ML20147B322

Responds to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions

ML20147B322
Person / Time
Site:	Waterford
Issue date:	01/28/1997
From:	Gaudet T ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, GL-96-6, W3F1-97-0017, W3F1-97-17, NUDOCS 9701310030
Download: ML20147B322 (9)

Text

• 1

\

g Ent rgy

- ~ -

ggymmu=. i.x.

Operations - u me Td 504-739 6650 l

l W3F1-97-0017 4 A4.05 l PR l

I January 28,1997  ;

1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

Waterford 3 SES Docket No. 50-382 I License No. NPF-38 l NRC Generic Letter 96-06 Assurance Of Equipment Operability and Containment Integrity During Design Basis Accident Conditions l 1

Gentlemen: l Pursuant to NRC Generic Letter (GL) 96-06, addressees were notified of safety significant issues that could affect containment integrity and equipment operability during certain accident conditions. The generic letter requested the following actions with respect to referenced scenarios:

(1) Determine if containment air cooler cooling water systems are susceptible to either waterhammer or two-phase flow conditions during postulated accident conditions; (2) Determine if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.

This response provides a written summary indicating actions taken relative to the generic letter along with, in Attachment A, conclusions reached regarding the susceptibility of Waterford 3's containment fan cooler cooling water system to waterhammer and two-phase flow. This evaluation concludes that the waterhammer and two-phase flow scenarios described in the generic letter are not a concern at Waterford 3. The provisions of the Waterford 3 design ensure single-phase flow to the unit containment fan coolers prior to operation in a design basis event and hence meeting associated heat removal assumptions.

i 9701310030 970128 PCR ADOCK 05000382 P PDR

j '

i NRC Generic Letter 96-06 Assurance of Equipment Operability and Containment Integrity During Design Basis

Accident Conditions

W3F1-97-0017 j Page 2 I j January 28,1997  !

I l l

\

! In Attachment B, determinations pertaining to containment penetration piping j susceptibility to overpressurization are provided. Also, incorporated within the j summary, is the basis for continued operation of affected systems and components, I

as applicable, and corrective actions, as appropriate, which will be completed by Refuel Outage 9, currently scheduled for the Fall of 1998.

4 There were 12 penetrations reviewed for effects caused by pressure locking or thermal binding in accordance with GL 95-07. None of the valves required to open post accident are susceptible to pressure locking or thermal binding.

Waterford 3 has concluded that none of a questioned population of 12 containment penetrations exceed burst pressure. The piping may yield and deform, but the piping will not fail. Since the safety function of the penetrations is not compromised, l a basis for operability in accordance with GL 91-18 is established. The associated i nonconforming condition is documented in the Waterford 3 corrective action program. Because all the affected containment penetrations retain their ability to perform their safety function and thus ma!ntain containment integrity, this nonconforming condition is not considered reportable under the 10CFR50.72/50.73 criteria.

Please contact me at (504) 739-6666 or Phillip R. Snowden (504) 739-6348 should any questions arise.

Very truly yours, yA W T.J. Gaudet Acting Director, Nuclear Safety and Regulatory Affairs TJG/ PRS /ssf Attachment cc: E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRR),

R.B. McGehee, N.S. Reynolds, NRC Resident inspectors Office

l UNITED STATES OF AMERICA -

! NUCLEAR REGULATORY COMMISSION i

In the matter of )

5 )

Entergy Operations, Incorporated ) Docket No. 50-382 j Waterford 3 Steam Electric Station )

l l AFFIDAVIT l

! Timothy Joseph Gaudet, being duly sworn, hereby deposes and says that he is Manager, Licensing - Waterford 3 of Entergy Operations, Incorporated; that he is duly l authorized to sign and file with the Nuclear Regulatory Commission the attached

Response to NRC Generic Letter 96-06; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, l

information and belief.

1 l l

[ A--itp ban AM j Timothy Jose $ DIaudet' j Manager, Licensing - Waterford 3 4

t i

STATE OF LOUISIANA )

) ss PARISH OF ST. CHARLES )

Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 28" day of _ d Aw<vt T .1997.

~

bc >IL Ic.

Notary Public My Commission expires W 'm 'i re

! . Attachm::nt to W3F1-97-0017 Page 1 of 6 ATTACHMENT A Containment Fan Cooler Cooling Water System Susceptibility to Waterhammer and Two-Phase Flow Generic Letter 96-06 describes circumstances wherein, Containment fan coolers  ;

may be subject to failure during a LOCA with concurrent loss of offsite power, due to boiling of the component cooling water (CCW)in the fan coolers. The CCW may boil because heat is transferred from the containment atmosphere while CCW flow is stagnant, until the CCW pumps are loaded onto the Emergency Diesel Generator j (EDG). When the CCW pump is restarted, the steam voids would collapse possibly j resulting in waterhammer loads.

