ML20058E299
| ML20058E299 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/21/1982 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| REF-SSINS-6835 5211-82-170, IEIN-79-22, NUDOCS 8207280144 | |
| Download: ML20058E299 (6) | |
Text
GPU Nuclear h __
g g{
P.O. Box 480 Middletown, Pennsylvan,a 17057 i
717-944-7621 Writer's Direct Dial Number:
July 21, 1982 5211-82-170 Office of Nuclear Reactor Regulation Attn: John F. Stolz Operating Reactors Branch No. 4 U. S. Nuclear Regulatory Commission Washir.gton, D.C.
20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 I&E Notice 79-22 I&E Notice 79-22 was issued by the NRC on September 14, 1979. This notice informed all Operating LWR's that certain non-safety grade equipment, if subjected to an adverse environment from high energy line breaks, could impact the design basis safety analysis and/or the protective functions performed by safety grade equipment. On September 17, 1979 a letter was issued by the NRC requiring all Operating LWR's to respond to the concerns of I&E Notice 79-22 within 20 days.
Metropolitan Edison Company issued a report on October 17, 1979 in response to the NRC letter. The report identified the non-safety grade control systems whose performance could be affected by adverse environment and committed to a more detailed evaluation of potential ef fects o f Iligh Energy Line Break (IIELB) accidents on non-safety grade control systems.
On November 20, 1979 (GQL 1441) Met Ed issued a subsequent letter indicating that a report of the results of our evaluation would be submitted in a separate report prior to startup of TMI-1.
The attached provides the results of our study that concludes that non-safety grade equipment subjected to an adverse environment resulting f rom high energy line breaks would not impact the safety analysis and the adequacy of the protective f unctions performed by safety grade equipment.
Submission of this l
report completes our response on I&E Notice 79-22 although related studies, including IEB 79-01B are continuing.
Sincerely, r207280144 820721 l
hDRADOCK 05000289 11. D. 'I kill 1
PDR llDil: LWil:vj f Attachment cc:
R. C. Ilaynes g[
R. Jacobs GPU Nuclear is a part of the General Pubhc Uhhties System
Safety and Non Safety Grade System Interaction Report Scope in our letter of October 17,1979, we identified the four non-safety grade control systems listed below as having the potential to undergo adverse environmental ef fects upon exposure f rom an HELB.
Reactor Power Centrol and Shutdown Control Rod Drive Control System Reactor Pressure Control Power Operated Relief Valve Pressurizer Heaters Pressurizer Spray Steam System isolation and Pressure Control Turbine Trip / Turbine Stop Valves Steam Line Isolation Valves Turbine Bypass / Atmosphere Relief Valves Feedwater System Isolation and Control Mein Feedwater Conirol Main Feedwater Isolation ~
Emergency Feedwater isolation Emergency Feedwater Initiation Emergency Feedwater Level Control The eval uation of potential l y adverse env i ronmenta l effects on the Emergency Feedwater components has already been performed by Met Ed in response to l&E Bulletin 79-01B.
The High Energy Line Breaks (HELB) assumed in this report are the licensing basis accidents which have been analyzed in the TMl-1 FSAR, Chapter 14 and/or the TMI-1 Restart Report, Chapter 8.
Thesc include the following:
1.
Large Main Steam Line Break inside the Reactor Building.
2.
Large Main Steam Line Break outside the Reactor Building.
3.
Large Feedwater Line Break inside the Reactor Building.
4.
Large Feedwater Line Break outside the Reactor Building.
5.
Lerge Loss of Coolant Accident.
6.
Small Loss of Coolant Accident.
O For TMI-1 the buildings in which an adverse environment would occur following a HELB would be as follows:
The Reactor Building (LOCA, MSLB or FWLB)
The Intermediate Building (MSLB or FWLB)
The Turbine Building (MSLB or FWLB)
Principal components and their associated trains of controi and power subcomponents were examined to determine the ef fects of an adverse environment on their performance.
In this analysis, attention was given to the acceptability or unacceptability of the resulting performance of the " principal" component as it impinged upon the saf ety analysis originally performed for Chapter 14 of the TMI-1 FSAR or for Chapter 8 of the Restart Report.
Attention was given to the service qual i f ication where this was obtainable. Where indications were that such information would delay the studies, assumptions as to the qualification of components were made in the direction of conservatism.
Assumptions As part of the analysis the following assumption were made:
Any systems and components which are classi f ied as Class ~ 1E or o
saf ety grade w i ll, for the purposes of this study, be considered capable of performing its intended functions under the postulated adverse conditions. Action concerning qualification of this equipment, is covered by I&E Bulletin in 79-01B.
o
.All CRD mechanisms are typical.
o Components in the Reactor Bui i d ing, Intermediate Suiiding, and Turbine Building only will be subject to an adverse environment.
Evaluation Reactor Power Control and Shutdown Control s All non-safety grade controi rod control and power supply subcomponents, which are located within potential areas of adverse, environment, are qualified to withstand aaverse environment or t h i. y are assumed to f all.
The latter category consists of the controi rod drive motors and their cable connectors for which there is no documentation to support qualification.
Failure of these subcomponents wili result in interruption of power to the control rod drive motors.
