TXX-4160, Forwards Revised Pages to Fsar,Correcting Inconsistencies Re Steam Generator Water Level low-low Setpoint,Per SER Confirmatory Issues 7 & 8.Revs Will Be Included in Amend 49 to FSAR

Forwards Revised Pages to Fsar,Correcting Inconsistencies Re Steam Generator Water Level low-low Setpoint,Per SER Confirmatory Issues 7 & 8.Revs Will Be Included in Amend 49 to FSAR

ML20084L276
Person / Time
Site:	Comanche Peak
Issue date:	05/08/1984
From:	Schmidt H TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	Youngblood B Office of Nuclear Reactor Regulation
References
TXX-4160, NUDOCS 8405140410
Download: ML20084L276 (9)

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Log # TXX-4160 TEXAS UTILITIES GENERATING COMPANY Fi1e # 10010

,,nywar Towen . .oo xourn ouve ~1aser, i..n. ,o . nau.4,.,1xxas mo, 907.2 May 8, 1984 Director of Nuclear Reactor Regulation Attention: Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446 STEAM GENERATOR LEVEL, LOW-LOW SETPOINTS REF: B. J. Youngblood to R. J. Gary letter of March 22, 1984, entitled " Resolution of SER Confirmatory Issues (7) and (8)

Pertaining to the Comanche Peak Steam Electric Station (Units 1 and 2)"

1

Dear Mr. Youngblood:

In response to the referenced letter, the Comanche Peak Steam Electric Station (CPSES) FSAR has been revised to correct previous inconsistencies with respect to the Steam Generator Water Level low-low setpoint. The revisions are attached and affect Sections 15.0 and 15.2 of the CPSES FSAR and the response to NRC question 032.103. These revisions will be included in Amendment 49 to the CPSES FSAR.

Please note that the specific values for the Steam Generator Water Level Low-Low Trip Setpoints for Unit 1 and 2 are no longer provided in the response to question 032.103. This response now refers to the Westinghouse setpoint methodology study where these setpoints were calculated (along with the available margins) for review as part of the CPSES technical specifications. Therefore, the discussions relating to margin of safety and the specific trip setpoint values should be deleted from the NRC Staff's evaluation findings with respect to CPSES SER Confirmatory Issue (7) (as enclosed in the NRC Staff letter referenced above).

Sincerely, H. C. Schmidt DRW/grr Attachment Distribution: Original + 40 copies 8405140410 840500 V PDR ADOCK 05000445 E PDR \g A DEVENDON OF' TERAN VTILITIEN ELECTRIC COMi'ANY

c CPSES/FSAR TABLE 15.0-4 (Sheet 2)

Limiting Trip Trip Point Assumed Time Delays Function In Analysis (Seconds) ,

Low reactor coolant flow 87% loop flow 1.0 (from loop flew detectors)

Undervoltage trip 68% nominal 1.5 +

Turbine trip Not applicable 2.0 l

Low-low steam generator 34.6%* (Unit 1) and 2.0 t level 0% (Unit 2) of narrow range level span -

i High steam generator 90% (Unit 1) and 81% 2.0 level trip of the (Unit 2) of narrow 49 feedwater pumps and range level span closure'of feedwater -

system valves, and I

turbine trip The basis for the Unit I limiting setpoint is the loss of Normal Feedwater analysis. The setpoint used in the Feedline Break analysis was assumed to f

be < 2%. I i

el 1

CPSES/FSAR If AC power were not lost for this incident the reactor coolant flow would remain at its normal value and the reactor would trip via the low-low steam generator level trip with no change in DNBR below the value at the start of the transient. The reactor coolant pumps would be manually tripped at some later time to reduce heat addition to the RCS. The Auxiliary Feedwater System has sufficient capacity, even 5 assuming the worst single failure, to preclude filling the pressurizer should the pumps not be tripped.

An additional assumption made for the loss of normal feedwater evalu-ation is that the pressurizer power operated relief valves are assumed

, to function normally. Operation of the valves maintains peak RCS pres-sure at or below the actuation setpoint (2350 pounds per square inch ,

absolute (psia)) throughout the transient. If these valves were assumed not to function, the coolant system pressure during the  :

transient would rise to the actuation point of the pressurizer safety valves (2500 psia). The increased RCS pressure, however, results in less expansion of the coolant and hence more margin to the point where water relief from the pressurizer would occur. ,

i Since the two Comanche Peak units will have different steam generators (see Section 5.4.2), the effect of this difference has been considered in the analysis. Both types of steam generators are integral preheater model s. The major difference, from the standpoint of accident analysis for this event, is the slightly higher secondary side mass as a function of power for the DS (Unit 2) model. In order to maximize the 5 time until reactor trip on low-low steam generator level occurs and to insure that the analysis is valid for both units, the initial steam generator secondary mass was assumed to be 110% of the higher D5 mass.

The low-low steam generator water level trip setpoint was assumed to be

, the same mass (ib. mass) for both units (see Table 15.0-4). Note that 49 while a higher secondary mass (larger heat sink) is, in general, a benefit for primary side heatup events, the assumption of a higher 5

initial mass results in a delay of approximately 4 seconds until the ,;

trip signal occurs, and thus produces a more severe transient.