Waterford 3 has evalusted these concerns and has concluded that they are not applicable to Waterford 3 for the following reasons:

1. The CCW pumps and containment fan coolers are simultaneously loaded on the EDG in the 7 second load block. Thus, following a LOCA with concurrent loss of offsite power, stagnant conditions would on;y exist for a maximum of 17 seconds (10 seconds for the EDG to start and begin loading plus 7 seconds for the sequencer to load the pumps and fans). j This relatively short period of time limits the amount of heat that is i transferred to the stagnant CCW in the fan coolers.

l

2. The Waterford 3 design incorporates a CCW surge tank located on the I roof of the Reactor Auxillary Building (RAB). This configuration provides increased pressure above atmospheric pressure. Due to this elevation ,

head, the saturation temperature of the water in the fan coolers increases. l The elevation head of 62 feet (between the bottom of the surge tank and the top of the fan cooler coils) creates an additive pressure of 26.7 psi, which increases the saturation temperature to 269'F. Since the maximum temperature calculated for a LOCA is 269'F, it is very unlikely that CCW in the fan coolers will boil within 17 seconds after a LOCA starts.

3. The worst case containment temperature excursion at Waterford 3 occurs during a steam line break event. During this ennt, the cooling coil tubes will have a water film on the outside (containment side) due to condensation. The temperature of this condensate film will be at the saturation temperature of the partial pressure of steam in containment.

The CCW water inside the tube will remain at a temperature lower than this temperature as long as the condensate film is on the tube. FSAR l analyses for a steam line break show that the containment temperature '

and pressure at 17 seconds after the break is about 375'F and 44 psia.

l l

3

]. .

. Attachment to W3F1-97-0017 Pags 2 of 6 However, the partial pressure of the steam at 17 seconds is less than 23 7 psia. Saturation temperature at this partial pressure, and thus, the temperature of the condensate film on the outside of the cooling coil, is i 235'F. Thus, the CCW temperature will remain be!ow saturation.

a Conclusion i

This evaluation concludes that the waterhammer and two-phase flow scenarios j described in the generic letter are not a concern at Waterford 3. The provisions of J the Waterford 3 design ensure single-phase flow to the unit containment fan coolers prior to operation in a design basis event and hence meeting associated heat 4 removal assumptions.

J i-

. . Attachm::nt to VV3F1-97-0017 Pago 3 of 6 ATTACHMENT B  !

1

, Evaluation for Susceptibility of Containment Penetration l Piping Overpressurization Due to Thermal Expansion of Fluid 4

1 An evaluation was performed of the Waterford 3 containment mechanical '

penetrations utilizing the following criteria developed for all Entergy sites:

Enterav Generic Letter 96-06 Penetration Evaluation Criteria l

1) Scope: All containment (Pressurized Water Fteactor (PWR) and Boiling Water Reactor (BWR)) and drywell f,WR penetrations.

2) A penetration piping system, including any connected heat exchangers, will be considered to be Potentially Susceptible if it meets aji of the following four criteria:

A.) The penetration must be full of liquid at the time of the accident. Pipes containing air, gas or steam will be excluded.

B.) The liquid contained in the penetration piping must be at a lower I temperature than the surrounding environment during operational or I accident situations. Piping that contains water at or near Reactor Pressure Vessel (RPV) or Steam Gerarator (SG) temperatures, such  !

as feedwater, letdown, blowdown or Reactor Water Cleanup (RWCU), I wou!d actually have initial fluid temperatures higher than those  !

expected during an accident.

1 C.) The penetration must be isolated during an event, i.e. plant heatup or l accident, that could cause a significant heat transfer to the fluid beiween the isolation valves. The valve arrangement used for penetration isolation must restrict flow out both directions. If the inboard isolation valve is a check valve or certain type and orientation of solenoid valve, (with a mechanism of pressure relief in the connecting piping) the penetration may possibly be excluded. This exclusion would also include piping open to the suppression pool, SG, RPV, or containment air space.

.. _ 7

. . . Attechment to W3F1-97-0017 Pago 4 of 6 l- In order to be excluded, the extended piping system available I l for fluid expansion inside containment must not constitute a i j closed system, so that the fluid volume can expand and prevent j damage to the containment isolation portion of the piping j penetration. I i

l Additionally, another closed valve further down the line inside l containment must not prevent expansion of the fluid volume in j the penetration, thereby isolating a penetration with an expected l available leak path i.e. check valve.

D.) The potentially susceptible penetration will not have any pressure relief  !

valves (with sufficient capacity and setpoint) or other method of overpressure protection (such as a check valve in parallel with the main inboard valve) between the isolation valves.

A penetration will additionally be considered potentially susceptible if it meets any one of the following two criteria:

The penetration will be considered potentially susceptible if a single-failure coupled with an accident would cause isolation, heatup and ,

overpressurization, of a normally open, low temperature, fluid filled I penetration.

