ML20096E924

Forwards Responses to Six NRC Recommendations Presented in SER Re SBO Rule & Calculations Re Subj

ML20096E924
Person / Time
Site:	LaSalle
Issue date:	05/15/1992
From:	Sheilds J COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20096E925	List: ML20096E924 ML20096E927 ML20096E931 ML20096E938 ML20096E945 ML20096E959 ML20096E970 ML20096E972
References
NUDOCS 9205200103
Download: ML20096E924 (9)

Text

gp a . -

+ .

'~

i'; 4 ,

a( , .

n '

p+.

WCommonwealth Edison Jy

j -. ( 1400 Opus Plece . .  ;

/ J Downers Grove, Illinois 60515 -

s hw-g, N.x -

hs;g . .

O cm, -

Hay 15, 1992 E

(M ; L t

U -Office of Dr.LThomes-E. Nurley,: Director ,

7s (Office of Nuclear keactor Regulation. ,

g%

.U.S. Sluclear Regulatory Commission .,

E' Washlagton, D.Cc ~20555 4 Attnt Document) Control' Desk LSubjecti LaSalle County Station Units 1 and 2 e '

~

Response to' Safety Evaluation on the Statian Blackout Rule

'tsC_Do.cLe.LNeic5h31LaDL5h314 r .

^m >-

Reference:

(a) . Byron L.iSiegel_ (NRC) letter to Thomas L-Kovach (CECO),

deted March 6. 1992, Safety' Evaluation of the LaSalle (County Station Response to the Station Blackout Rule w (b)- Peter L.EPlet (Ceco)4 1etter to USNRC dated September 23, '

Q-p -o 1991,' Supplemental Response to Station Blackout (SBO) Rule

!Dr. Mdriey,-

4' In the-Safety Evaluation for the. Station Blackout (SBO) rule,lrcference 9

,;(a),'the NRC Staff concluded that the d? sign.of.LaSalle County Stattor b

? r -

conformed with'the'SB0 rule contingent upon the satisfactory resolutior of the .

sixfrecommendations presented in the Safety Evaluation. Attachments A through ,

?,

F ofJthis-lettee contain LaSalle's: responses to these six recommendations.

1. L cAttachment G-contains copies of theLcalculations referenced in the Attachments.

s

• J -

Reference;(b) providedisupplemental'information on the Station Blackout iRulekLitem~5) in the Attachment.toTthat letter provided a list of: equipment

^

i 1which required a QALprogram in accordance with Regulatory Guide l!.155 z

LAppendices A and'B. Further review of- t'i'v11st revealed:that valves 6 ;1G33v2001-32A,?1G33-Z001-32B, IG33-Z001 M,-2G33-2001-32A, 2G33-Z001-1.B. and D 12G33-Z001-32CLare not required:forisaf- %utdown. These valves-are Reactor

' iMater CleantUp filter:demineralizer isution' valves, and not primary I

>c ir.ontainmentxisolation valves. -Therefore, these valves will not be; included in

~

y#

s the"SBO,QA program and'will-be deleted from this-list.

ih w

f y

, a i

?

'ZNLD)1806/1--

g, A nnnnn - i 1

) )

PDR 9205R00103 T20515

ADOCK O9000373

.P 2PDR m

. 4 i t t --

p

x

JDr. T.E..Murley May 15,1992 If there'are any questions regarding this matter, please contact this

, ' office.

Sincerely, Jr /h

/

JoAnn M. Shields Nuclear Licensing Administrator 1

Attachments: A. Effects of loss of ventilation on control room, AEER, and RCIC room temperatures

-B. Effects of loss.of ventilation on drywell temperature profile C. Reactor coolant inventory analysis demonstrating core

. coverage-10 . Suppression pool 1 temperature effects on RCIC and HPCS E. -Suppression pool temperature effects on RHR F. Modification documentation and retention

.G. _ Supporting Calculations

cc: A.B. Davis, Regional Administrator - RIII D. Hills,_ Senior' Resident Inspector - LaSalle B.L. Siegel - NRR Projp?t-Menager P. Gill NRR Electricat Systems Branch

--Office of. Nuclear Facility-Safety - IDNS 1

2NLD/1806/2 e

4 ~

l

-Attachment A Effects 2of Loss of Ventilation on Control Room,: AEER, and RCIC Room Temperatures A revision has been made to the control room, AEER and RCIC room temperature transient calculations based on an additional evaluation which included a surveillance review of summertime room _ temperatures. The  !

