ML20099J849

Forwards Response to NRC 920720 Request for Addl Info on FSAR Chapters 4 & 5,Amends 83 & 84

ML20099J849
Person / Time
Site:	Comanche Peak
Issue date:	08/19/1992
From:	William Cahill, Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-92397, NUDOCS 9208240029
Download: ML20099J849 (8)

Text

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Log # TXX-92397 File # 904 915 r

x TUELECTRIC August 19. 1992 William J. Cahllt, Jr.

(iro,,ener tresident U. S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) UNIT 2 DOCKET NO. 50-446 REQUEST FOR ADDITIONAL INFORMATION ON CPSES FINAL SAFETY Af4ALYSIS REPORT (FSAR) CHAPTERS 4 AND 15, AMENDMENTS 83 AND 84 REF:

(1)

NRC Letter from Mr. Brian E. Holian to Mr. William J. Cahill Jr., dated July 20, 1992 (2)

Comanche Peak Steam Electric Station, Unit 1 - Amendment No. 10 to Facility Operating License No. NPF-87. dated June 8, 1992 Gentlemen:

TV Electric's responses to the nine questions in reference 1 are attached.

Two additional questions, which were asked by Mr. Tai Huang at the meeting in Rockville. Maryland on June 4, 1992, are also included as questions 10 and 11.

Should clarification or additional information be required to enable the NRC Staff to com;,iete its t eview, please call David Bize at (214) 812-8879.

Sincerely.

William J. Cahill, Jr.

By:

(1/#

D. R. Woodlan Docket Licensing Manager DNB/dnb Attachment Enclesure c - Mr. J.

L. Milhoan. Region IV Resident Inspectors, CPSES (2)

Mr. T.

A. Bergman, NRR Mr. B. E. Holiar, NRR 92002/'00p9 920819

1. 4 1

ADOCK 05000446

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PDR PDR l;

A30 N, Ouve Siteet L.B. 81 DaHas. Texas 75201

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' Attachment to1TXX-92397 Page 1. of: 6 ADDITIONAL INFORMATION REGARDING

-CHAPTERS 4 AND 15 0F FINAL SAFETY ANALYSIS REPORT COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 1.

Question-Section 4.2.2.3 of. Chapter 4 of the FSAR. page 4.2-21, mentions the desirability of a negative moderator coefficient when greater than 75 percent 9f full power.-

However, Figure 15.0.6 shows a positive value up-to 100 percent of full power.

Explain this discrepancy.

'Responsg Section 4.1.2.3 of the FSAR will be revised in Amendment 86, consistent j

with Amendment 6 OL Technical. Specifications, to reflect that the moderator temperature coefficient may be positive at power levels below

-100%-Rated Thermal Power

-2, Duestion; Use of hafnium as the absorber material in the control rods is mentioned throughout Chapter 4..

NRC Information Notice No. 89 31, " Swelling and-Cracking of Hafnium Control Rods," alerted PWR licensees of swelling _ and 1 cracking of_-hafnium control: rods at several PWRs.

Did you consider this

-information-in your-application of hafnium as a control rod material for Unit 27 Resoonse Ag-In-Cd _ control rods are used in CPSES'as shown in Table 4.1-1B and stated on page 4~2-12.

Section 4.2.2.3.1 of the FSAR will be revised in-Amendment 86 to reflect that Ag-In Cd _ alloy is the primary design and-

. hafnium,- the alternate design.

3.

' Question-i Section:4.3.2.2;8 of Chapter'4 of the FSAR specifies that tests performed-at'the beginning of each reload cycle are limited to

. verification of steady state power _dist-ibutions.

Explain'why control-rod worth measurements *and moderator temperature. coefficient surveillance areinotfalso performed at this time.

Rgsoonse Section 4.3.2.-2.8 was revised in Amendment 85 to acknowledge the-tests

_ erformed at the beginning of each reload cycle.

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h Attachment to.TxX-52397

. Page 2 ofE6 4.-

Question Section 4.3.2.~6fof Chapter 4 of the FSAR refers to the use of the LEOPARD and Pu0 computer codes for fuel storage r;'ticality calculations. _NRC Information Notice 92-21, ' Spent Fuel Pool Reactivity Calculations,* indicates inaccuracies discovered in the use of these codes to predict 'the criticality in fuel storage racks.

