system in the event of the hypothetical main feedwater line break. Flow capacity was evaluated based on stored energy and decay heat without attempting to determine the minimum possible size j for the auxiliary feed pumps, to retain additional conservatism, and pre-j vent discharge of coolant from the pressurizer for the more credible case 1 of loss of main feedwater without a feedwater line rupture. ] The two loop auxiliary feed system recommended by Westinghouse consists of two (2) 200 gpm motor driven pumps and one 400 gpm turbo-pump. This system ' ) provides diversity of energy sources, allows adequate time for operator action in rerouting flow paths and has been favorably accepted by the AEC for the Ginna Plant. (Recall also, that in their Safety Evaluation, the AEC m 'N TY @Me E E' UYE TYTf NY MY F "E# E Y A ,wYY I Mi?

f f.V?,.090 g& 4h %f. yJ),; %iP64 ? H lO $f f ~- } WMfb%'f. Mr. D. M. Leppke September 5, 1969 1] PIW-P-519 P KW-P-546 cb !? had given Northern States Poder and Wisconsin Public Service credit for the 'J diversity inherent in the two motor driven - one steam driven pug system.) The turbine driven ptsap provides redundancy plus additional flow capacity for the credible feedwater accident. This added capacity was determined by the feedwater line rupture analysis to be necessary for recovery action following this accident. Our analysis indicated that with the recommended system, the operator has about ten (10) minutes to reroute flow paths to prevent over-pressurization of the primary system. In addition, for the more credible loss of main feedvater flow without a feedvater line break, the reconomended system will insure that no water is relieved from the .1 4 ') pressuriser. 1 We recognize that other possible auxiliary feedwater systems could meet the i same criteria with less installed flow capacity. However, decreasing the f l installed flow capacity has the direct consequence of reducing the time for operator intervention. In our judgment anything less than ten (10) l minutes for operator action is unacceptable and would require automatic action to reroute flow paths. Such an autoinatic schene should meet the single active failure criteria as being part of the protection system, i.e. the system would require re-dundancy of sensors, logic channels and valving. Due to the variety of failures to be considered (pumps, valves, flow indicators, etc.) any such scheme would be very complex to design, will require periodic testing and maintenance, and is difficult to justify for the protection against a very rare event such as a main feedwater break. Complex automatic systems to protect against rare accidents of ten jespordize the response to the more credible failures. If Pioneer elects to take an alternate approach to the Westinghouse recom-mended system, their detailed system should be submitted to Westinghouse for review. A sununary of our eniculations is presented in the appendix for your infor-mation. We will be pleased to answer any questions you may have regarding these calculations. Very truly yours, 0<v Attachment A. C. Hall Control & Protection Analysis cc: D. M. Leppke 10L J. S. Moore IL a V. H. Simon IL J. M. Geets it C. D. Albrecht 2L A. C. Hall IL F. M. Sovis 1L H. J. von Hollen 1L L. C. Bird it V. P. Buscemi 1L A. B Cutter, Manager R. C. Straub IL H. J. Bruschi 1L prairie Island & Kevaunee Projects R. J. Jensen 2L D. C. Marburger IL C. A. Lins 1L A. R. Collier IL D. C. Beatty 6L F. D. O'Brien IL A. B Cutter 4L

.M ([fMh gf @r) $h NMYdd5 g./.*dhk N.4bfft. b M PENDIX 2 ( J TWO LOOP FLANT AtTIILIARY FEEDWATER SYSTD( SIZING b INTRODUCTION i The prinicpal problem in conducting the ana.1yses involves evaluation of heat i removal credit for the steam generator in which the rupture occurs. This quantity is critically dependent on the amount of liquid carried over by the steam phase. (Water which is ejected in the liquis! phase, is clearly not available to cool the system by evaporation.) The initial liquid mass inventory is 92,500 pounds per steam generator. This is equivalent, purely on the basis of heat removal, to 40 minutes flow at 200 gym. s The transient is also sensitive to the time of reactor trip. The steam generator associated with the rupture depressurires fairly rapidly, and this produces a tendency for the level to rise, even though the water inventory is dropping. Until trip occurs, the turbine load demand remains constant, and so the good steam generator stems flow rises. This again produces a rise in level. It is, therefore, not possible to take credit for trip associated with low level. There is a small section of feed main outside the containment but downstream of the check valve. If the rupture occurs at this point, no credit can ba taken for high containment pressure. Trip will occur as a result of a safety injection signal based on low steam main pressure on either loop. This trip is lead / lag compensatzd and has 2/3 logic. The time of this trip is dependent on the lead / lag ratio and on the carryover as s ump tions. Reasonable variations in basic 'asstreptions, sut.h as carryover, produce large changen in the transient. It is clearly essential that the system should be adequate to prevent a public hazard, and so some conservatism must be included in our analyses. Analyses were conducted using a newly developed long term multi-loop digital code. The code has a detailed natural circulation calculation, and permits realistic representation of the primary to secondary heat transfer for very low steam generator water levels. Bases of the analysis. This section describes the " ground rules" adopted for the analyses, and discusses their nrigin. 4 g 1. Carryove r. As a result of the depressurization of the steam generator i associated with the break, there will be an initial rise in level thus tending to keep the feed ring submerged. Falling pressure vill cause the saturation temperature to fall. At the same time, the lack of feed-water will cause the downconer enthalpy to rise. These two effects in combination will result in early flashing in the downcomer. This will disrupt the recirculation process and produce an extremely large rise in 1s. val. At this stage the whole steam generator will in all probability be filled with a frothing mass.

