ML20117G781

Discusses Water Sys Operability Evaluation for Water Hammer Impact on Piping & Support Sys.Procedure Encl

ML20117G781
Person / Time
Site:	Point Beach
Issue date:	08/28/1996
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20117G782	List: ML20117G781 ML20117G800
References
VPNPD-96-059, VPNPD-96-59, NUDOCS 9609050327
Download: ML20117G781 (20)

Text

_ _ .

l Wisconsin

Electnc l POWER COMPANY 231 W Mchigan. PO Box 2046. Mihuoukee, WI 53201 2046 (414)221 2345 VPNPD-96-059 August 28,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION Mail Station PI-137 Washington, DC 20555 Gentlemen:

DOCKETS 50-266 AND 50-3_0_1 SERVICE WATER SYSTEM OPERABILITY DISCUSSION POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 On August 8,1996, Wisconsin Electric personnel and NRC staff participated in a confere ice call. The purpose of the call was to discuss our ongoing evaluation of the applicability of the issues detailed in Westinghouse Nuclear Safety Adsisory Letter, NSAL 96-003, to our Point Beach Nuclear Plant. The NSAL discusses the potential for boiling in the cooling water for j containment accident fan coolers under design basis accident conditions. .

The design basis transient assumes a loss of off-site power (LOOP) concurrent with a loss of coolant accident (LOCA).

During the short period of time between the loss of power and the sequencing of safeguards loads on the emergency power supply, cooling water flow to the fan coolers stops. While the fans are coasting to a stop, containment atmosphere at elevated temperatures and pressures is projected to be drawn over the cooling coils, transferring heat to the stagnant cooling water in the coils. This heat transfer may result in water in the coils boiling. Upon resumption of cooling water flow, collapse of the steam bubble in the piping may result in a waterhammer of sullicient magnitude tojeopardize the integrity of the piping.

Prior to the call, we provided the NRC our discussion notes. The notes provided an oveniew of our system and our analysis >

to date. It also provided simplified drawings of our fan cooler and senice water piping elevations and a Probabilistic Safety Analysis of the efTects of this phenomena on core damage frequency and fission product release probability. Our notes and discussion provide our rationale for determining that the Senice Water System and Containment Fan Coolers remain operable at Point Beach under the conditions postulated in the NSAL.

As requested by Mr. Mark Ring of your staff, attached is a copy of our discussion notes. As committed to in our conversation, the formal analysis and final operability determination will be completed by September 9,1996.

If you have any questions, or require additional information, please contact us.

Sincerely,

/ '

t/

V Pr sident _ 9609050327 960828 PDR Nuclear Power ADOCK 05000266 JFM PDR Attachment ec: NRC Resident inspector, NRC Region III Administrator A subs &qofHismmhimqvCoqwathn

Service Water System Operability Evaluation - Water Hammer impact on Piping and Support System SIMPLIFIED HEAT TRANSFER MODEL (boilina of water in coolers)

This simplified model consists of one longitudinal foot of 1/2" copper piping, along with the corresponding number of fins. l l

The heat transfer coefficient due to condensation at the outside surface of the copper tubes is obtained from Point Beach FSAR. This peaks at 385 Btu /hr-ft' *F.  ;

i The heat transfer coefficient due to boiling of water inside the copper tubes is <

conservatively assumed to be the maximum value for nucleate boiling. This is I 2

approximately 7700 Btu /hr-ft ,.F, and it allows for more rapid boiling. <

l The results show that the entire water inventory in the coolers will boil in less than 10 ,

seconds. ,

1 Since the volume of one cooler unit (8 coolers) is approximately 8.5 ft', and the volume of the return piping for any of the four cooler units is less than 100 ft , the produced steam will be sufficient to push the water columns down to the bottom'of the discharge riser.

SIMPLIFIED FLUID FLOW MODEL (velocities of water in oioino)

This simplified model includes one unit, and it consists of three pumps, four coolers, and the corresponding valves, fittings and piping. However, the results are bounding for both units.