Did you consider-this information on potential computer code inaccuracies in rclation to your Unit 2 fuel storage analyses?

BesDonse TV Electric has obtained confirmation from Westinghouse that the

. enrichment limits set with LEOPARD and PD0 contain substantial conservatism and that-higher enrichment. limits can be justified using

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more sophisticated analysis.

IN-92-21 was mainly concerned with the use l'

of Boraflex, which is not employad at CPSES.

S.

Question The first-footnote to Table 4.3-[3]B of Chapter 4 of the FSAR for Unit 2 refers to a value which includes a 0.1 percent delta-rho uncertainty.

What value is being referred to?

Resoonse The footnote applies to the Doppler defect.

The omission will be corrected in-FSAR Amendment 86.

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Question

-Table 4.344 of Chapter-4 of the FSAR which is supposed to summarize the comparisons of _ criticality calculations with 101 critical experiments is missing.

Resoonse (Table _4.3-4 -is still' listed as an ef fective page_ in the current FSAR.

A O

copy of _the table le encloseW.

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1 Attachment to TXX-92397 Page 3 of 6 7.

Question-Explain why the control rod drop time has decreased to 2.4 seconds for Unit 2 compared to 3.3 seconds for Unit 1.

Ellp_q_njg Current Comanche Ptak Units 1 and 2 Combined Techr.ical Specifications (Proof and Review copy for Unit 2) require the measured control rod drop time to be less than or equal to 2.4 seconds.

In some analyses for Unit 1. Westinghouse used a control rod drop time of 3.33 seconds for the FSAR non-LOCA analyses assu'.iing that the control rods at CPSES would be B,C control rods.

The B,C control rods were never used and were replaced by Ag-in-Cd control rods which have a f aster drop time than the B,C control rods.

Ag In-Cd control rods are also installed for Unit 2.

Thus, any control rod drop time greater than or equal to 2.4 seconds is acceptable for use in the Unit 2 analyses.

8.

Question The analysis for the uncontrolled rod cluster control assembly bank withdrawal from a cubcritical or low oower startup condition [ assumes two reactor coolant pumps to be in optration.

Technical] Specifications allow fewer than two pumps to be in operation during shutdown.

The analysis should be performed from flow conditions correspending to the minimum number of allowable operating pumps.

Resoonse 7

The FSAR presents an analysis for the uncontrolled rod cluster control assembly bank withdrawal event in Mode 2.

An occurrence in Modes 3, 4 or 5 with two or more reactor coolant pumps in operation would be bounded by the analysis in Mode 2.

This is based upon the FSAR analysis assumption that reactar trip does not occur until the power-range (low setting) high neutron flux setpoint is reached and that two banks are withdrawn sequentially at maximum speed (72 step / min).

These conservative assumptions result in the core returning to critical and generating some power prior to trip.

Therefore, the primary system flow rate becomes an important consideration as a factor in DNB evaluation.

(Note that in Mode 3, the Technical Specification 3.4.1.2 requires that at least two reactor coolant pumps to be in operation whenever the reactor trip breakers are closed.)

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-' Attachment to TXX 92397

'Page 4 of 6 In Modes 3, 4 and's, the source range high neutron flux trip will be asailable (Technical Specification 3.3.1 Table 3.3-1) to terminate the event by tripping-any. withdrawn'and withdrawing rods before any significant power level could be attained.

Also, the reactivity insertion rate would be slower because a f ailure in the rod control system could cause, at most, the withdrawal of only-one bank and its

c withdrawal rate'would be slawer than the maximum rod speed which is possible when -in automatic rod control.

Under these conditions, DNB (and fuel: failure) would not be credible.

9.

Qutstion Recent nonconservatisms were identified at Comanche Peak related to the input assumptions and boundary conditions (inverse count rate--ratio data and flux-multiplication setpoint).in the analyses of the licensing basis boron dilution event. -Based on this. justify the automatic actions to terminate the dilution"and start boration which were assumed in the boron diiution analyses for Unit 2.

ResDonse :

The'CPSES Unit 12 licensing analysis for the Boron Dilution event has been completed and is consistent with the-analysis supporting a License Amendment (reference 2) whi'ch was granted for CPSES Unit-1.

The License Amendment temporarily removes the_ operability requirement for the Boron Dilution: Mitigation, System (B0HS.) f rom the Technical Speci fications.

CPSES Unit 1 BDMS operability will not be required'until six months following-the'second refueling outage for CPSES Unit 1.