A reactor trip occurs whenever power has been removed from the control rod drive motors.
The TMI-1 design provides two stored energy breakers which do not require power to interrupt the electrical feed to the rod drive control power supplies and a second set of circuit interrupting devices in' series with the output of the power supplies.
All devices
have interrupting capacity of suf ficient rating to open under any group load configuration.
Reactor trip is f urther assured by providing series trip devices, split buses and provisions for periodic testing.
Trip redundancy is provided by series breakers while availability and testability are provided through dual power sources.
(TMl-1 FSAR 7.2.2.3.1)
The reactor trip system is covered by our response to I&E Bulletin 79-01B.
Functioning of the saf ety grade trip mechanisms to -insert the CRD's will occur prior to any HELB induced environmental damage to the non-saf t ry grade CRD motors or the cable connectors.
In the unlikely event that an undefined malfunction were to cause a spurious withdrawal of all CRD's, the rapid tr ip due to the resultant high neutron flux would occur prior to any environmentally induced damage to the safety g ade trip mechanisms.
The study, therefore, demonstrates that reactivity control in accordance with the safety analysis is ensured in all postulated HELB's.
Reactor Pressure Control The ultimate RCS over pressure control during a HELB will be accomplished passively by the spring loaded pressurizer safety valves, The setpoints for PORV lift and high pressure reactor trip are at 2450 psig.and 2300 psig respectively. Therefore, the PORV is not expected to operate following a LOCA or a MSLB.
In the scenario after a FWLB where the PORV can be actuated and potentially remain open, creating a LOCA, analysis have been performed to demonstrate that the ensuing transients can be safely mitigated (as defined by 10 CFR Part 50, Paragraph 50.46(b)) by the ECCS.
These analyses include a small break LOCA occurring simultaneously with loss of all feedwater and, therefore, a FWLB is bounded by them.
Further, the Control Room Operator, observing installed and improved PORV sensing and position indicators, has remote manual control to close the environmentally qualified PORV block valve installed upstream of the PORV, thereby limiting the RCS discharge.
Documentation is not available to support the environmental qualification of the non-safety grade actuators for the pressurizer spray valves, nor to support the qualification of the normal pressurizer heater power dis-tribution centers. However, due to their prolonged time constants, neither the pressurizer sprays nor the pressurizer heaters will significantly affect the transients following a postulated HELB.
The study, thus, concludes that none of the non-safety systems or components which are part of the Reactor Pressure Controls will adversely af fect the transients following a MSLB, FWLB or a LOCA.
Steam System Isolation and Pressure Control i.
~
The steam isolation and pressure control valves are qualified to with-
~
stand HELB induced adverse environmental conditions with the exception of associated actuation and/or control equipment.
High pressure control for the secondary system will rely upon the Main Steam System spring loaded safety valves. Therefore, taifure of the Turbine Bypass Valves (TBV's) and Atmospheric Dump Valves (ADV's) to open upon control command would not effect the safety analysis.
Failure of the TBV's and ADV's in the open position would result in an uncontrolled blowdown of both steam generators. However, this failure is bounded by the TMI-l Restart Report section 8.3.9.
For this FSAR accident, double ended ruptures of all steam lines was assumed.
The conclusions of this analysis were:
- 1) that "both steam generators" would blowdown in forty-three (43) seconds af ter the ruptures, 2) feed-water would be isolated, terminating the accident, and 3) the core does not return to criticality."
"An equiiibrlum reactor system cooldown and cepressurization is achieved by operator controlled emergency feedwater flow with steam relief out of the steam line breaks."
In view of the fact that the main steam isolation valves (MSIV's) are normslly open during plant operation and that these valves are remote-manualIy controlled, these valves do not function during any analyzed accident transients. Failure of these valves in any position, therefore, will fall within the bounds of the accident described above.
Failure of the electro-hydraulic actuators for the Turbine Stop Valves (SV-1 thru 4) will cause these valves to trip (TMI-1, FSAR 10.3.1).
Therefore, the safety analysis will not be affected.
l The study concludes that possible failure of any of the actuators i
and/or controllers of the steam system valves due to HELB induced I
environmental conditions will not present unacceptab le secondary system pressure control conditions.
The Restart Report Section e.3.Y worst cr.se" analysis discussed above ef f ectively bounds any failure of the TBV's or ADV's.
Feedwater I solation and Control The Main Feedwater and Startup Feedwater block valves to Steam Generator B may be subject to adverse environmental conditions in the event of a MSLB.
Ultimate level control of the secondary system to provide a secondary system heat sink will rely upon the Emergency Feedwater System af ter isolation of the Main Feedwater to both OTSG's.
Therefore, the portions of the Emergency Feedwater System which are in the Intermediate Building and the Main Feedwater block valves are required to be environmentally qualified.
The adequacy of the environmental qualifications and any corrective actions required will be taken as part of GPUN's response to I&E Bulletin 79-01B.
Conclusion The necessary systems will perform their functions during the accidents considered by virtue of location and/or environmental 1
qual i f ication, and will not result in an adverse ef fect on the safety analysis.
.The HELB interactions on non-safety grade control systems during accidents are such that plant response is bounded by the safety
~~ analyses.
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