ML20153G752
| ML20153G752 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/31/1988 |
| From: | Gridley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8809080364 | |
| Download: ML20153G752 (5) | |
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TENNESSEE VALLF.Y AUTHORITY CHATTANOOoA. TENNESSEE 37401 SN 157B Lookout Place AU8 311988 U.S. Nuclear Regulatory Comission ATTN: Document Cot.*.rol Desk Washington, D.C.
20555 centlemen:
In the Matter of
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Docket Nos. 50-327 Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SQN) UNITS 1 AND 2 - REACTOR COOLANT SYSTEM (RCS)
POSTTRIP COOLDOWN AND SHUTDOWN MARGIN (SDM)
As requested by letter from Suzanne Black dated August 19, 1988, enclosed is our submittal on the above subject.
Included in the submittal, ene.losure 1, is a description of the subject phenomenon, its probable cause, the effect on SDM, and the immediate correctite actions that have been implemented for operation of SQN units 1 and 2.
The submittal also contains a discussion of the long-term corrective actions that are presently under invostigation.
Summary statements of tt.e ca nitment contained in this submittal are provided in enclosuro 2.
If you have any questions concerning this submittal, please telephone M. A. Cooper at (615) 870-6549.
Very ttvly yours, TENN S E VAf.T Y AUTHORITY s
R. Gridt f, a.s *ar Nuclea-L4.cattvin'. and Regui ;*ty m t-tes 4
g Enclosures i
ect See page 2 i
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An Equal Opportunity Employer
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U.S. Nuclear Regulatory Comtesion i
cc (Ruclosures):
Ms. S. C. Black, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Mr. F. R. McCoy, Assistant DLeector for Inspection Progrant TVA Projects Division U.S. Nuclear Regulatory Commission Region 11 101 Marletta Street, NW, Suite 2900 Atlanta, Georgia 30323 Sequoyah Resi6ent Inspector Sequoyah Nuc. Lear Plant 2600 Isou Ferry Road Soddy Dalsy. Tennoesee 37379 l
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4' Enclouure 1 Der,criotion if problem on June 14 1988, TVA identified un issue associated with cooldowns fciloteing 3
a reactor trip.
Both units 1 and 2 are affected. Following a reactor trip.
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cooldown below the design no-load temperstyre of $47 dogrees Fahrenheit (F).
the curre..t plant configuration results in a reactor coolant system (RCS) l Operational history indicates that the RCS will generally cool uown to approximately S20 degrees F following a reactor trip from full power. Tha safety issue associated with this cooldown is in the area of suutdown margin (SDH). The design core SnH assumes that, on a reactor trip, the RCS stabilizes at S47 degrees F and, as a result, does not consider the input of pocitive reactivity because of a subsequent posttrip cooldoen. Therefore, i
conditions cou1<* exist that would result in a f el'.ure to rnintain t'io
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1,600-peiront mLile (pem) technical specification (TS) SDM r.,quirement if, j
- past occurs.
under certain core conditic ns, a,cooldown of the magnitudo experf enced in Llw I
Corrective Actions For Continued Operation When this iscue was identlfied, the following itnrudiate correctivo acticas l
were taken to ensure continued operation of unit 2 within denigr. basis assumptions for the remainder of cycle 3 and to provide resolus.lon to support I
restart of unit 1:
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Westinghorse Electtic Corporation was contacted to det'emine the maximum allowable cooldawn for which the TS 1,600-pc.m SDH requirement would bo maintained. The Westinghouse analysis provided the mininum e.ooldown terperature as a functL7n of burnup under various pretrip condLtions.
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SQH Erergency Instsuction ES-0.1, "Reactor Trip Responso," was revised to require the operation *: crew to initiate a manual RCS boracion tf average RCS terperature drops below the alnituun allowable comperature s:rpp1 Lod by l
Westinghouse. The intent. of this boration.8s to ensure that actions aro l
taken consistent with TS tequirements to mitigat$ the potentici loss of l
SDH su a e alt of the posttrip cooldown.