Similarly, a six_. month evaluation period of the BDMS following initial criticality hast een requested for the-draft Unit 2 Technical Specifications.

The b

~ basis for the License Amendment-was-a revised Boron Dilution analysis

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.and additional-administrative controls.

-The revised Baron--Dilution analysis for Unit 1 and Unit 2 does not

--credit automatic actions of the BDMS.

The licensing. analyses for Modes 3 : 4 and -5; demonstrate that there are 'at.least 15 minutes f rom' the -

initiation of-a boron dilution betree shutdown margin is lost.

These m

analyses. provide reasonable. confidence that the reactor operitors have-sufficient' time during performance of their routine duties to identify

'and mitigate;a boron dilution event.

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Altachment to TXX-92397

Page 5 of.6 In addition to'the revised analyses, the following compensatory action.;

are specified for-the duration of the temporary Technical Specification revision:

1)

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of entry into MODES 3, 4.-or 5 from MODES 1, 2, or y

6, (and once per_every 14 days thereafter while in MODFS 3, 4, or 5). TV Electric will verify (unless startup is in progress) that.

ett'er valve CS-8455 or valves CS-8560. FCV-111B, CS 8439 CS-84'.1, and CS-8453 are closed and secured in position: or 2).

-Following_ entry into MODES 3, 4, or 5 from MODES 1. 2, or 6, each crew of the Contror Room Staff will receive a briefing to discuss the type of reactivity changes that could occur during a oilution event; the-indication of a dilution event: and the actions

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required to stop dilution, commence immediate boration and i'

establish the required shutdown margin.

For extended shutdowns, this briefing will be repeated for each crew prior to resumption of control room. duties following an off duty period which exceeds 7 days.

During time periods when this option is used. the sourca

. range will be monitored for indication of unexplained increasing counts and1 inadvertent boron dilution every fi. teen (15) minutes.

In addition, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of entering MODE 5. TV Electric will ensure that on?y one Reactor Makeup Water Pump (dilution source)'

is a'igned to the supply header; Even though credit is not taken for the BDMS, its use during plant n

operation provides additional assurance that an inadvertent dilution event will be detected and mitigated prior to a retarn to 'ritical.

In c

i addition, other alarms an1 indications (as provided in Section 15.4.6.1

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of-the CPSES FSAR)lare available to the operator which allow for the -

detection of an inadvertent boron dilution.

In view of'these alarms-and indications, tcgether with'the procedures, training, and activities.previously mentioned, reasonable assurance has 1been provided to rninimize the' likelihood of an inadvertent boron Edilutior, event during the time interval for1the tamporary TS revisions.

Should such an-event occur, thase actions provide reasonable assurance of timely detection and mitigat on.

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l A'ttachment to TAX-92397 Page 6 of 6 10.

Question When comparing Figures 4.3-22A and 4.3-22B, why is the distribution of peak linear power, kw/ft, shifted about the point of deltc I - 07

Response

In Figure 4.3-22A, Unit 1, some of the data for the end-of-life load follow cases were inadvertently omitted.

If those points are included, then rigure 4.3 22A will look similar to Figure 4.3-22B which is correct for Unit 2.

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11. Question BOL and E0L Doppler temperature coef ficient curves, Figt.res 4.3-27A and 4.3 27B, are different.

What differences between 0FA and standard fuel design or Operating characteristics cause the differences in the shape and orientation of the curves?

Response

The Unit 1 curve. Figure 4.3-27A, was gcnerated with an earlier

.c methodology than the Unit 2 curve.

If the curves for both Units are calculate) with a spatial weighting factor applied, they exhibit similar characteristics.

Incidentally, Figur? 4.3-27B will be updated to reflect CPSES Unit 2 specific results, with 3-D spatial weighting to be included.

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_Page.1 of_1 T' '

CPSES/FSAR TABLE 4.3-4 BENCHMARK CRITICAL EXPERIMENTS Description of Number of LEOPARD k Using eff Experiments

• Experiments Experimental Bucklinas UO 2 Al clad 14 1.0012 SS clad 19 0.9963 Borated H O 7

0.9989 2

Subtotal 40 0.9985 U-Metal Al c'ao 41 0.9995 Unclad 20 0.9990 Subtotal 6i 0.9993 Total 101 0.9990

• Reported in Reference [121 i

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