ML20138L874

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

ML20138L874
Person / Time
Site:	Crane
Issue date:	02/14/1997
From:	James Knubel GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6710-97-2065, GL-96-06, GL-96-6, NUDOCS 9702250151
Download: ML20138L874 (6)

Text

s

{

GPU Nuclear,inc.

(

Route 441 South NUCLEAR P**' 0"ic' B** ***

Middletown, PA 17057 0480 Tel 717-944 7621 Febmary 14,1997 6710-97-2065 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington,DC 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 GPU Nuclear Response to Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions" Generic Letter (GL) 96-06 requested that licensees determine if the piping which penetrates containment is susceptible to overpressurization from thermal expansion of fluid trapped between the containment isolation valves or if the containment air cooling water systems' are susceptible to either waterhammer or two-phase flow conditions during postulated accidents. We were further requested to assess operability of any systems found to be susceptible to these phenomena and to provide a report on the actions taken, conclusions reached relative to susceptibility (inc:uding:

identification of systems affected and the specific circumstances involved); the basis for continued operability of affected systems and components, as applicable; and, corrective actions implemented or planned.-

GPU Nuclear has performed the evaluations as requested by the GL, including those specific scenarios referenced in the GL, and the results show that the affected systems remain operable.

Design verification of the calculations and analyses required to support these results is not yet complete. Therefore, if any of these results should change, this will be reported in a supplemental response.

The concern identified by GL 96-06 involves the potential for heat transfer from an accident 2

environment to adversely affect the ability of the containment fan coolers and the piping which

' For TMI l the containment air cooling water system to which the GL refers is the Reactor Building Emergency Cooling (RBEC) System w hich is comprised of three air handling units (AH-El A/B/C) cooled by river water under 9Y accident conditions.

The Loss of Coolant Accident (LOCA) arxl Main Steam Line Break (MSLB) inside containment are the only accidents which could add significant thermal energy to the containnent. The Large Break LOCA was found to be the bounding accident for TMI l.

! g 9702250151 970214 PDR ADOCK 05000289 p

PDR

(

4 6710-97-2065

(

Page 2 of 5 l

]

penetrates containment to perform their intended functions. Specifically, this concem addresses whether the liquid within containment fan cooler cooling coils would boil leading to degraded heat i

s transfer and possibly water hammer as well as whether heating of the liquid within the isolated piping penetrations could over-pressurize the piping. To assess the significance of this concern, GPU Nuclear evaluated the vulnerability of the fan coolers to void formation as -11 as the potential stresses on piping penetrations when isolated.' The results of these evaluations were used to determine equipment operability as well as the need for procedure changes or design modifications.

l The following is a summary of our evaluations:

j

- A.

Containment Cooler Evaluations In surpon of the efforts to evaluate the GL 96-06 issues a fan cooler model was developed to assess ie vulnerability to void formation within the cooling coils and distribution piping. The -

l GOTHIC computer code (version 5.0e) provides the analytical tool to model the fan cooler j

system. The model incorporates heat transfer to the cooling coils as well as the system piping within the containment for one train of the Reactor Building Emergency Cooling System.

Relief valves (RR-VI 1 A/B/C) are located at the top of the cooling coils' and the cooler isolation valves (RR-V4A/B/C/D) are provided at the system outlet' in an initially closed position. A surge tank is connected to the inlet side of the piping for leak detection purposes (through a check valve, NS-V12). The Surge Tank (NS-TI) is from the Nuclear Services Closed Cooling System which provides an overpressure on the Reactor Building Emergency Cooling System piping inside containment. The surge tank,'which must be manually isolated, would therefore maintain an overpressure on the coolers during an accident.