1 The imediate corrective actiot.s ensure that the operator takes manual action in the event of a posttrip cooldown to maintain validity of Final Safety Analysis Report (FSAR) accumpt!ons concerning SDH for a steam line break design basis event. This act.lun resolves the irrcediate safety issue.
The corrective actions constitute a compensator / measure and have been evsduated in accordanca kith Administrative Insttvetton 49, "Control and Tracking of Cowpensatory Heasures." ihe long-term corrective actions w8.11 resolve this compensatory measure, i.on&-Tet
'orraetive Action LTo HitIg*te posttrip (ooldowna s
TVA believec the rostttip cooldown to be the result of the combination of an excessive mass locu from the steam generators via the steam dump system and I
miscellaneous secondary-side steam leake. The RCS overcooling is further conpounded by the refilling of the steam generators with water from the condensate storar.e tank, which is typically 70 degrees F, via t'io auxL11ary tM A.
l feedwater (AFW) systen. The SQN AFW system is declgned with an automatic steam generator level control system. AFW to each steam generator is controlled to automaticelly maintain no-load steam generator levels. On a reactor trip, two 100-percent-capacity, motor driven AFW pumpf and one a
200-percent-capacitw, turbine-delven AFW pump start and deliver full flow
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until steam generator levels are returned to 33-percent narrow range level.
TVA is presently investigating options to roduce the mass loss from the steam canerators. This reduction in nass loss following a reactor trip would have a twofold effect. First, the initial RCS cooldown from at-power tenditions would be slowed and it.s magnitude reduced.
Secondly, because less mass would be removed from thw steam generators, the amount of AFW that would be required to regain steam generator level would be reduced, thereby further llmLting the RCS cocidown.,
To reduce the nass loss from the steam generator following a reactor trip. TVA is investigating various methods of optimizing steam dump operation. These methods are currently being analyzed by Westit.shouse to determlno their feasibility and effect on RCS cooldown.
The Wantinghouse analyses will also include a study of the sensitivity of the RCS cool 4own to AFW flow rate. The Westinghouse study will also provida a basis for determining if the current automatte AFW 1evel e.ontrol system should J
be defeated by taking manual control following retctor trips. TVA be1 Loves
+. hat autcmetic operation of the AFW 1evel control system is desirable and is first pursuing correctise action to prevent the posttrip cooldown wLthout having to take manual control of the system.
Following a thorough review of the Westinghouse analysis, TVA will determine which plant modifications and/or procedure revisions will bent serve to mitigate the posttelp cooldown.
Any require] plant modi 1 cations and/or procedure revisions will be initiated and implemented in an expedient and controlled manner, j_
To further limit rass loss fra1 the steam sonorators, TVA is providing the appropelate management attention to ensure that steam leaks, which could contribute to the QCS cooldown, are adequately controlled by the existing maintesance program. Sources of steam teaks idantifled curing unit 2 operation are bLing evaluated for applicability to unit 1.
conctonion TVA is preser.tly investigating sergral dif ferent options to resolve the posttrip cooldown b91ng experienced at SQN. Following completion of the review of these options TVA wlli determine appropriate long-term resolution.
4 TVA will provide NRC with a status of our investigation results and plan for resolution implementation by October 14, 1988.
In the interim, continued operation of unit 2 or restart of unit I will not result. in any ufety concerns as a result. of the cooldown because of procedural controls presently in place that require manual boration of the RCS in tho event of an excessive posttrip cooldown.
This boration requiremos.t will ensure that actluns are taken consistent with TS requirements to mitigato the potential loss of SDM because of posttrip cooldowns.
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Encicsuro 2 List of Cossnitmnts 1.
TVA is presently investigating several different options to resolve the posttrip cooldown being experienced at SQW. Following completion of the review of these options. TVA will determine appropriate long-term resolution. TVA will provide NRC with a status of our investigation results and plan for resolution implementatlon by October 14, 1988.
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