ML19269E713

Supplements 790411 Response to IE Bulletin 79-05B.Discusses Commitments to Anticipatory Reactor Trips.Forwards Rept Indicating Trip on Low Steam Generator Level Would Not Be Anticipatory.W/Proposed Safety Evaluation Program

ML19269E713
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/21/1979
From:	Trimble D ARKANSAS POWER & LIGHT CO.
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
1-059-25, 1-59-25, NUDOCS 7906290513
Download: ML19269E713 (32)

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ARKANSAS POWER & LIGHT COMPANY POST OFFICE box 551 LITTLE ROCK. ARKANSAS 722C3 (501)371-4000 May 21, 1979 1-059-25 Mr. K. V. Seyfrit, Director Office of Inspection & Enforcement U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011

Subject:

Arkansas Nuclear One IJnit 1 Docket No. 50-313 License No. DPR-51 Anticipatory Reactor Trips (File: 1510.1)

Gentl emen:

OJr response to IE Bulletin No.79-05B, dated May 4,1978, committed us to install control-grade, anticipatory reactor trips on loss of main feedwater and/or on a turbine trip, and to upgrade tnese to safety-grade trips commensurate with our letter of May 11,1979. Item 5 of Bulletin 79-05S also requested an anticipatory reactor trip on low steam gene-rator level . Attached is a report on the assessment of these proposed trip signals wnich shows that a trip on low steam gererator level would not te anticipatory. This trip is, therefore, not included in our design.

Appendix A to the attached report is a proposed Safety Evaluation Pro-gram for Anticipatory Trips. Arkansas fower and Light intends to fully complete Phase 1 of this proposed program. A determination will be made at that time of the necessity of continuation of the program. Another determination for continuation will be made u:;on completion of Phase 2, provided Phase 2 is implemented.

Appendix B to the attached report is our proposed design for safety grade anticipatory trips on loss of main feedwater and/or on turbine trip. The design of these trips is submitted for your review as re-quested in Item 5 of Bulletin 79-05B. Following approval of the design of the trips by the staff, we will implement these changes during our next outage (following completion of the design change engineering) to cold shutdown conditions which is of sufficient length to accommodate the change but no later than the next refueling outage.

2148 129

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? I 1-059-25 Mr. K. V. Seyfrit, Director May 21, 1979 In response to Item 7 of IE Bulletin 79-05B, we have previously sub--

mitted two Technical Specification Change Requests for your review and approval. We have no further changes to request as a result of the TMI-2 incident at this time.

Very truly yours, M d.

David C. Trimble Manager, Licensing DCT: ERG:vb Attachment 2148 130

9 f ANTICIPATORY TRIP FUNCTI0 tis FOR 177 FA PLANTS 2148 131

If;DEX 1.0 INTRODUCTIOt1 2.0 ASSESSMENT OF POSSIBLE ANTICIPATORY TRIPS 3.0 FUf,'CTIONAL ANALYSIS

4.0 CONCLUSION

S AND

SUMMARY

5.0 P,EFEP, ENC ES APPENDIX A: SAFETY EVALUATION PROGRAIl FOR A"TICIPATORY TRIPS APPENDIX B: DESIGt! FOR SAFETY GRADE ANTICIPATORY TRIPS 2148 132

1. 0 INTRODUCTION For the purposes of this report, an anticipatory trip is defined as a trip function that would sense the start of a loss of OTSG heat sink and actuate much earlier than presently installed reactor trip signals. Possible anticipatory trip signals indicative of changes in OTSG neat removal are: turbine trip, loss of main feedwater, and low steam generator level.

This report evaluates the effectiveness of anticipatory trips compared to the existing high RC pressure trip for a LOB!. Qualitative and quantitative arguments are presented which support elimination of the level trip in the stean generators from final design considerations of anticipatory trips.

Functional response is presented in terms of a parametric study of time to trip. Thus, irrespective of the plant specific trip signals and actuation time, the hardware design can proceed with greater flexi-bility. That is, by presenting systen parameters, such as pressurizer fill time, as a function of tine to trip, then if one plant's turbine trip signal occurs 2.1 secs after initiation of the event and another plant's trip signal occurs at 2.5 secs, this study will still be applicable to both.

Some of the results presented in this report have already been sub-nitted to the NRC in Reference 1. The analyses are performed with the revised setpoints, i.e., high RC pressure trip at 2300 psig and PORV setpoint at 2450 psig. It is shown that anticipatory trips provide additional margin between the peak RC pressure after the reactor trip and the PORV setpoint, but provide little additional margin in the longer tenn repressuriza tion to the PORV setpoint with continued delay of auxiliary feedwater initiation.

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9 9 2.0 ASSESS!1ENT OF POSSIBLE A!!TICIPATORY TRIPS In accordance with Bulletin 79-05B, item 5, an evaluation for design basis for anticipatory trips on turbine trip, loss of main feed-water, and low steam generator level has been completed.