The isometric drawings are used to construct this simplified model, regarding elevations, fittings, and pipe lengths and diameters. The overall pressure drop across the cooler units is obtained from calculation 96-0117. This is 4.04 psi at 663 gpm, for clean coolers.

As the LOCA/ LOOP event is initiated, SW pumps will lose power, water in SW system will decelerate and come to rest, and accordingly column separation will occur in the retum lines of the cooler units at higher elevations. In addition, due to boiling of water inventory in the coolers, it will be conservatively assumed that all the water in discharge i I

header inside containment has been evacuated.

For conservatism, the following initial and boundary conditions are considered in the pump start transient:

._ m _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _

1- It is assumed that the steam produced due to boiling of water in the two cooler units at lower elevations will rise and escape to the discharge header, and that these coolers, along with the supply and return lines, will be water solid.

Therefore, no credit is taken for flow through these two coolers as the pumps start.

2- Only flow through the two coolers at higher elevations is considered in the l analysis. For conservatism, it is assumed that the return lines for these two coolers are filled with steam all the way to the 20" SW return header. This allows l for more time, and accordingly higher flow velocity, as the water hammer occurs.

It is also assumed that the back pressure at the retum lines of these coolers is equal to the vapor pressure at ambient temperature throughout the transient, and no credit is taken for pressurization due to steam generation at the coolers.

3- Since check valves exist on the supply lines, these lines are assumed to be water solid at the start of the pumps. The results show that by the time the return l lines are filled with water and water hammer is initiated, all three pumps would j have started and steady state flow conditions would have been achieved.

Therefore, any supply line water loss due to check valve leakage would only delay the water hammer but would not affect its magnitude.

4- The magnitude of the water hammer is based upon the relative velocity between the two water columns as the vapor bubble in the retum line collapses. This relative velocity is reduced due to the 8" retum line drainage as it is filling up.

For conservatism, no credit is taken for drainage.

l 5- Two phase flow will occur in the retum lines and not in the supply lines.~ For conservatism, frictional losses in the retum lines are not considered in the analysis. Instead, a conservatively low back pressure is used at the cooler outlet ]

throughout the pump start transient. Therefore, two phase flows will not affect  ;

the results. l 6- For conservatism, it is assumed that each pump reaches its rated speed instantaneously upon start, and no credit is taken for inertia of impeller. The refill is started at time zero with the three pumps starting at 0, 5 and 10 seconds, respectively.

7- The results of the analysis show that the return lines will be filled with water, and accordingly water hammer will be initiated in less than 20 seconds from pump start. The results also show that by then, steady state flow would have been achieved and the corresponding relative velocity in the 8" retum line will be 16.4 ft/sec. This is the maximum velocity in that line throughout the pump start transient event, and will be used to calculate the water hammer pressure in that line. Based on the above assumptions, these results are considered to be bounding for both units.

CALCULATION OF WATER HAMMER PRESSURE IN THE 8" RETURN LINE Consistent with NUREG-5220, the maximum impact overpressure is P = 1/2 r a V

= 1/2 (62.4/32.2)(4500)(16.4)

= 71500 psf '

= 500 psi  !

As explained in NUREG-5220, this value represents an upper bound, and actual loads !

are usually lower by a factor from 2 to 10. Reductions are due to cushioning by uncondensed steam or non-condensable gas, to compliance of piping, hangers and mounts, to oblique impact, and to reduction in slug length.

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Pioina Hydraulic Transient Analysis Ooerability Evaluation

1. Description of Model l . A PIPSYS computer model of the 8" and 2-1/2" return piping to cooler 1HX-15A was selected as representative of the piping attached to the four coolers. The model includes 2-1/2" piping from the connections to the cooler up to the containment penetration. A plot of the piping model is shown in Figure 1.0.

)

The following analyses were performed on the piping model: l Dead weight  ;

Thermal Hydraulic Transient Loading i Combined Stresses and Reactions  !

l Details of the transient evaluation are provided below.