s. APPMDIX Page 2 4 Under these circumstances it is extremely difficult to predict carryover. We arbitrary, but conservative, assusption has been made that the exit quality is qual to the mean quality in the steam generator, i.e. steams generator phases are homogenized. 2. Discherme flow. Flow discharge through the rupture is dependent on the length of pipe between the steam generator and the rupture. A break near, the steam generator would result in more break flow at a given time and hence less cooldown from this steam generator. However, this effect would be more than offset by the more rapid trip resulting from high containment pressure. A mervative estimate of flow was made, based on an outside containment break. 3. _Initisi steam generator level / initial pressurizer level. The transient is reasonably sensitive to both these levels. For simplicity they were assumed to be at their nominal values, and the necessary conservatism provided elsewhere. (Specifically in the flow and carryover asstamptions). 4. Trip time. Trip was assumed to result from a safety injection signal re-sulting from low steam main pressure. Asstmaing an initial steam generator pressure of just over 800 psi (typical for observed UA's af ter T ad-AVC j2stment) the steam main pressure of the ruptured steam generator will reac% 500 psi in 25 seconds. The maximum credible lead /las ratio of 6:1 would advance the trip to 15 seconds. However, separate investigations have indicated that this may be an unattainably high ratio. For this reason, and also in order to include some small margin for error, the scram time was assumed to be 20 seconds. 5. power level. All analyses were conducted for 102% power. (100% = 1722 MW,) tower power levels will result in higher steam generator pressures and hence larger trip delays. However, the core energy re-lease in this period will be correspondingly less. Also the steam gen-erator inventories will be higher, primary temperatures lower and pressurizer level lower. These effects will combine to inake the accident less severe at part powers. The above assumptions are felt to introduce a balanced, but not excessive conservatism. (Bearing in mind current uncertainties) However, no reason-able auxiliary feed system will stop the pressurizer from filling on the above basis. Since this is a hypothetical accident it was felt that safety valve discharge of liquid water should be acceptable. The prime criterion was therefore taken to be no unacceptable over-pressurization of the system. Cases Considered h e design of the auxiliary feed system is the responsibility of the architect-engineer. However, in order to size the system it is necessary to know some-thing of its characteristics. For the purpose of the study, three groups of runs were conducted as follows:

APPENDIX Fage 3 t i ]q y 1. It was assuesd that no flow was available until after operator action. This corresponds to the type of system employed by Rochester Cas and i Electric, if we have a f ailure of the motor driven pump associated with 3 the good steam generator. In this case the motor driven pump on the ruptured stems generator and the backup pump both drain to the break.. Flow to the good steam generator commences af ter the operator diagnoses the situation and carries out the necessary valving manipulations. The time delay for operator action was assumed to be ten (10) minutes. 2. It pas asstaned that all the auxiliary feedwater was available one minute af ter scram. This corresponds to a totally automatic system which diagnoses tbs break and automatically switches all flow to the good steam generator. 3. It was assumed that 200 gpm was availeble one (1) minute af ter scram, and a further flow ten (10) minutes after the scram. This corresponds to a system where automatic valving initially guarantees equal flow to both steam generator, and where the operator can divert further flow to the good stems, generator af ter diagnosing the break. The results of the runs are plotted in the attached figures. Description of Figures. Three figures are presented for each of the three types of sys tea. The first figure shows level in the intact steam generator as a function of time. This level in conjunction with the steam generator delta T (Primary to secondary) sets the effective heat transfer parameter. The second figure shows the normalized net power to the primary, i.e. the decay heat less heat removal. So long as this number remains positive the primary system stored energy continues to increase. pethirdfigureshowspressurizerlevel. the dotted line indicates the integrated Af ter the pressurizer fills at 1000 f t liquid release. The resumed solid line indicates that a str..a bubble has again been drawn in the pressurizer. o

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ENCLOSURE 5 Pioneer Service and Engineering Letter PIP N-3G3 Dated October 6,1969 " Auxiliary Feedwater System - Feedwater Line Rupture Accident"

_1,no .s. O COP.Y .;i73 ma" PioneerService&EnQine5rin Go.s a sqlvE ste10E PLA2A SUILDING I 3 Norr tw essvtse DIDE PLA2 A CHIC ADO. 4LLINDIS 409 06 s A '.'s i A t/ '. 'r October 6,1969 PIP-N-353 Hr. A. V. Dienhart i Assistant Vice President - Engineering Northern States Power Company l 414 Nicollet Mall Minneapolis, Minnesota 55401 Attention: Mr. D. E. Larsen