-combinati e of higher, more conservative initial room temperatures and -i l

-smaller, more realistic electrical heat loads in the calculation resulted in higher final room temperatures and necessitated the procedural requirement-to open panel doors in both the controi-room and AEER during the event. The AEER ha been analyzed with closed access doors to maintain less than 120 *F room temperatures. THe doors will only be opend for personnel access. The other conclusions stated in the original control room and AEER temperature transient calculations remain valid. The final RCIC room temperature is also slightly higher in its revised calculation although its original conclusions remain ']

valid.

'The basis for reasonable assurance of equipment operability using the I results of revised calculations 3C7-0290-001, Rey, 1, 3C7-0289-001, Rev. I and ,

3C7-0290-002, Rev. I are listed in the tab'e below:

Area InLt taLIempa finaLI.emp, RAUustification Control Room 90'? Il6*F Less than 120' (open panel doors)

Ul South AEER 90*r 110*F Ul-North AEER- 90'F 119.7'F U2 CE AEER 90*F 119.7*F U2 NW AEER 90'F 108.6*F RCIC Rooms 124.0*F 164.7'r less than the design temp.

of 212*F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> The-Center Desk Shiftly Surveillance LOS-AA-52 which monitors area temperatures in accordance with Technical Specification 3/4.7.7 will be

revised to verify the 5B0 assumption that the initial control room and AEER temperatures are less than or equal to 90*F. Ifithis limit is exceeded, appropriate action will be taken to investigate the problem and resolve it in a timely menner. -The procedure revisions required to open control room and AEFD panel doors and monitor their area temperature daily for 5B0 will be compieted withir one year'after the issuance date of the Safety Evaluation.

h l

l-l L

ZHLD/1806/3

_ __ __ _ _ _ . _ _- _ _ ._ , . , _ _ . . _ , _ . _, ~ ~

Attachment B-

-Effects of-Loss of-Ventilation on Drywell Temperature Profile

~

-- The drywell_. temperature transient calculation 3C7-0390-002 was revised to include the-affect of venting to-the suppression chamber and the prer'ure decay of the_ vessel during the coping pertod.__ The resultant maximum drywell 1

- temperature during a four-hour SB0 is 251'F. The six ho'ur drywell te'nperature is_245'F. The equipment qualification curve for the drywell_is a step

- function _with the following temperature limits:

340'F from 0 - 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 320*F f rom 3 -. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 250'F from 6 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Thus, the EQ equipment inside the d;ywell is designed to operate at 320'F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A review of the drywell temperature analysis while taking into account the SB0 time duration shows that the EQ temperature profile envelopes the SB0 conditiens.

l 1

i

ZNLD/1806/4 i

d

'.. - .',_,..,.._w, .. _ . . ,, , , , . , . ,_ ,,,, ym,,',.

.. . -. - . . - ~ - - . ~ _ . . - - . . - - . -

Attachment C Reactor Coolant Inventory Analysis Demonstrating Cove Coverage Either the RCIC or HPCS systems may be utilized to provide reactor coolant inventory makeup while coping-during a SBO, Original calculatter.

-3C7-0189-001, Rev. 2, SB0 Condensate. Inventory Coping Assessment shows that the suppression poo.1 has sufficient capacity to compensate for primary system losses due to SRV steam discharge, RCIC turbine steam requirements and the assumed RPV leakage. Additionally, the Suppression Pool Temperature Transient calculation 3C7-0390-0W, Rey, 1, models the RPV level for either the HPCS pump or RCIC pump operations, Per the analysis, the lowest reactor vessel level occurs during RCIC makeup at approximately -130". This level does not result in core uncovery since the Top of Active Fuel (TAF) is -161" per LaSalle Technical Specification Bases figure B3/4 3-1, l

l l

l 1

i i

ZNLD/1806/5

Attachment 0 Suppression Pool Temperature. Effects'on RCIC and HPCS The maximum 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 15 minute _ suppression pool temperature is__234.2*r when using the HPCS system and 217.1?f when using the RCIC systam for decay-heat removal and_ reactor ccolant inventory, as stated in the Suppression Pool Temperature Transient calculation 3C7-0390-001, Rev. 1. The qualified

-temperature rating for the RCIC pump materials is 121'T and for_the HPCS-materielsLis 300'F per calculation CQD-05SO9frRev. O. Thus, the maximum suppression' pool' temperatures have no adverse effect on the RCIC and HPCS materials.