2. Hydraulic Transient Analysis A simplified hydraulic transient evaluation was performed by running a direct l integration forced vibration analysis (PIPSYS Dynamic N option). A simplified force-time history was created for each elbow in the model. Each time history consisted of six points as illustrated below-Force

I to tr td tr 0.20 sec Time

l The maximum force in each leg was calculated as follows:

V = 16.4 ft/sec peak fluid velocity P = (1/2)*r*c'V' r = water density P = 500 psi c = sonic velocity (4,500 ft/sec) l-F = P*A A = pipe area F = 25,500 lbs for 8" pipe F = 2,450 lbs for 2-1/2" pipe The time steps input for each leg are summarized below; te = time for shock wave to travel from origin to end of leg where the force is applied (distance divided by speed of sound in water .

assumed as 4,500 ft/sec) tr = rise time arbitrarily chosen as 1 millisecond td = time for wave to travel from one end of the pipe leg to the other (leg length divided by speed of sound)

The integration duration was set at 0.20 see with 0.0001 sec intervals. A damping ratio of 2% was specified.

3. Impact on Cooler To evaluate the impact on the cooler, the hoop stress in the tubes was calculated and the potential water hammer impact on the bends in the coils was assessed. The calculated hoop stress is less than the allowable stress. The force on the tube bends will be bounded by F = 2*P*A = 270 lbs, where "2" is conservatively used for a dynamic load factor. Since this is a tensile force applied to the tube it is well within the capability of the tube material.

3. Analysis Results Intermediate runs showed relatively high loads on piping supports. Although these loads may exceed individual component allowables, the loads would not be expected to result in total failure in all cases (i.e. high side loads on extended ,

U-bolts). Therefore, for conservatism, subsequent runs were made assuming all the supports on the system failed.

The resulting analyses showed that pipe stresses were within the Point Beach operability stress limits. Based on the conservative nature of the analyses performed, the resulting water hammer will not impact piping operability.

1 PSA Evaluation of Water Hammer in SW Supply to Containment Accident Fan Coolers A recently identified scenario brings into question the ability of the fan coolers to respond to a Large Break Loss of Coolant Accident, (LOCA) with a Loss of Offsite Power (LOOP). It is postulated that once the service water pumps stop due to the diesel generator start, and Si sequencing, that the water in the fan coolers will flash. When the service water pumps restart, and the cold water is introduced into the fan coolers, the rapid cooling will cause a water hammer condition that couldjeopardize the integrity of the fan coolers. His condition is only postulated to occur for Large Break LOCAs and Steam /Feedline breaks inside containment, because the containment temperature must be high at the time of the SI sequencing to cause the service water flashing.

l This PSA evaluation will address the probability of the LOCA and LOOP occurring together, the possible consequences based on different assumed fan cooler conditions, and mitigating conditions that can be put in place to reduce the probability that the negative consequences will occur.

m

For purposes of simplifying the discussion, whenever LOCA is mentioned in the rest of this document, it will actually be referring to either a Large Break LOCA accident, or a steamline/feedline break inside containment.

Scenario Probability:

{ It is postulated that the scenario could occur anytime that a Large Break LOCA occurs, or a Steamline/Feedline 3 break inside containment occurs, in coincidence with a Loss of Offsite Power. From the PBNP PSA-93 model, the most recent available, the initiating event frequency for a Large Break LOCA is 2.5E-4/yr. He initiating event probability for a Steamline/Feedline break inside containment is 7.0E-4/yr. He sum of these initators is 9.5E-4/yr.

To determine the probability that a Loss of Offsite power will occur coincident with these initators, it is first necessary to describe how two accident can happen simultaneously. He LOOP and LOCA could occur at roughly the same time due to dual random failures. Also, the LOOP and LOCA could occur as dependent failures, such that the occurrence of one somehow led to the occurrence of the other. Since a LOCA will result in a reactor trip, and terminate the generation of electricity by the affected unit, the grid will be impacted which could cause an instability

] which could lead to a LOOP. It is not considered feasible that an initial LOOP could result in a Large Break LOCA j or a Large Steamline/Feedline break.

The PBNP PSA estimates that the probability that a LOOP will occur following a reactor trip is 1.42E-3. Thus the 1 total vrobability ofa devendent LOCA and LOOP is 9.3E-4 x 1.42E-3 or 1.35E-6hrr.

