ML20039E190

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Forwards Addl Info Requested by NRC Re B-SAR-205.Info Will Be Incorporated Into Amend to B-SAR-205,scheduled for 770831 Submittal
ML20039E190
Person / Time
Site: 05000561
Issue date: 08/08/1977
From: Taylor J
BABCOCK & WILCOX CO.
To: Boyd R
Office of Nuclear Reactor Regulation
Shared Package
ML111090060 List: ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8201060628
Download: ML20039E190 (10)
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Teteptione (90173MS 4 i

DOCKET ST1t 50-561 August 8,1977 Office of tbclear Rector Refplation Attention:

Mr. Roger S. Goyd, Director y

U. S. Phclear Regnintury CommissionDivision of Project rianagement Washington, D.C. 20555 Suldect:

B-SN1-205 - Outstanding Issues Rofenmce:

1.

Report to the Ady t r.ory Committen on Ronctor Safegion ti,e Office of Nucienr Reactor Regulation, II.

Company Reference Safety Analysis Report S. PAsclear I

Docket No. STri 50-561. July 9,1977.

2.

July 15,1977.J. H. Taylor to R. S. Doyd, "R-SAR-205 - Out 3

J.11. Tayinr to R. S. Boyst, "D-SAR-205 - Outstanding Issu July 21, 1977.

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tagust 4, 1977.J. H. Taylor to R. S. Boyd, *B.SAR-205 - On

Dear Hr. Dayd:

We have been inforried by the Staf T tint a ktitions)

This letter is specifically intendert to provirte act fitiona of Refere outstanding issuer. 3. fi, anil il at information conceni listeil in Sectioh 1.6 of Reference 1 In t repsnf the infon=ation in the at tachrant to this int ter supple already provided to the Staf f in Rarorences 2. 3. and 4 as appilc h rnnts infoma ti as a ranuit of a discussion with the Staf f on /,ugust 5 n

n.

In add attaclwent a member of revisions which should he mado to

, we a re inciwfing with outstanding issue ntsnber 12 In reference 3 and outstaniing issue 8 i position conr.

Me hope that the additional or clarified positions will resol n referenc issues before the ACRS Full Comittee Pfecting ve these outstand 8201060628 810403 G

PDR FOIA MADDEN 80-515 PDR

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I Balscock 5.Wih The Babcock & Wilcox Company coemits to forn. illy ine'uding modiffed materf a}

described in the attactrient to this lett.cr in a fieture an< ndment to th l

j which is scheduled for sutxnittal prior to August 31. 1977.

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truly yours.

  1. I Jnes H. Tayine Manager, Licensing JIIT:cr s

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I Aumist 8,1977 Page 1 of 8 l

(3) liigh Pressure Inlection Line Dreak_

i' Staff Position We require additional analyses to evaluate the consequences and agierator cctions required for the worst case high pressure injection line rupture.

This matter is discussed further in Section 6.3.4 of this report.

BMI Position _

The following supplemental information will he added to B-SAR-205 Section 6.3.2.17.2 in response to a staff request made on August 2.

states associated

" Table 6.3-4b (attached) sumarizes the finw alarrt with each break as a function of the HPI pump operating status and Closure changes to these states resulting from the nitigating action.

of only one valve is required for any HPI line break and any single active failure to assure that the ninirmm flow required for the break Thre operator action rettigation is achieved through an Intact HPI ilne.

to be taken is very straightforward in that he need only close the valve in the line which has high indicated flow and need not ascertain Furthermore, for this particular break, the which line is broken.

operator action time of 20 minutes following indication of the break, consistent with the assumptions of SRP 6.3. results in a margin of over ten minutes after operator action before core uncovering would 7

occur *.

Check valves MO-MY131 A, R. C, and D (refer to figure 9.3-1) which were inadvertently omitted from figure 6.3.7A wi11 be adtfed to this Also, the HPI flow indicator ranges given in Table 7.5-1 figure.

will be corrected to show 0-500 gpm in lieu of the presently shown 0-400 gpm.