When HPCS or RCIC'15 taking suction from the-suppression pool, the-suppression pool-temperature affects the pump's Net Positive Suction Head Available (NPSHA ). An evaluation of the suppression pool during SB0 and untti pool cooling becomes available shows that the NPSHA for the HPCS or RCIC pump -

exceeds the Net Positive Suction Head Requirements (NPSHp). The following

-table summarizes these _ results as stated in calculation ATD-Oll7, Rev. 0:

Eltmp MP.5H p, NPJ1Q

. RCIC-- 15 ft. 22.6 ft.

HPCS 1.5 ft. 16.5 ft.

The RCIC turbine backpressure was determined based on worst case ,

suppression pool water levels, suppression chamber pressure and RCIC turbine exhaust flow following the $B0.

- The-calculated maximum RCIC turbine

. backpressure is 23.1 psig at four hours following SBO and is 24.5 psig at four hours and-15 minutes following SBO. These pressures are below the RCIC-turbine backpressure. trip setpoint of 25 psig (see calculation ATD-0117, Rev.

0),

u e

ZNL.0/1806/6

Attachment E Suppiession Pool Temperature Effectt on RHR e

The maximum suppression pool temperature when utilizing HPCS for decay.

heat removal and reactor inventory. control is 234.2'F; The RHR materials are unaffected since their qualified temperature rating-is 300*F.(calculation.

CQO-055096 Rev, 0). An evaluation of the suppression pool up until the time suppression pool cooling becomes-available shows that the NPSHA for the RHR pumps is 16.2 ft. while the NPSH Thus the NPSHA exceeds the l NPSH R (see calculation ATD-Ol'7,g Rev. 0). is 11.5 ft.

l l

I

.l l

f 6

l' r

ZNLD/1806/7 l:-

, =. ..- , - . -. .- __ _

___ _. __. -~ _ _ _ _ _. _ _ _ .- _ _ _ . _ _ _ - _ _ _ _ , _ . . . _ _ _

Attachment f Modification Documentation and Retention LaSalle Station has six Class IE 125 Vde-batteries and 2 Class IE 250 Vdc i batteries.fDue to the addition of DC loads since fuel load, the safety

---related DC systems had reached their maximum load for the existing size.

batter 125. Subsequent to the decision to replace these batteries the Station Blackout analysis was completed and reviewed against the replacement batteries for.its impact. The calculations concluded-that the' replacement Class IE 1 batteries have-adequate capacity to feed 5B0 loads for a four-hour duration and-toirestore ac power following coping assuming non essential load

-shedding. All Class IE batteries ~have been replaced except for the Unit I division 3 battery which is scheduled for the fourth quarter of 1992.

However, the division 3 battery.is not. required to cope with an SB0 and the ,

division 3 battery. charger would be available if the division 3 generator was utilized during an SB0 event. The table below references the modifications and the-SB0_ load calculations associated with the replacement batteries. A full description of the nature and objective of the modifications can be found in these documents.

DES _c tik tio!L_._.. Modificat.ionJo__ CdculationloL_

U1 DIV-1 125Vdc-Bat M01-1-88-004 (4266/19D30 Rev, i U1 DIV-2 125Vdc Bat M01-1-88-003- &

U1 250 Vdc Bat M01-1-88-001 4266/19031 Rev. 0)

U2 DIV-1-125Vdc Bat M01-2-88-004

-U2 DIV-2 125Vdc Bat M01-2-88-003 U2.250Vdc bat M01-2-88-001 U1 DIV-3 125Vdc Bat M01-1-90-011 (sched-4th quarter 1992)

U2 DIV-3 125Vd'c Bat M01-2-90-009 ZNLD/1806/8.

1.

V __

i.

__ w___.,, . . - , , , , - - - ~ . ~ . , . - , , , , ,, ,,--,- , . . ~~..,,e - ,- -

i

~ '

Attachment G ,

Supporting Calculations 3C7-0290-001 -Rev;-1, Main Control Room Temperature Transient following-Station Blackout 3C7-0289-001, Rev.-1, AEER Temperature Transients During Station Blackout. ,

3C7-0290-002, Rev. 1, RCIC Pump Room Temperature Transient following Station -

Blackout 3C7-0390-002, Rev.-1 Drywell Temperature Transient following Station Blackout

- 3C7-0390-001, Rev. 1 Suppression Pool Temperature Transient following Station Blackout .

C00-05S096. Rev. O, Calculation for EQ Impact due to-Station Blackout Conditions ATD-0117,:Rev. O. Evaluation of NPSH requirements for HPCS, RHR, and RCIC Pumps and Back. Pressure Limitations of RCIC Turbine

~Following Station Blackout ,

e d

2 ZNLD/1806/9-I

- - , _m.__ ., _ _ __._ _ _ _ _ _ ____ _ _ __