ML20039A229

Forwards Revision to Justification Submitted in Util 811103 Request to Delete Startup Test SU-9.5,concerning Rod Group Drop & Plant Trip,From Startup Test Program.Test Deletion Will Not Result in Program Degradation

ML20039A229
Person / Time
Site:	Sequoyah
Issue date:	12/09/1981
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Adensam E Office of Nuclear Reactor Regulation
Shared Package
ML20039A230	List: ML20039A229 ML20039A240 ML20039A247
References
NUDOCS 8112160390
Download: ML20039A229 (4)

Text

i TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374ot 400 Chestnut Street Tower II December 9, 1981 l_ 1 %,

Director of Nuclear Reactor Regulation # ,

Attention: Ms. E. Adensam, Chief 4 Licensing Branch No. 4 Division of Licensing Jff)/

N@p d m U.S. Nuclear Regulatory Conunission Washington, DC 20555 kph. j h ,

Dear Ms. Adensam:

4 s.

In the Matter of ) Docket No. 50-328 Tennessee Valley Authority )

As required by item 2.c(3).a of the Sequoyah Nuclear Plant unit 2 operating ,

license, TVA must have NRC approval before making a " major modification" to )

the initial test program. Startup test SU-9.5, " Rod Group Drop and Plant Trip Test," is required. for unit 2 at the 50-percent power level as part of our initial test program. In my November 3, 1981 letter to you, we requested approval to delete startup test SU-9.5 from our startup test program. As requested by members of your staff and by the Office of ]

Inspection and Enforcement in subsequent telephone conversations, we have revised the justification for our regaest to include additional information. Based on the enclosed information and other information provided to the NRC, by TVA and Westinghouse, during the review of the

" dropped rod" issue, we believe that startup test 3U-9.5 can be deleted without any degradation of the startup program for our Sequoyah Nuclear Plant unit 2.

Please provide us a response to our request as soon as possible in order to delete the test at the appropriate point in our startup test program.

Very truly yours, TENNESSEE VALLEY AUTHORITY

'b 7 L. M. Mills, Manager Nuclear Regulation and Safety Sworn to and , subs ibed before me

'this day o 4/A/1981 ol

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• Notary Public' My Commission Expires cv Enclosure rail 216 0 390k ,

1 An Equal Opportunity Employer

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% M- ENCLOSURE JUSTIFICATION FOR DELETION OF STARTUP TEST SU-9.5

" ROD GROUP DROP AND PLANT TRIP"-

SEQUOYAH NUCLEAR PLANT f

Westinghouse Electric Corporation identified to TVA in November 1979 a ,

concern with regard to certain assumptions employed in the dropped rod safety analysis. The concern came primarily from the potential for an unanalyzed power overshoot while in automatic rod control following selected dropped rod events without immediate reactor trip.

This item was a reportable deficiency under 10 CFR 50.55(e). Westinghouse g recommended plant operating restrictions for reactor controls as an interim measure which would keep our safety analysis valid while a long-term solution was determined.

i The proposed interim solution was discussed at a November 19, 1979 meeting with NRC and involved a change in plant operating procedures. The calculated consequences for this event were dependent upon whether the reactor was being operated in an automatic or manual mode. The concern was limited to reactor operation in the automatic mode. The analysis in the Safety Analysis Reports (SAR) for the rod drop event with the reactor in a manual mode remained valid. This analysis indicated that the DNB limit was not exceeded. If a rod drop event occurred when the reactor was in the automatic mode, the reactor control system would respond to both the reactor power drop (mismatch between turbine power and reactor power) and the decrease in the core average temperature and attempt to restore both quantities to their original values. This restoration of reactor power by the reactor control system might result in some power overshoot depending upon the excore power signal that was used. Therefore, the simple and straightforward way to prevent power overshoot was to either operate in manual rod control or limit the potential overshoot by restricting rod insertion at high power levels.

The proposed change was as follows:

1. In manual mode of reactor control from 0-100-percent power, there is no change from current procedures.

-2. In automatic mode of reactor control from 0-90-percent power, there is no change from current procedures.

3 In automatic mode of reactor control above 90 percent of reactor -

power, control bank D must be withdrawn 215 steps.

By implementing these changes, a dropped rod event during automatic rod control would not result in an overshoot above rated thermal power. For power levels 2:90 percent, a dropped rod event would result in a withdrawal demand from the - rod control system. Since differential rod worth of the D bank while above 215 steps is negligible, the reactivity required for a power overshoot following a rod drop is not available. For rod drops below 90-percent power, analysis has been performed to show that the reactor will not overshoot above rated thermal power and thus the DNB design basis is j met. The above procedures resulted in no overshoot for a dropped rod 1 event. The above procedures were documented by letter dated November 28, j

1979 from Westinghouse to the NRC (attachment 1).