The evaluation showed low steam generator level not to be antici-patory and, therefore, it has not been reconmended as an anticipatory trip function. Figure 2-1 shows tne OTSG startup level fran site data and the CADDS calculated 0TSG mass inventory as functions of time following the TMI-2 event. The time of reactor trip on high RC pressure is noted on the figure and clearly demonstrates that a stean generator level trip would not have been anticipatory for a level setpoint that would not interfere with normal operations and naneuvers. The initial rapid fall in OTSG level occurs as the turbine stop valves close, momentarily stopping steam flow out of the generators.

The mass inventory increases during this pariod due to the loss of flow friction aP. By the time the reactor trip occurs, at 8 seconds, steam flow is re-established through the bypass system, flow friction aP re-establishes the level and both mass and measured level start to decrease unifonaly. An OTSG level trip set to trip on the initial drop shown in Figure 2-1 would need to be set restrictively high for normal plant maneuvers and/or lower power levels.

Further level information (in tenas of mass inventory) is given in the figures for the analysis in Section 3.0. The results for those cases also indicate that the steam generator low level trip function would not be sufficiently fast to be considered anticioatory.

Anticipatory trips for loss of feedwater and turbine trip can be designed to trip the reactor in a care expedient manner than the high 2148 134

RC pressure trip for sone overheating transients. An anticipatory trip will provide more margin to PORV setpoint during the initial overpressuri-za tion resulting fran loss of feedwater and/or turbine trip. These trips will provide slightly more time to PORV setpoint and pressurizer fill for delayed auxiliary feedwater initiation conditions.

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Figure 2-1 LOFW (TI.il-2 EVENT) 2.0 t i i 1 160 TRIP ON HIGH RC l.6 I

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3.0 FUNCTIONAL ANALYSIS A series of LOFW evaluations were oerfonned at 100% full power (2772 MWt) with a reactor trip assumed on an anticipatory signal. With the new high RC pressure setpoint of 2300 psig, a reactor trip would be expected at about 8 seconds after the LOFW. The anticipatory trip study considered reactor trips with 0.4 sec, 2.5 sec, and 5 sec delays fran time zero.

These studies also included sensitWities to AFW failure and reactor coolant pump coastdown.

The anticipatory trip study modeled a generic 177 FA plant, and is considered applicable to raised or lowered loop designs. A feedwater coastdown similar to that estimated to have occurred at the fiarch 28th TMI-2 event was used to generate separate heat demands for each CADDS analysis. The heat demands will change as the reactor trip time is delayed, because the additional heat input will boil off the fixed steam generator inventory at different rates.

For the cases where AFW floa was modeled,1000 gpm was assumed, starting at 40 seconds. With proper stean generator level and pressure control, the systen parameters will begin to stabilize at 195-290 seconds, depending on trip delay time and RCP operation; see Table 3-1 and Figure 3-12. The PORV will not be actuated, n~ would the pressurizer fill or empty.

With the assumption of no AFW, the PORV will be actuated about three ninutes into the event, as a result of systen swell, the pressurizer fills at 10-12 minutes (see Table 3-1). A delay of reactor trip of 2-3 seconds is seen to reduce PORV time to actuate by about one minute, and pressurizer fill by about 2 minutes. For PORV setpoints other than 2450 psig, the times will vary and can be datermined from Figures 3-3, 3-4, 3-8, and 3-10. 2148 137

In each of these cases, the mass addition and cooling effect of expected make-up system operation is not modeled. One make-up pump running rtill add about 10 inches per minute to pressurizer level, and

$1/2% heat demand. It should be noted that the Itay 7 report used a heat denand which reproduced the T11I-2 LOFW event; it has been reported by the operator that two make-up pumps were running from 13 sec into the event, creating a higher heat denand than the anticipatory trip studies of the report assume. The difference is shown in Figure 3-12.

The steam generator heat demands, reactor power, RC system pressure, pressurizer level, and RC inlet / outlet tenperatures are given in Figures 3-1 through 3-5 for the trip at time zero case and Figures 3-7 through 3-11 for the trip on high RC pressure (t=8 secs) case. The effects of delayed auxiliary feedwater initiation are also shown on the high RC pressure trip curves.

2148 138

TABLE 3-1 LOFW EVENT (LOFW at T=0 sec)

TIME OF REACTOR REACTOR AUXILIARY PORV PRESSURIZER S/G LEVEL CONTROL TRIP (" DELAY") C0OLANT PUMPS FEEDWATER OPERATES FULL (400") (P =1025 psig) s tm 0.4 Run at 40 sec - - 195 sec 2.5 Run at 40 sec - - 225 sec '

5.0 Run at 40 sec - - 275 sec 0.4 Run None 235 sec 790 sec -

2.5 Run None 180 sec 685 sec -

5.0 Run None 140 sec 575 sec -

0.4 Coastdown at 40 sec - - 255

0. 4 Coastdown None 190 sec 700 sec -

LOFW EVENT - TIME =0 sec REACTOR TRIP AT 2300 PSIG TIME OF TRIP _RCP AFW PORV PRESS. FULL S/G LEVEL CONT.