15.2-23

CPSES/FSAR In addition, all steam generators for both units will be equipped with separate feedwater connections for injection of auxiliary feedwater and main feedwater at low power operation. The major effect of injecting auxiliary feedwater into the upper section of the downcomer is that most of the flow will bypass the preheat region due to the higher  !

resistance to flow in the preheater. This will result in a slight decrease in heat removal capability. However, the auxiliary feedwater injection point is now much closer to the steam generator, resulting in a much smaller volume of hot feedwater which must be purged before the colder auxiliary feed enters the units.

Plant characteristics and initial conditions are further discussed in Section 15.0.3. Plant systems and equipment which are available to  !

mitigate the effects of a loss of normal feedwater accident are dis-cussed in Section 15.0.8 and listed in Table 15.0-6. Normal reactor control systems are not required to function. Power operated pres-surizer relief valves are assumed to function in order to present a more limiting transient, as described above. The Reactor Protection System is required to function following a loss of normal feedwater as analyzed here. The Auxiliary Feedwater System is required to deliver a minimum auxiliary feedwater flow rate. No single active failure will prevent operation of any system required to function. A discussion of ATWT considerations is presented in Reference [2].

Resul ts t

l Figures 15.2-9 and 15.2-10 show the significant plant parameter tran-f sients following a loss of normal feedwater.

l l Following the reactor and turbine trip from full load, the water level in the steam generators will fall due to the reduction of steam genera- ,

tor void fraction and because steam flow through the safety valves con-r 15.2-24 l

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W{s ~ gj .1 CPSEShSAR

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j 3. $_afety Function Setpoints, ,

3 ,,n '

A .' ' Steam Generator Water Level Trip Setpoints The only high-energy line rupture within the containment for which the steam generator water

j. level provides the primary trip function is a secon& ry high energy line rupture from an initial high power condition.

1. 1. In a continuing' effort to monitor protection system operability, Westinghouse has reviewed reported LERs for RPS and ESF trip functions, 3 and when available, plant data on functional

. test results. During this monitoring process, there has been no 1.ndication that the trip functions experience any more additional drif t

' than trip functions with setpoints greater than 5% from the top or bottom of the instrument

. span. As part of the calibration procedures, several, points across the calibrated instrument range incicding the 0% and 100% span values are 49 checked. . Inspection of as left/as found data frem these checks do not indicate any

- significant variances from the instrument range intermediate points. In addition, transmitters

, are purchased such that the upper and lower

. range limits for the transmitters are greater than the upper and lower range values, (See T SAMA Standard PMC 20.1-1973),i.e.,the calibrated instrument range is not the design limit for the device. Based on the above, Westinghouse does not believe that functions which have 95% probsbility setpoints (when

~

1 including uncertainities using the approved '#

032-130 s ,

3

CPSES/FSAR Westinghouse setpoint metholodolgy) with'in 5%

of the top or bottom of the instrument range 49 will respond any differently than any other protection function.

Because large steam generator pressure changes are not expected before trip, only the reference leg heatup effects need be considered, and not the effects of system pressure changes.

The basis for determination of the low-low setpoint is the Loss of Normal Feedwater and Feedline Break events. The setpoints were determined by considering the level used in each of the analyses for each unit.

Unit 1 Unit 2 Loss of Normal 34.6% 0%

Feedwater Feedline Break 2.0% 0%

49 For each unit, the setpoint was determined by considering the following errors for feedline break:

- Normal errors (normal channel accuracy, etc.)

- Post-Accident effects on transmitter (radiation and temperature)

- Reference leg effects (post-accident heatup) 9 032-131

e CPSES/FSAR Only the normal errors were considered for the Loss of Normal Feedwater event since it does not result in an adverse environment.

The errors were calculated using the approved Westinghouse setpoint methodology based on plant specific data for each unit. For Unit 1 the Loss of Normal Feedwater analysis setpoint plus normal errors was determined to be more limiting than the Feedline Break analysis 49 setpoint plus errors (normal, post-accident and reference leg effects). For Unit 2, the Feedline Break analysis setpoint resulted in the limiting setpoint.

At CPSES the reference leg effects have been addressed in the determination of the low-low setpoints. Considered a feedline break and peak compartment temperature of'2270 F, the error due to reference leg heatup will be approximately 5.2%.

The high-high steam generator water level trip is not required for accident situations that 36 could cause significant errors in level indication. The setpoint of this trip will remain unchanged.

Steam generator wide range level does not provide any automatic trip functions.

032-132 e

CPSES/FSAR B. Pressurizer Water Level .

Pressurizer water level provides no trip function following an accident which results in an adverse environment inside containment.

4. OPERATOR ACTIONS A graph depicting the level measurement error due to steam generator reference leg heatup is shown in 36 Figure 032.103-1. Error due to system pressure changes are shown in Figures 032.103-2 and 032.103-3 for the steam generator and pressurizer indicated water level respectively. Plots similar to these will be made available to the operators to ensure they are aware of the potential level measurement errors.

Furthermore, a remote possibility exists that the fluid in the open reference legs may flash to steam in the depressurized steam generators following a secondary high energy line rupture. Therefore to alert the operator to the possibility of erroneous indications, Westinghouse has recommends that the following caution be inserted in all plant emergency instructions for indicated steam generator water

.l evel .

CAUTION The operator should not rely upon steam generator water level indications in any depressurized steam generators following a high energy line rupture inside containment. This is due to the possibility of reference leg '

boiling.

032-132a

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CPSES/FSAR i

The Westinghouse reference Emergency Operating - 36 f

Instructions take the post accident indicated water level errors into account in the specification of the minimum levels required for safety injection terminations.

Similar cautions will be included in the CPSES emergency procedures.

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032-132b

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