ML20203A063
| ML20203A063 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/13/1998 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 6L20-98-20063, GL-96-06, GL-96-6, NUDOCS 9802230266 | |
| Download: ML20203A063 (5) | |
Text
_ ______
e e
l GPU Nuclear. inc.
(
Route 441 South NUCLEAR Post Office Box 480 Middletown PA 17057 0480 Tel717 S44-7621 February 13, 1998 6L20-98-20063 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555
Dear Sir:
1 Subject.
Three Mile Islarid Nuclear Station, Unit I (TMI-1)
Operating License No. DPR 60 Dockct No. 50-289 GPU Nuclear (Revised) 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 licens es determine if the piping which penetrates contaironent 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 rater hammer or two-phase flow conditions during postulated accidents. We wete further requested to assess operability of any sys: ems fcund to be susceptible to these phenomena and to provide a report on the actions taken, conclusions reached relative to susceptibility (includirg: identification of systems affected and the specific circumstances involved);
the basis for continued operability of affected systems end 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 mtems remain operable. The purpose of this letter is to revise our rebmary 14,1997 response to reflect Jitional analysis which further demonstrates the capability of the RBEC system and to reflect the completion ofmodifications installed during the Cycle 12 Refueling (12R) Outage which ended in October 1997.
The concern identified by GL 96-06 involves the potential for heat transfer from an accident emironment to 2
adversely affect the ability of the containment fan coolers and the piping which penetrates containment to perform their intended fimetiens. Specifically, this concern addresses whether the liquid within containment
' For TMI l the containment air cooling water system ic which the GL refers is the Reactor Building Emergency Cooling (RBEC) System which is comprised of three air handling units (AH-EINB/C) cooled by river water under accid:nt conditions.
I 2
The less of Coolant Accident (LOCA) and Main Stea n Line Break (MSLB) inside containment are the only accidents which could cdd significant thermal energy to the containment. The Large Break LOCA was found to be the bounding accident for I
TMI-1.
/
{i Lj. ng,
.,****,;h,.,,h.'
'g 9802230266 980213 PDR ADOCK 05000289 u
a 6'
I P
PDR w
6L20-98-20063 Page2or4 fan cooler cooling ceils would boil leading to degraded heat 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 vt.Incrability of the fan coolers to void formation as well as the potentid 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. The following is a summary of our evaluations:
A.
Containment Ceqler Evaluations in support of the efrons to evaluate the GL 96-06 issuea a fan cooler model was developed to assess the vulnerability to void formation within the cooling coils and distribution piping. Tae GOTHIC computer code (version 5.0e) provides the analytical tool to model the fan cooler system. The model incorporates heat transfer to the cooling coils a; well as the system piping within the containment for one train of the Reactor Building Emergency Cooling System Reliefvalves (RR-V11 A/B/C)which were located at the top of the cooling coils' have teen relocated outside containment. Cooler isolation valves (RR-V4A/B/C/D) are provided at the system outlet
- in an initially closed position. The Nuclear Senices Closed Cooling System Surge Tank (NS-TI), which is normally connected to the inlet side of the piping for leak detection purposes provides an overpressure on the Reactor Building Emergency Cooling System piping inside containment. The surge tank would normally be available to provide an overpressure on the coolers during an accident. However, analysis has demonstrated that this overpressure is not required to prevent water hammer oc two phase flow.
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), failure of the relief valve to close following actuation, inlet and discharge valves in either open or closed position, and with or without the overpressure normally provided by the NSCCW System. 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 (0 to 130) psig and a temperature of 130
- F. The temperature assumes thermal equilibriam 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 relief valve setpoint. Tht, relief valve 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 cf the fluid is accommodated by flow past the valve. The analysis reveals minimal (< 1%) voiding of the system during this evolution. The event is terminated upon start of the RBEC pumps (RR-PI A/B).
' Each of the three cooling units contains a relief valve.
- RR-V4A and RR-V4R are the nutlet isolation valves for Emergency Cooling Coils A and B respectively. RR-V4C and RR-V4D are provided a isolation valves in parallel for Emergency Cooling Coil C.
' Subsequent to modification during the 12R Outage, RR-V-6 now has a safety grnde air supply.
- Norma: Reactor Buildmg temperature is approximately 90
- E
' In the consrvalve case the relief valve is assumed to stick open.
,n
6L20-98 20063 Page 3 of 4 ThereSre, since our analyses show that only minimal voiding occurs even in the worst case design-basis o Ments, neither two phase flow nor water hammer present a concern for TMI-l containment air coolers.
B.