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- 'NRC agreed with the interim operational restrictions and placed operational restrictions in supplement 1 of the Sequoyah Safety Evaluation Report (SER)-

NUREG-0011 with the accompanying statement. The following is an excerpt from the SER.

'l 8

, . The basis for our finding this interim position proposed by

. Westinghouse acceptable is that a dropped rod event will-not result in an overshoot above full rated thermal power. For' power levels equal to or greater than 90 percent, a dropped rod event will result in a -

withdrawal demand from the rod control system. Since differential rod worth of the D bank while above 215 steps is negligible, the reactivity required for a power overshoot following a rod drop is not available. .For rod drops below 90-percent power, -analysis by Westinghouse shows that the reactor will not overshoot above rated j power. Thus, the DNB design limit is not exceeded for the proposed position and, consequently, we find the interim position acceptable.

3 The negative rate reactor trip is intertwined with this issue. It was 4

thought that this trip was needed to prevent an unanalyzed power overshoot, but this was before Westinghouse had instituted the interim operating restrictions or completed its long-term evaluation. Initially, Westinghouse thought its plants with negative rate trip circuitry might not experience a reactor trip as a consequence of rod drop due to a reduction of conservation in the error allowances and the application of a more conservative core physics no6el than previously utilized.

Westinghouse has now completed its long-term evaluation, and in August 1981 notified NRC of its conclusion that, based on a considerable quantity of work, the interim restrictions on operation above 90-percent power could be removed. A meeting with members of the Core Performance Branch of the NRC staff was held in August 1981 for the purpose of presenting the basis for this conclusion. It was also agreed that, based on preliminary review of information to be formally submitted, the staff would be able to issue an interim position which would result in removal of the operation restrictions and the return to normal operation in automatic rod control.

Westinghouse was to provide this information by letter to NRC in November.

I While NRC has not formally reviewed or approved the results of Westinghouse's long-term evaluation, under the-interim operating -

restrictions the negative rate trip is not necessary to prevent unanalyzed power overshoot. Therefore, Westinghouse no longer believes the Rod Drop Plant Trip, SU-9.5, needs to be repeated (see attachment 2). Further, Westind1ouse is confident its long-term evaluation will remove the need for both the operating restrictions and negative rate trip as documented in the November 10, 1981 letter to TVA from Westinghouse (attachment 3). -

The analytical work Westinghouse has done on the rod drop issue provides sufficient justification for the deletion of SU-9.5. Beyond the analytical work, there are additional reasons this startup test need not be repeated for Sequoyah unit 2. These are outlined below.

The objectives of Startup Test SU-9.5, Rod Group Drop and Plant Trip, i.e.,

to confirm that the negative rate trip circuit will' trip the reactor as a result of dropping two rods and to obtain preliminary data for systems response to plant trip before performing the turbine trip and reactor trip from 100 percent, are fulfilled by other testing requirements. Therefore, there is no reason to perform this startup test on Sequoyah unit 2.

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> The negativa rato tripLeircuitry receivss a channel calibration (see IMI "

PRM-CAL Section 5.2.7,~ attachment 4) and a channel function test (see IMI-

.92-PRM-FT, attachment 5) before initial entry into MODE 2 operations. In

. addition, the channel calibration and channel fbnctional test are repeated

_g at 18-month and 1-month intervals-re'spectively. These tests verify the

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operability of the entire negative rate trip circuitry with the exception of the power-range detectors before the production of any reactor power.

The second objective of startup test SU-9.5 is verified in startup test '

SU-1.2. A, Shutdown From Outside Control Room. In this test, the plant is at approximately 30-percent power then tripped from outside the control room. Subsequent to every plant trip, nur operators verify control rod, pressurizer, steam generator, and reactor coolant system status. SU-1.2.A can be performed at a higher power level if NRC still requires additional testing. In addition, verification of the following acceptance criteria of l'

SU-9 5 pertaining to system response can be obtained with SU-1.2. A.

1. All full' length RCCAs shall have released and bottomed.

2. The pressurizer safety valves shall not lift.

3 Steam generator safety valve shall not lift.

4. Safety injection is not initiated.

A new acceptance criterion to startup test SU-1.2A could formally verify-the operability of the entire negative rate circuitry including the power-range detectors. The acceptance criterion would be:

Verify each power range negative rate bistable is tripped.

However, TVA has no plans to revise SU-1.2. A unless SU-9.5 is deleted from the startup program.

In conclusion, even though the Westinghouse long-term evaluation of rod

- drop has not been reviewed and approved by NRC, the repetition of startup test SU-9.5 on Sequoyah unit 2 should not be required since:

I

1. Under the interim operating restrictions on rod controls, reactor trip is not required for multiple dropped rods in order to meet safety analysic design basis.

2. All- the objectives of startup test SU-9.5 are fulfilled by other testing requirements.

However, should Westinghouse determine the need for a negative rate trip to meet our safety analysis design basis without the interim operating restrictions, we will perform a test of this function.

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