1 ne probability that an independent LOCA and LOOP will occur is first dependent on the time period under evaluation. This postulated scenario could occur anytime a LOOP were to occur following a LOCA as long as containment temperatures are elevated. For purposes of this calculation, we will evaluate a 1 hr time frame following LOCA initiation. He probability of a LOOP is 6E-2/yr. Thus the random probability of a LOCA, with a LOOP occurring within the next I hour is (9.5E-4/yr x 6E-2/yr x ! hour /7315 hours per reactor-year) = 7.8E-9.

This number is insignificant compared with 1.35E-6/yr and can be ignored.

In conclusion, the total probability of a LOCA and LOOP simultaneously is calculated to be 1.35E-6/yr.

Consequences:

The consequences of this scenario are dependent on the assumed condition of the Service Water system. Two scenarios will be described:

1) Under a worse case assumption, if the water hammer was very severe, it can be postulated that the service water piping in the fan cooler would fail, and the rupture of the system would fail the service water system entirely.

Under this assumption, every time the initiating event would occur,(LOCA and LOOP, or 1.35E-6/yr)it would also result in a core damage event, and a containment breach (due to de failure of the Service Water piping, which penetrates the containment.) and Large Early Release. This number represents an increme ofour total CDF freauency (1.69E-4hr) ofless than 8% and an increase ofour FPRF (3 67E-$h r) of 3. 7%

2) Using less severe assumptions, it is possible that the containment fan cooler coils would be damaged during this scenario, and the fans would be unavailable, but the resulting service water leakage that results is small enough that it does not impact the operability of the Service Water system to perform its other functions. Based on this assumption, the resulting core damage frequency will not be a function of the initiating event frequency for LOCA, but rather the subsequent CDF frequency as a result of a LOCA initiator. He CDF due to a Large Break  !

LOCA is 6.51E-6/yr and the CDF due to Steamline/Feedline break inside containment 3.73E-7 for a total of 6.9E-6/yr. As was calculated above, the probability of this coincident with a LOOP, (ignoring the insignificant  !

independent failure mode) is 6.9E-6 x 1.42E-3 = 1.0E-8/yr. Once again it is postulated that this results in a containment breach and a Large Early Release. This frequency represents about .006% of our CDF frequency, and

.03% of our FPRF. l l

NE1 has attempted to describe an industry methodolaev to use PSA evaluation to determine the severity of l operability issues This is documented in EPRI TR-103396. "PSA Anclication Guide " This guideline describes a 1 change in core damage probability (CDP) as a result of a plant change. The CDP is simply the change in core l damage frequency times the length of time that change is in effect. The application guideline than gives a figure l (Figure 4-3, Quantitative Screening Criteria for Temporary Changes) which is used to determine whether the temporary change is risk significant or not. An equivalent concept is used to evaluate changes to Large Early 1 Release Frequency, called LERP.

If it is assumed that the temporary change would be in effect for 6 months, the resulting CDP for Scenario I above would be (CDF Change) x (Duration) or 1.35E-6/yr x 1/2 year = 6.8E-7 CDP. He LERP would be (LERF change) x (Duration) = 1.35E-6/yr x 1/2 year - 6.8E-7 LERP.

The table describes three regions, of increasing level of severity. For evaluating the CDP, anything less than 1 E-6 is I considered non-risk significant. This means the change does not significantly increase the overall plant risk, and the change can be justified without the need for additional mitigating actions or analysis. The calculated CDP for this scenario is 6 8E-7 and would fall in the non risk signiReant nortion ofthe Rgure. In evaluating the calculated LERP of 6.8E-7. the same table comiders any LERP change of between IE-7 and IE-6 as in the erav area where the change is less than the Potentially risk signiReant area butgreater than the non-risk signi_Reant area. In this region. the PSA Avnlications Guideline states that it is necessary to assess non-auantiRable factors to determine the accectability of the temvorary change.

The resulting CDP and LERP assuming scenario 2 represents the actual final condition would be 1.0E-8/yr x .5 =

SE-9 for both CDP and LERF. His value is well into the non risk-significant regions for both CDP (<lE-6) and LERP (<1 E-7.)