Containment Penetration Pipe Section Evaluations in this evaluation, a total of ning.'seven (97) containment penetrations were identified for consideration. Each of these penetrations was reviewed against a specific set of criteria for susceptibility to the GL 96-06 concems and ten (10) were found to be susceptible to the GL 96-06 concerns.' Those ten (10) piping segments which w ere 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 Drain Tank (RCDT) Transfer Line, Reactor Coolant Pump Cooling Return Line, and the "A" and "B" Core Flood Tank Sampling Lines.
Tl ese pipe sections will have liquid trapped between two closed valves following an accident. The accident conditions inside containment will promote heat transfer and increase the pressure of the fluid trapped within the piping. The pressure inc. ease will place stresses on the piping and could potentially chall mge the integrity of the pipe and compromise containment integrity. A stress calculation was perfo med for each of the pipe sections identified to assess the poter.tial impact on the primary contaiament integrity.
The firsi step in performing the pressurization analysis was to identify the appropriate initial fluid conditiors prior to being exposed to an accident environment.' The next step was to incotporate models ol'the piping systems into a containment model (GOTHIC) and 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 each pipe section and incorporated into the subsequent stress calculation. The stress calculation evaluates the fluid structure interaction using steam table propert'es and stress strain relationships of the piping material. The fluid pressure and piping stress for each piping segment is obtained using an iterative solution technique which solves for a fluid stmeture equil'brium condition.
It was found that seven (7) of the ten (10) affected piping segments could exceed material yield stresses, howevet none exceeded the ASME Section III, Appendix F criteria. Although the penetrations that were affected by the GL 96-06 concem remained operable, GPU Nuclear installed pressure reliefdevices during the 12R Outage to prevent the overpressure condition from occurring.
- This event was reported in Licenset Event Report (LER) 97 001 as a condition that us outside of the design basis of the plant.
' Initial!y the kcactor Coolant Pump (RCP) Scal Water Return Line piping segment was found to be potentially susceptabic; but it w s later shown to be acceptable by adlitional calcuations.
1
e 6L20 98-20063 k
Page 4 or4 C.
Relief Valve Leakage Concerns in evaluating the potential for overpressurization ofisolated piping, the GL requested that any relief valves installed to prevent overpressure conditions be considered for flooding or radiation hazards in the event they were to lifl and fail in the open position.
The RBEC System was the only system with relief valves (RR-VI 1 A/B/C) that could relieve inside containment and contribute a significant amount ofwater to flood the Reactor Building during a design-basis accident ifit were to lift and fail to close.'" The source ofleakage through RR-Vi 1 is an unlimited supply of river water. Failure of an RR-VI1 relief valve to close aller opening was evaluated for post accident radiation hazards, flooding, and boron dilution concerns. Rather than rey upon operator actions to control this potential flooding event inside the R eactor Building, GPli Nuclear relocated the RR-V-I l relief valves to a location outside of the containment during the 12R Outage.
The details ofour evaluations have been documented in a GPU Nudear Technical Data Report (TDR) No.
1212. Therefore, GPU Nuclear has completed all of the actions required in response to GL 96-06.
Sincerely,
?g&
/
James W. Langen ch Vice President and Director, TMI Attachment MRK cc:
Administrator, Region 1 TMI Senior Resident inspector TMI Senior NRC Project Manager
' Other system piping with relief valves inside containment (e.g. Intermediate Closed Cooling Water, Decay Heat Removal. Core Flood, Nuclear Services Closed Cooling Water, Reactor Coolant System, etc.) have already bum included in the Reactor Building flooding calt.fations or would remain isolated and therefore could not contribute a significant amount of miter to flooding.
- 6L20-98-20061 Attachment h1ETROPOLITAN EDISON COhiPANY JERSEY CENT RAL POWER AND LIGHT COMPANY PENNSYLVANIA POWER AND LIGliT COMPANY GPU NUCLEAR INCORPORATED Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 1, James W. Langenbach being duly sworn, state that I am a Vice President of GPU Nuclear, Inc. and that I am duly authorized to execute and file this response on behalf of GPU Nuclear. To the best of my knowledge and belief, the statements contained in this document are true end correct. To the extent that these statements are not based on my personal knowledge, they are based upon infom1ation by other GPU Nuclear employees and/or consultants. Such information has been reviewed in 9ccordance with company practices and I believe it to be reliable.
/ }p%) fh 4%
James W. Langentac Vice President and Director, TMI Signed and sworn before me.his h
lb day (plen~A998.
d.
qulL
/ NotaryPu'blic (
w sev
\\
Sura w C U T " r y Punbc C8 Fly
(. nre 1
Ml GC' L
- 0. 1 W.fi Memcer, PennS/Wama Assecution ct Natanes n
0