A variety of cases were analyzed using conservative worst case assumptions including: loss of offsite power with and without subsequent loss of air to the pressure control valve RR-V6 (loss of air causes RR-V6 to fail open), and failure of the relief valve to close following actuation. The conditions downstream of the RR-V4 valve are conservatively assumed to be at standard atmospheric conditions. The containment environmental conditions are based upon the bounding Equipment Qualification (EQ) temperature and pressure profiles. The coils and piping are initially assumed to be filled with liquid and maintained at a pressure of approximately 55 psig and a temperature of 130

• F. The temperature assumes thermal equilibrium between the stagnant fluid and the Technical Specifications maximum containment temperature (130

• F)'. The heating of the fluid causes it to expand raising the pressure to the reliefvalve setpoint. The reliefvalve opens relieving the pressure and closes several times prior to opening of the outlet valve.' During this time there is no voiding within the system as a direct result of the elevated pressure within the system being well above the saturation pressure of the fluid. As the outlet valve opens, further expansion of the fluid is accommodated by flow past the valve. The analysis does not reveal any voiding of the system during this evolution due to the surge tank providing makeup and assisting to maintain system 5 Each of the three cooling units contains a relief vahe.

• RR-V> A and RR V4B are the outlet isolation vahrs for Emergency Cooling Coils A and B respecthcly. RR-V4C and RR-V4D are prmided as isolation vahts in parallel for Emergency Cooling Coil C.

5 Normal Reactor Building temperature is approxinutely 90 'F.

• In the conservative case the relief vahr is assumed to stick open.

t 6710-97-2065 l

Page 3 ors i

pressure as the RR-V4 valve begins to open. The aystem pressure is maintained above the i

saturation pressure of the fluid. The opening of the valve into a low pressure downstream condition produces a rapid outflow of fluid. This allows the system to de-pressurize, but voids do not form within the cooling coils because the surge tank is able to provide sufficient makeup flow and pressure control. The event is terminated upon stait of the RBEC pumps (RR-PI A/B).

l l

In the case where the control valve has air available to control the position, the analysis is

}

basically the same as that described above. The main difference is that surge tank makeup is i

not required.

4 In the cases where reliefvalve RR-VI 1 is assumed to fail open with or without control air to RR-V6, the analysis still shows that voids do not form. The main difference is the surge tank makeup rates.

l Therefore, since our analyses show that voiding does not occur even in the worst case design-basis accidents, neither two phase flow nor water hammer present a concern for TMI-l i

containment air coolers.

a B.

Containment Penetration Pipe Section Evaluations

[

In this evaluation, a total ofninety seven (97) containment penetrations were identified for consideration. Each of these penetrations was reviewed against a specific set ofcriteria for susceptibility to the GL 96-06 concems and eleven (11) were found to be susceptible to the GL 96-06 concerns.7 Those eleven (11) piping segments which were determined to be affected are as follows: the "A" and "B" Once Through Steam Generator (OTSG) Sampling

[

Lines, Intermediate Closed Cooling Water (ICCW) Return Line, Reclaimed Water Supply.

Line, Makeup and Purification Letdown Outlet Line, Pressurizer and Reactor Coolant Sampling Line, Reactor Coolant Pump (RCP) Seal Water Return Line, Reactor Coolant Drain Tank (RCDT) Transfer Line, Reactor Coolant Pump Cooling' Retum Line, and the "A" and "B" Core Flood Tank Sampling Lines.

j

'l nese pipe sections will have liquid trapped between two closed valves following an accident.

j The accident conditions inside containment will promote heat transfer and increase the pressure of the fluid trapped within the piping. The pressure increase will place stresses on the piping and could potentially challenge the integrity of the pipe and compromise l

containment integrity. A stress calculation was performed for each of the pipe sections identified to assess the potential impact on the primary containment integrity.

The first step in performing the pressurization analysis was to identify the appropriate initial fluid conditions prior to being exposed to an accident environment. The next step was to incorporate models of the piping systems into a containment model (GOTHIC) and

' This event will be reponed in Liansce Event Report (LER) 97 001 as a condition that was outside of tle design basis of the plant.

i 6710-97-2065 Page 4 of 5 i

' analytically expose them to the accident. The computer model of the piping incorporates full heat transfer analytical capabilities. The peak fluid temperature is ~obtained in this manner for i

each pipe section and incorporated into the subsequent stress calculation. The stress i

calculation evaluates the fluid stmeture interaction using steam table properties and stress strain relationships of the piping material. The fluid pressure and piping stress for each piping l

l segment is obtained using an iterative solution technique which solves for a fluid structure equilibrium condition, j

4 Seven (7) of the eleven (11) affected piping segments could exceed material yield stresses,

[

however none exceeded the ASME Section III, Appendix F criteria. Although the penetrations that are affected by the GL 96-06 concem remain operable, GPU Nuclear is l

j considering hardware changes or procedural actions to assure compliance with applicable j

design code requirement.

i i

l C.