Tne above changes will be included in Amendcent 17 to BSAR-20531, 1977.

which is scheduled for submittal to NRC on or prior to August

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4 TABLE 6.3-4b HPI LINE BREAX FLOW ALARN STATES (1) (2)

MUPumpsOperating(train)

Mitigatir.g Action reak location Follcwing ESFAS Actuation FE-43A FE-438 FE-43C FE-430 (Valve Closed)

CP-181 Discharge A

Hi(Lo)

Lo(Hi)

Lo Lo V43A B

Lo lo Hi(La) to(Hi)

V43C AaB Hi(Lo)

Lo(Hi)

Hi Lo V43A AaB Hi Lo Hi(Lo)

Lo(Hi)

V43C CP-182 Otscharge A

to(Hi)

Hi(Lo) to to V43B B

Lo lo Lo(Ht)

Hi(Lo)

V430 A&B Lo(Hi)

Hi(La)

Lo Hi V438 or A&B Lo Hi Lo(Hi)

Hi(Lo)

V430 P-1A2 Discharge A

Hi(La)

Lo(HI)

Lo lo V43A B

Lo Le Hi(Lo)

Lo(Hi)

V43C

+

AaB Hi(La)

Lo(Ht)

H1 lo Y43A Gr AAB Hi La Hi,(Lo)

Lo(Hi)

Y43C

7-1A1 01scharge' A

Lo(Hi)

Hi(Lo)

Lo lo Y438 B

Lo Lo Lo(Ht)

Hi(Lo)

Y430 AaB Lo(H1)

Hi(Lo)

Lo Hi V438 or AAB Lo Hi Lo(Ht)

Hi(Lo)

V430 ites:

) Alarm states in parenthesis are those resulting frca indicated mitigating action.

Where no parenthesis are shown, no change in state results from the indicated mitigating action.

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Instrument ranges are given in Table 7.5-1.

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August 8, 1977 Page 3 of 8 (6) _0verpressure Protection t

Staff Position i

"We will require Babcock & Wilcox to s9 ply additional information regarding the B-SAR-205 capability to protect the reactor pressure vessel from excessive pressures during low temperature operation i

such as startup or shutdown.

This issue is discussed further in Section 5.2.2 of this report."

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_B&W Position

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This discussion supplements the information supplied in the B&W Ictter from J. H. Taylor to R. S. Boyd of NRC dated 7/21/77.

i Features used to prevent overpressurization i

To prevent exceeding the ASME Code,Section III Appendix G limits.

l B&W relies on two passive features:

t (1)

Pressurizer safety valves (setpoint = 2500 psig) i (2)

Decay Heat Renoval System relief valves (setpoint = 455 psig)

The bounding curve famed by these two,"entures is shown as a heavy

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line on the attached figure.

Typical Appendix G Limit Curves Shown on the attached figure Js a typical Appendfx G Ifmit curve for 32

[

EFPY.

This curve shows'that the bounding curve will be an effective means for ensuring positively that the Appendix G limit i

will not be exceeded.

j These curves are typical based on:

(1) An expected initial RT of 10F.

This temperature is l

representative of forghs used recently in reactor vessel fabrication.

(2) An expected maximun copper content oT.03% for the forgings, i

This valve is representative for forgings used recently in reactor vessel fabrication.

Regulatory Guide 1.99 was used for analysis of the effects of this chemical composition.

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_BSAR Changes The changes cited in the letter of 7/21 and the clarifications and

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additions stated here will be made in the next Amendr.ent to the B-SNI-205 which will be submitted prior to August i

31. 1977.

An additional change requested verbally be the NRC Staff will be made.

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B&W will specify, as an interface requirement, the maximmn allowable accartulation of the DHRS relief valves.

5 August 8 1977 i '! '

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MOS PRESSURIZER SAFETY YALVE I

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.s.h' SET P0lHT (2500 PSIG) i (400 t

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7200 TYPICAL APPEl;DIK S

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TECH SPEC LilJIT.

2000 (32 EFPY)

THIS CURVE DILL b,.

I II E 1800

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LIRITS FOR DSAR (l:DMAL !!EATUP e

205 PLANT 1000 A G C00LU M )

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1400 lc

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1200

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1000 r'

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Dr3 SYSM SAFETY YAM 000

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SETPolNT (455 PSIG) 4DO j

TEMPEP.ATURE BELON WHICH THE i

DHRS ISOLATION VALVES NUST 200

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305*

BE OPEN i

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100 200 300 400 500 i

RC Tes17,*F I

I i ASME CODE SECTION III APPE OVERPRES9JRE LIMIT 9

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Augusc as, 3y//

Page 5 of 8 j

D. ray Heat Removal System Isolation Please stake the following corrections to the BfM position stated on Page 4 of reference (4):

j (1)

Change the first sentence under proposed Section 8.3.1.2 to read "Special provisions, which enable DH suction isolation I

valve D;i-VilA to be transferred....'

(2)

Change the third sentence under proposed Section 8.3.1.2 to read "Special provisions, which encbie DH suction isolation volve DH-V118 to be transferred...."

