ML20236K794
| ML20236K794 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/04/1987 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 5211-87-2142, NUDOCS 8708100005 | |
| Download: ML20236K794 (6) | |
Text
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i GPU Nuclear Corporation
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Ga. u luclear
- ,omgr8o Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number
August 4, 1987 5211-87-2142 l
U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Gentlemen:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. OPR-50 Docket No. 50-289 Response to NRC Request for Additional Information Concerning TSCR No.154 As requested by your letter of June 30, 1987, attached is the additional information concerning the proposed changes contained in Technical Specification Change Request (TSCR) No.154.
Sincerely, D.
<ill Vice President and Director, TMI-l HDH/JCA/ls 5040g/0951A Attachment cc:
R. Conte, NRC.
J. Stolz, NRC W. Russell, NRC G. Edisen, NRC 8708100005 870804 DR ADOCK 05000, 9
\\
hh GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation g
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION CONCERNING THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMI-1) l TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) N0.154 1.
Considering the turbine bypass capabilities and other plant specific features for the TMI Unit 1, will raising the reactor trip on high i
pressure to 2355 psig and raising the arming threshold for the ART on turbine trip to 45% result in more frequent lifting of the first bank or additional banks of the mair. steam safety valves (MSSVs)? Also, will the proposed Technical Specification changes result in more frequent lifting I
of MSSVs than for the original plant design (i.e., 2390 psig high pressure trip, 2255 psig PORV setpoint and no anticipatory reactor trip on turbine trip), Provide the bases for your response.
RESPONSE
l This question involves several subquestions, _.ach one concerning whether the proposed Technical Specification (Tech. Spec.) changes will result in more frequent lif ting of the MSSVs than another combination.
The frequency of reactor trip followed by MSSV lifting depends on the initiating events. These can be generally categorized into reduction of 1
steam generator heat transfer (e.g., a reduction of feedwater flow event) 1 and reduction of steam flow (e.g., turbine trip) events.
Each will be addressed separately.
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(a) RPS at 2355 psig versus 2300 psig for Reduced 0TSG Heat TransferEveny The RPS high pressure trip at 2355 psig corresponds to a higher RCS T
than does a trip at 2300 psig.
Other conditions equal, this avg means more RCS water stored energy needt to be removed by the OTSG through the MSSVs and the Turbine Bypass System which is comprised of the turbine bypass valves (TBVs) and atmospheric dump valves (ADYs).
This will result in either a longer duration blowdown through MSSVs or a lifting of additional MSSVs but will not increase MSSV lift frequency during any one trip.
In addition, the higher RPS trip setpoint will reduce the total number of reactor trips experienced.
Fewer reactor trips without turbine trip will reduce the frequency with which the MSSVs are challenged, and therefore, will reduce the frequency of MSSV lifts.
(b) Anticipatory Reactor Trip ( ART) at 45% Power Versus 20%
for Turbine Trip For the new setting, the reactor will not trip on a turbine trip below 45% power unless the RCS high pressure trip setpoint is exceeded.
The MSSVs will lift in conjunction with Turbine Bypass 1
System cperation while the reactor power is run back by the ICS.
For i
the old setting (20%), the reactor will trip, and the MSSVs will lif t until the primary to secondary temperature differential is reduced.
Although additional MSSVs may lift, the MSSV lifting frequencies will be the same for both ART settings.
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. (c) RPS at 2355 psig and ART at 45% Power (Proposed Tech. Spec. Changes)
Versus RPS at 2390 psig, PORV Opening at 2255 psig and No ART (Original Plant Design) For the original plant desis n, minor s
overheating events were handled by lifting of the PORV without a reactor trip.
Only for major overheating events did the RCS pressure increase to the reactor trip setpoint of 2390 psig.
This corresponds to a higher T,yg and thus to more energy which is removed via MSSV lifting. Also, a turbine trip at high power l
without reactor trip (i.e., no ART) would lift more MSSVs.
The j
higher RCS pressure trip setpoint combined with the lower PORY l
setpoint for the original design would result in fewer reactor trips l
l and associated MSSV lifting.
l In summary, we conclude that the proposed changes will result in fewer challenges to the MSSVs than the current design of the plant. This reduction is realized through a reduction in reactor trips.
While MSSV blowdown time may be extended by implementation of the proposed changes, the frequency of lifting for the MSSVs during any one trip will not increase.The original plant design had an even higher RPS trip setpoint than the proposed setpoint of the TSCR. The original plant design also had a lower Power Operated Relief Valve (PORV) setting than the current or proposed plant design and did not include an anticipatory trip of the reactor on turbine trip as do the current and proposed designs.
The original design was less likely to trip than the current or proposed design, however, the original design would have resulted in more of the MSSVs lifting following a turbine trip from high power.
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l 2.
If the proposed changes will result in more frequent lifting of the MSSVs than for the current conditions or the original design, then state why the proposed changes are acceptable.
RESPONSE
The merits of the proposed changes are a reduced reactor trip frequency and a wider operating region. This has been demonstrated in detail in l
our TSCP. submittal.
A reduction in reactor trip frequency which would i
result in a decrease in the MSSV lifting frequency is significant. Also, the derived benefits from fewer challenges to the plant safety systems l
l provide additional merit for the proposed TSCR. A comparison between proposed and original design is not appropriate because of the reduced i
PORY setting associated with the original design.
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. 3.
Confirm that no credit was taken for the. anticipatory reactor trip on turbine trip in the accident analyses of Chapter 15 of the Final Safety Analysis Report to provide assurance that the ihcrease to 45% reactor power is bounded by these analyses, f
RESPONSE
The TMI-l Updated FSAR Chapter 14 " Accident Analysis" takes no credit for the Anticipatory Reactor Trip on Turbine Trip.
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