Dear Mr. Dienhart:

Prairie Island Nuclear Generating Plant - Units 1 & 2 i Project 21-6197 f Auxiliary Feedwater System - Feedvater Line Rupture Accident

Reference:

Westinghouse letter PIW-P-540 dated October 1, 1969 This letter is written to reconfirm the adequacy of the 200 gpm auxiliary feedwater ptusp for the Prairie Island Nuclear Generating Plant. As per the referenced West-inghouse letter, the 200 gpm pump is sufficient for the auxiliary feedvater flow. It may seem at firs t sight that this is marginal, however, we should bear the following points in mind which should also provido an answer to the questions raised by Westinghouse in their letter. 1. The 200 gpm flow rate meets the WORST possible condition and therefore does not require any additional margin. This worst condition includes the single j failure criterion, no auxiliary feedwater flow until operator action in aligning the flow to the proper steam generator (10 minutes), no reactor trip for 10 seconds after the accident. However, we do have additional margin as explained below. { I 2. The Westinghouse analysis used a 10 second reactor trip which is 2 seconds more than High Containment Pressure Safety Injection - Reactor Trip. This 2 seconds represents approximately 40 gpm flow in the pump stee required to perform its function. To elaborate further, as per analysis given in a pre-vious Westinghouse letter PIW-P-519 dated September 6,1969, a 20 seconds reactor trip with auxiliary feedwater flow commencing at 10 minutes after the accident requires a flow rate of 400 gpm. Under similar conditions with a reactor trip at 10 seconds as given in the recent analysis, the flow rate required is 200 gpm. Hence, a 8 seconds reactor trip (which is what the Containment Transient Analysis shows), with all other conditions remaining the same the flow required is only 160 gpm. Therefore, we do have a 40 gpm margin in the 200 gpm pump and there is no special need to increase this margin. L

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J.ms. r.. -;;,. 3p y,.3. n y. ,,,y . ~. .p.... g g I PionecrSonice &Engincoring 60. Mr. A. V. Dienbart October 6,1969 3. The effect of the Safety Injection on the Primary Coolant System has been neglected in the Westinghouse, analysis. This will further reduce the cooling requirements on the auxiliary feedwater system. 4 The system as proposed does reduce the diversity of energy sources. But again the turbine driven scheme is unique oniv to the auxiliary feedwater system. In other safeguards system requiring coolant to be pumped or injected the prime-movers are of only one kind (usually motor driven). This therefore should not prevent us from proceeding with the proposed scheme of two motor driven pumps. 5. The likelihood of drying out a steam generator m a normal loss of feed (black-out) is NOT GREATLY increased as stated otherwise in the Westinghorse letter. The reasoning behind this is the fact that in this situation the reactor trips essentially at zero second af ter the blackout, or 10 seconds earlier than in the Westinghouse analysis. In addition, there is no blowdown from ':he steam generator since there is no break in the feedline which enables us to pick up the additional water inventory of the broken steam generator. 6. Additional check valves in the feedwater line are not required to satisfy failure criteria. In the proposed scheme with a failed check valve (the single failure criterion) the reactor trip is delayed for 20 seconds (see P1W-P-510). However, both pumos are available to deliver 400 gpm at 10 minutes af ter the accident and is sufficient to meet the requirements. 7. The only one difficulty that we can see in the proposed scheme 4= -= N q maintenance requirements. Any maintenance on one pump will require that the other pump be operational and perhaps running during maintenance. Of course this situation is ne dif ferent than the case of maintenance on S.I. pumps and RHR pumps, in conclusion to recapitulate we state that: 1. Two 200 gpm pumps are sufficient for the auxiliary feedwater system to meet all its requirements. 2. Piping will be such that each pump will normally supply one steam generator ~ with provision through normally closed valves for auxiliary feedwater flow te be diverted to the other steam generator. It shoula be of further interest to you to learn that two other nuclear plants art following in this same directioa. However, ther.t are 4-loop plants. An early confrontation with the AEC-DRL on the proposed change should be helpful. S 2 l

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,, s p o,;. . e,,p,. ,; s c nc,",,N O, 9' r.. Pioneer Sorvicc &Engineerin(160. Mr. A. V. Dienhart October 6, 1969 If you should follow with the proposed scheme on your plant it should be noted that you are replacing the two turbine driven ptumps by two motor driven pumps with no sharing between Unit 1 and Unit 2. We therefore request that you forward your coussents and recomumendations to us at the earliest. Please refer to our letter PIP-N-414 dated June 9,1969. It is suggested that a meeting with Westinghouse prior to a AEC-DRL aceting on this subject should be arranged. A complete write-up on the Auxiliary feedwater Ryste= vill t,a trans=itted in the near future. If you have further cuestior.s please do not hesitate to call upon me. Very truly you'rs, s c wa-Yh a W Jay S. Iyer, Nuclear Engineer H. Hollingshaus, Manager Nuclear Power Engineering JSI/agh H Eollingshaus DH 1eppke VH Simon 3 l a}}