ReligfValve Leakane Cecsm_s l

In evaluating the potential for overpressurization ofisolated piping, the GL requested that any j

relief valves installed to prevent overpressure conditions be considered for flooding or radiation hazards in the event they were to lift and fail in the open position.

)

i i

The RBEC System is the only system with reliefvalves (RR-VI 1 A/B/C) that relieve inside

]

t containment and which could contribute a significant amount ofwater to flood the Reactor f

Building during a design-basis accident ifit were to lift and fail to close.' The source of leakage through RR-VI1 is an unlimited supply ofriver water. Failure of an RR-VI1 relief valve to close after opening was evaluated for post accident radiation hazards, flooding, and boron dilution concerns.

RBEC system pressure is higher than maximum containment pressure so any leakage would be into the Reactor Building. This prevents any leakage of the contaminated post accident containment envirorunent from becoming a radiation hazard.

j Preliminary evaluations show that adequate time is available for detecting and isolating the i

leaking reliefvalve before dilution of reactor coolant or flooding of safety related equipment would occur. Adverse conditions related to boron dilution or decreased pH of Reactor

{

Building sump water would be detected from Post Accident Sample results. The Abnormal Transient Procedures (ATP's) for Large and Small Break Loss ofCoolant Accident (ATP 1210- 6 and 1210-7, respectively) both require post accident sample results for boron concentration and pH. Isolation of the leakage source at this time would not impact the effectiveness of the Emergency Core' Cooling System (ECCS) which includes the Reactor Building Emergency Cooling System because only one of the three fan coolers would be j

j required to cool the Reactor Building. Although the affected equipment remains operable, to

' Other system piping with relief valws inside containment (e.g. Intermediate Closed Cooling Water, Decay Heat i

Removal, Core Flood, Nuclear Senices Closed Cooling Water, Reactor Coolant System, etc.) have already been j

included in the Reactor Building flooding calculations or would remain isolated and therefore could not contribute a j

significant amount of water to flooding.

I

j a

i 6710-97-2065

{

Page 5 of 5 i

avoid reliance upon the operators to detect and isolate a leaking reliefvalve and also because such leakage would be undesirable, GPU Nuclear is considering hardware changes or procedural actions that would eliminate the potential for this type ofleakage to occur.

1 The details of our evaluations are being documented in a GPU Nuclear Technical Data Report.

(TDR) which will be finalized upon completion of the design verifications for the calculations and the identification oflong term conective actions. Overpressurization concems for some of the affected containment penetration piping segments may possibly be resolved through additional 4

i analysis. GPU Nuclear commitments for the long term conective actions will be described m a l

l supplement to this response by May 31,1997.

l Sincerely, 1

t

[

~.

fp bb I

l.

J.Knubel Vice President and Director, TMI l

i j

i i

t Attachment

{

MRK i

I cc:

Administrator, Region I l

TMI Senior Resident Inspector i

l TMI Senior NRC Project Manager d

4 J

t

)

i

.~

l 1

i i

i i

e

i i

6710-97-2 % 5 l

. Amchment 1

I METROPOLITAN EDISON COMPANY

{

JERSEY CENTRAL POWER AND LIGHT COMPANY j

PENNSYLVANIA POWER AND LIGHT COMPANY GPU NUCLEARINCORPORATED i

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 t

Docket No. 50-289 i

l I, James Knubel being duly swom, state that I am a Vice President of GPU Nuclear, Inc. and that I am duly authorized to execute and file this response on behalt'of GPU Nuclear. To the best ofmy j

knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information by

)

other GPU Nuclear employees and/or consultants. Such information has been reviewed in accordance l

with company practices and I believe it to be reliable, j

{

l i

t j

J. Knubel i

Vice President and Director, TMI Signed and sworn before me this l

[h dayof

/unao 1997.

~

/

ll/A4

//

No~tary Public ~/

Notarial Seal Debra S. Klick, Notary Public Londondorry Twp., Dauphin County My Commission Expires July 4,1908 i

Member, Pains /tania Asso222cn of itf;vios r-