8. 0 Run at 40 sec - - 260 sec 8.0 Run None 175 sec 620 sec - 2148 139

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4. 0 CONCLUSI0tF AND

SUMMARY

A spectrum of delay times, representing anticipatory trips, has been analyzed for the loss of feedwater transient. The spectrum included trips at time zero, with a 0.4 second instrument delay, up to high RC pressure trip at time 8.0 seconds. Since a high RC pressure trip occurs very soon after a loss of heat sink (overpressurization) transient from 100% FP, only turbine trip and direct loss of feedwater detection trips would be considered anticipatory.

For all trips considered, including high RC pressure, the PORV is not actuated when normal systen operations occur. The pressure rise in the primary side is less for the anticipatory trips providing additional margin to PORV lift. If auxiliary feedwater is significantly delayed, then an anticipatory trip will, at best, provide about 1 minute addi-tional time to PORV iift and about 3 minutes additional time to filling of the pressurizer. These results can be seen in Table 4-1 which shows the sequence of events for a LOFW transient with trip on high RC pressure (2300 psig) and trip at tine zero.

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TABLE 4-1. LOFW-SEQUEfCE OF LOW EVENTS COMPARISON 40-s 120-s TRIP AT ZERO, EVENT A Fl. DELAY NO AFW NO AFW Loss of feedwater initiated 0 0 0 0 (trip occurs)

(0.4 delay)

High-pressure trip (2300 psig) 8 8 8 a PORV opens (2450 psig) a a 175 235 Peak RCS pressure 10 10 175 235 Pressurizer full a a 620 790 Does not occur for these cases 1

2148 153

5. 0 References

1) Babcock and Wilcox Report entitled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" dated May 7,1979.

2148 154

APPENDIX A SAFETY EVALUATION PROGRAM FOR ANTICIPATORY TRIPS The following constitutes a proposed safety evaluation program for the 177 FA plants with anticipatory trips installed as part of the RPS.

The scope of the study is to provide the necessary review of all transients.

A comolete safety evaluation of the operating plants is complicated by the numerous licensing requirement changes that have occurred since the FSAR's were completed. The following work scope is predicated on a review and analysis which meets current regulatory fcnnat and requirements as applicable. If current designs or equipment do not meet all of the requirements or standards, these will be noted.

The program allows for changes to be incorporated in the design, prior to plant unique analyses being co.upleted.

The proposed safety evaluation prosides a logical sequence of review, general analysis to limit scope, and plant specific evaluations, as follows:

Phase 1: Safety Evaluation Review The transients will be revieaed for possible impact by anticipatory trips in tenns of the following categories of events:

- Increases in Heat Removal by the Secondary System

- Decrease in Heat Removal by the Secondary System

- Decrease in Reactor Coolant Syste.n Flowrate

- Reactivity and Power Distribution Anomolies

- Increase in Reactor Coolant Inventory

- Decrease in Reactor Coolant Inventory 2148 155

Each transient within the above category will be discussed in terms of what normal trip function occurs (and when), what type of signal would be anticipatory, and what vauld be a qualitative assessment of the impact of the anticipatory trip. This assessment would be as generic as possible, with plant specific characteristics mentioned where they would impact the conclusions.

Phase 2: Generic Screening Process Phase 1 of this evaluation will identify those transients that may be adversely impacted by an anticipatory trip. The purpose of this phase of the program is to assess quantitatively the magnitude of impact. This process, to be done generically, will possibly allow elimination of those transients that are impacted but to a small or negligible degree.

Phase 3: Analytical Evaluation The exact scope will De dependant on the results fror, Phases 1 and

2. It is envisioned that for those transients that are significantly impacted by anticipatory trips, each plant specific design will need analytical evaluation. This analysis will account for what signals have been installed which may vary frcc. plant to plant. The goal will be to show that, although adversely affected, the transient results meet the acceptance criteria.

2148 156

T User a Group Approval /NRC gm Requirements NRC NRC N g'"

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APPEliDIX B DESIGN FOR SAFETY GRADE ANTICIPATORY TRIPS The following descrioes the implementation of safety-grade reactor trips into the RPS-I for loss of main feedwater turbine trip.

Loss of Main Feedvater Trip - Control oil pressure switches on both main feedwater pumps will input an open indication to the RPS on feedwater pump trip. Contact buffers in the RPS will sense the contact inputs and initiate an RPS trip when both pu:nps have tripped. This trip will be bypassed below a predetemined flux level, typically 20% FP. Reference Figure 1.

Turbine Trio - Contact outputs fron the main turbine electro-hydraulic control unit will input an open indication to the RPS on turbine trip.

Contact buffers in the RPS will sense the contact inputs and initiate an RPS trip when a turbine trip is indicated. This trip will be bypassed beloa a predeterminej flux level, typically 20S FP. Reference Figure 2.

Figure 1 is a simplified drawing of the main feedwater puelp trip.

Figure 2 is a simplified drawing of the turbine trip.

Drawing 510790GB-1 shows the generic logic for the new trips.

Drawing 51079MLG-1 is a legend for the generic logic drawing.

2148 158

iYPICAL RPS CHANNEL EX ISilliG .

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