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Coev11ance with IEEE-323 (1974) for Makeup Pump and Decay Heat Pump Motors Staff Positicre "Ms will require that Babcock & Wilcox specify, as interface data, l

that they have elected the " ongoing qualification" option and that t

referencing applicant compliance with Standard 323-1974 will require periodic testing, including seismic qualification testing.

This i

catter is discussed in Section 7.8 of this report."

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BEM Position Follo:dng the discussion on this subject with the NRC Staff on August 2. B&W has decided to revise B-SAR-205 as follows. The following statement will replace the existing discussion in BSAR Section 3.9.2.4.3(6) " Err frorunental' Motor Testing."

" Class IE motors for safety related pumps will be qualified according to IEEE Standant 323-1974.

The qualification program including methods and acceptance criteria for envirorsnental qualification will be described in a suitable generic documentation such as a B&W topical report, or Appendix to the B-SAR-205 which will be submitted prior to the final design. app 1Mhtbr. h:e)Mh ybw: tun O-B&W cousilts to adopting the resolution of issues arising from the j

review of issues which may arise during the review of this qualification program in the B-SAR-205.

Environmental design conditions are defined in Table 3.11-2.

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I August 8,1977 Pa p 7 of 8 i

(12) _ Anticipated Transients E

! Page 5 of reference (3).. Please substitute the follouing for the B

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t The use of non-safety grade equipment for an t

i turbine trip via CRDCS and the L

transients (moderate frequency) turbine bypass system.

The anticipated group withdrawal at startup and at power, control rod misopera i

chmical and volume control system sialfunction (Boron Dilution n,

trip, loss of nomal feedwater, excessive heat removal

, turbine operation of the ECCS.

, and inadvertent credit for the bypass system is discussed later.The few cases of usel i

For the above listed, transients, except turbine trip I

ist initiated prior t i

o tripping the turbine.

, a reactor trip is that the heat demand during the transient and iThe significance of this the reactor trip is indep;mdent of the use of the rmediately following ne bypass systue.

trip is the same whether or not the bypass system is as i

ng turt>ine In addition, the heat demand used in the analysis more clos i

sumed to function.

,i the hea bypass.t demand wf thout turbine bypass than the heat demand w pproximates Analyses perfomed with the heat demand simulating no tu bypass system action showed negligible differences to the j

Fresented in B-SAR-205, Chapter 15 analysis of its nomal role, but is not a required functionin a ss used as part the atmospheric dump and/or safety valves with n on through transient response ths bypass is discus (sed)pt for turbine trip analysis where failure of exce g

e effect on j

The turbine trip transient is the only one in the categary of trip prior to reactor trip.

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(2),itea13 with and without turbine bypass.The turbine trip is addressed in The art >1trary use of an initial power level of 1 l

posf tive moderator coefficient produces DNR results tMt bound all ypass.

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of turbine trip fors: 102%

turbine bypass operation. power, regardless of the assustptions used for cases i

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i Papa 8 of 8 Tim only anticipated transients listed in Table 15.1.4 that assume turbine trip without turbine bypass action (discussed above) are the loss of four pumps, break in prirrary systear penetration lines, and control mo:n uninhabitability.

Tha equipment assumed to function in tha event of control room uninh:bitchility is dependent on thd initiating event.

A spectrum of postulated events is listed in BSAR Sectica 15.1.23, and for these events, the systems assumed to operate are sesamtely identified in TabTe 15.14.

The loss of four Ptmps and the break in primary system penetration Ifnes both cause a reactor trin prior to turbine trip.

Therefore, the effect of turbine bypass is negligible as it was for the bypass action discussed above.

Furthamore, as these two transients are under-cooling in nature, assu:2ing a turbine trip is conservative.

Tie above cases all discuss turbine trip with and without bypass.

Tha concern of no turbine trip following a mattor trip is addressed as follows.

For those accidents above that are overheating ty2a conditions, e.g. rod group withdrawal, the asstaption of a turbine trip is conservative.

For overcooling transients, the conclusions are not readily aparent.

Hawever, of primary interest in overcooling transients of moderate frequency is the DraR associated with the decrease in primary pressure.

For a bounding event, Section 15.1.36 of BSAR ms amended to show the case of 15% increase in steam flow.

For this case, the miniram DNBR occurs without a reactor trip, hence turbine as rot assumed to trip.

Considering overcooling transients of greater mgnitude f.e. where the increased steam flow causes a reactor tr.ip at the point of ofnisun D7GR, rod insertion in the first few seconds, ternifnstes the DtGR transient independent of the availability of turbine trip.

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