A penetration will be considered potentially susceptible if trapped pressure can prevent safety-related isolation valves from opening when required to mitigate an accident, i.e. pressure locking. Ref.:

Generic Letter 95-07.  ;

A Potentially Susceptible penetration may be eliminated from concern if l qualified calculations or analyses demonstrate that the penetration piping 1 system, which includes the valves, remains within its Design Basis.

Penetrations that do not meet their Design Basis requirements shall be considered Susceptible, and have a basis for operability established in accordance with Generic Letter 91-18 guidance.

1

. . , Attachment to VV3F1-97-0017 Pago 5 of 6 At Waterford 3, seventy-four (74) containment penetrations were evaluated per the

, preceding criteria. Of these 74 containment penetrations, the following 12 were considered potentially susceptible:

Penetration # System Fluid Dwg 7 Primary Makeup Water G-161 Sh 2 24 CCW Retum From RCP's Water G-160 3h 4 28 RCS Sampling Water G-162 Sh 2 29 PZR Surge Line Sampling Water G-162 Sh 2

, 3^ _

PZR Steam Space Sampling Water G-162 Sh 2 4e Containment Sump Pump Discharge Water G-173 Sh 3 43 Reactor Drain Tank Outlet Water G-171 Sh 1 44 RCP Controlled Bleedoff Water G-168 Sh 2 )

51 Refueling Water Supply to Refueling Cavity Water G-163

59 SIT Fill / Drain From RWSP Water G-167 Sh 4 62 Refueling Cavity Drain Pump to RWSP Water G-163 l 71 Demineralized Water Water G-161 Sh 2 l Initial analysis determined that all 12 potentially susceptible penetrations exceeded their normal code allowables and would require a basis for operability established in accordance with the guidance contained in Generic Letter 91-18. In accordance with GL 91-18, a penetration is considered operable if analysis demonstrates that  ;

the penetration piping system mairitains its ability to perform the safety function it l was designed to perform.

l A calculational analysis of the 12 containment penetrations has been prepared. This calculation determines the maximum internal pipe pressure due to thermal expansion of the water, and considers the expansion of piping due to pressure from the encased water and the increase in pipe temperature. The change in pipe volume will result from the elastic and plastic deformation of the pipe due mostly to the internal pressure, and partly due to the temperature rise. No consideration was given for leakage past the isolation valve seats. This assumption is conservative and results in an internal pipe pressure greater than actual values.

The resulting internal pipe pressure was then comparcd to the burst pressure of the pipe. The calculation concludes that none of the 12 penetrations exceed burst pressure.

.j._..__._.___..._-._.__._____._...___.-______._

..- . . Attachment to

{ .

W3F1-97-0017 j- Pago 6 of 6 i

j Conclusion i None of the 12 penetrations exceed burst pressure. The piping will yield and deform j to relieve the pressure, but the piping will not fail. Since the safety function of the penetrations is not compromised, a basis for operability in accordance with GL 91-18 is established. This conclusion is further supported by available Local Leak Rate Testing (LLRT) leakage data; valve leakage would decrease the internal pressure j and provide additional margin for operability.

i-

! These 12 penetrations were reviewed for effects caused by pressure locking or j thermal binding in accordance with GL 95-07. None of the valves required to open

post accident are susceptible to pressure locking or thermal binding.

l Reportability Determination l

i j in performing the review requested by Generic Letter 96-06, Waterford 3 noted i several containment penetrations that were potentially susceptible to i overpressurization. While detailed engineering evaluations demonstrated that the

' postulated overpressurization of these penetrations would not jeopardize the ability l of the penetrations to perform their safety functions (i.e., containment isolation); it i was determined that in some instances that ASME Ill Class 2 code limits could be i exceeded. Waterford 3 considers this to be a nonconforming condition as described

) in Generic Letter 91-18 and thus addressed the issue accordingly.

I l

! In accordance with Generic Letter 91-18, the noncenforming condition was

{ documented within the corrective action program and a prompt determination of

i. operability determined and documented for each affected penetration. Following assurance that containment integrity would be maintained (i.e., no safety concern exists), the nonconforming condition was evaluated for potential reportability requirements. In this instance the reportability determination was contingent on the interpretation of the phrase "outside the design basis". Waterford 3 believes guidance provided for making this interpretation directs one to focus on preservation j of defense in-depth particularly as it relates to protection of fission barriers.  !

In this case since all affected containment penetrations retain their ability to perform their safety function and thus containment integrity is maintained, this nonconforming condition is not considered reportable under the criteria of 10CFR 50.72/50.73. As with other nonconforming conditions, appropriate corrective action to restore the condition to within the required quality requirements will be taken in accordance with the safety significance of the issue.

J

, - -, - . , . - - + . -.- -- -