Assess Non-Quantifiable Factors There are additional factors that could serve to mitigate the consequences of this cvent. For the first scenario, it is probable that the operators would identify the failed service water system, isolate the containment fan coolers and be able to restore the service water system to service. His would make the risk consequences of the first scenario much more similar to the consequences for scenario 2.

Point Beach has two containment spray pumps in addition to four fan coolers per unit. Two containment spray pumps are capable of protecting the containment from an overpressure condition. In addition, operating containment spray serves to reduce containment pressure and scrub the containment atmosphere. This would reduce a containment release.

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-- --;-----~~---- - --- -

WlSCONSIN ELECTRIC DG-M10 NUCLEAR POWER DEPARTMENT Revision 1 DESIGN AND INSTALLATION GUIDELINES Pace 2 ofJ Accendix B PIPE SUPPORT GUIDELINES

f. ,:1 y

tbpies to Ab$oo, Frieser,1.Lpke, Nwton - CWT 11/10/89

,.**'. a s .,,s, UNifs0siATeS

/i UCLE AR R EGUL ATORY CCMMISSION 2.p

. .,7 // seas u.a t o,s. o. c. noses

• ',?.f.4 Novesnber 8,1989 Docket Ns.s. 50 256 and 50-301 Mr. C. W. Fay, Vice President Nuclear Power Department .-

Wisconsin Electric Power Company NUCbCO ...** ' ' ~ b 3 231 W. Michigan Street, Room 3C8 till.4WLee, Wisconsin $32:1 Cear Mr. Tay:

5'BJECT: INTERIM CPERASILITY CRITERIA FCR SAFETT RELATED P! PING AfC ASSOCIATED SUPPCRTS (TAC N05. 7a503 AND 74504)

In a letter dated August 8,1989, Wisconsin Electric Pomer Cor.pany OitPCO) submitted for NRC review and approval a docunnnt entitlec *

" Criteria for Determining Justification for Continued Operation

= bun Encountering Major Discrepancies in 'As. Built' Safety Related Piping

• for the Potet Beach Nuclear Plant. WEFC0 further notes that these ' criteria are the same as those previously submitted to I the rtRC by Northern States Power Cos:pany under Occket Nos. 50 ZB2 and 50-3CG for the Prairie Island Nuclear Plant."

According to WIPCO, the referarced criteria are based on the ASME

! action !!! Appendix F values (1983 Edittun through Winter 1985 Addenda) and are intended to assure the opersoItty requirements of safety-related piping and associated supports if stresses are foued to exceed allo ables presented in the Point Beach Nuclear Plant Final Safety Analysis Report (FSAR). The criterit peral; operation for an inttrim perted only. Corrective modifications, restoring the system to FSAR allo ables, are to be made by the next refueling outage or sooner.

In your August J.1989, response to the Notice of Violation arising from inspection reports 50-266/89-004 and 50-301/B9-004 WIPC0 indicated that an operability evaluation was perforr.ed for each cf tre four integral welded attachments found to exceed 831.1 code 411o ables. The NRC requires licensees to make prompt operability determirations in those instances where degraded or non-conforming coeditions are found to esist. We note that WEPCO has Nde suCh a cetermination and that appropriate nuclear industry experience was utilized in doing so. We have no objection to your operability determination.

Etctivte fl0V 101W

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--- f f nl?

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\

. l WISCONSIN ELECTRIC DG-M10 NUCLEAR POWER DEPARTMENT Revision 1 DESIGN AND INSTALLATION GUIDELINES Pace 3 of 9 l Accendix B PlPE SUPPORT GUIDELINES

• 2'

)

1 tir. L.V. fay I This completes cur review relative' to 1AC Nos. 745C3 and 745C4.

If you have any questices, please contact sne. I sin erely, tn J '

Varren H. Sweeson. Project Manager  !

Project Directorate !!! 3 /

Division of Reacter Projects - !!!,

IV, V anc Special Projects Cffice of hutlear Reactor Kegulation cc: 5:e next page l

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