ML20044E282

Summary of 930413-15 Meeting W/Util in San Jose,Ca Re Resolution of Draft SER Open Items in Chapters 3,5,6,9,10 & 11.List of Attendees,List of Items Discussed,Summary of Status of All Splb Items & One Oversize Drawing Encl

ML20044E282
Person / Time
Site:	05200001
Issue date:	05/05/1993
From:	Lyons J Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9305240143
Download: ML20044E282 (120)

Text

g May 5, 1993 t

i Docket No. S2-001 l

f APPLICANT:

GE Nuclear Energy (GE) j PROJECT:

Advanced Boiling Water Reactor (ABWR)

SUMMARY

OF MEETING WITH GE ON APRIL 13, 14, and 15, 1993 On April 13,14, and 15,1993, members of the Advanced Reactor Section of Plant Systems Branch (SPLB) met with General Electric in San Jose, California, to discuss the resolution of Draft Safety Evaluation Report (DFSER) open items in Chapters 3, 5, 6, 9, 10, and 11. is a list of attendees. lists the items discussed during the meeting and their status. As part of the discussions, GE provided the attached mark-ups of the Standard Safety Analysis Report that the staff will use to begin development of the final Safety Evt.luation Report input.

A summary of the status of all SPLB items identified in the DFSER is provided in.

l GYMW 5+*u y i

James E. Lyons, Chief l

Advanced Reactor Section Plant Systems Branch Division of Systems Safety

.i and Analysis Office of Nuclear Reactor Regulation L

Enclosures:

As stated cc w/ enclosures:

See next page p yu..~x (*j qv T? p;N3y un ~.

v., -

a DISTRIBUTION w/ enclosures:

Docket'Filei PDST Rdg File JLyons CPoslusny PDR SPLB File WBurton PShea j

i w/o enclosures:

j TMurley/FMiraglia 0GC ACRS-10 EJordan i

AThadani GHolahan DCrutchfield l

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SPLB:DSSA SPL " SA PM:PDS B:DSSA WBurton{g JL n CPoslusny

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930524o143 93050s I

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PDR ADOCK 05200001 1

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GE Nuclear Energy Docket No.52-001 cc: Mr. Patrick W. Marriott, Manager Mr. Joseph Quirk Licensing & Consulting Services GE Nuclear Energy i

GE Nuclear Energy General Electric Company 175 Curtner Avenue 175 Curtner Avenue, Mail Code 782 i

San Jose, California 95125 San Jose, California 95125 Mr. Robert Mitchell General Electric Company 175 Curtner Avenue l

San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy i

12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 e

t Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C.

20460 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.

20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Room 8002.

Washington, D.C.

20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 Washington, D.C.

20006 Marcus A. Rowden, Esq.

Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.

Suite 800 Washington, D.C.

20004 Jay M. Gutierrez, Esq.

Newman & Holtzinger, P.C.

1615 L Street, N.W.

Suite 1000 Washington, D.C.

20036 o

7-ABWR MEETING APRIL 13-15, 1993 LIST OF ATTENDEES J. Fox GE J. Pcwer GE G. Elhert GE A. Beard GE B. Genetti GE C. Oza GE E. Nazareno GE M. Munson GE B. Strong GE C. Sawyer GE U. Saxena GE G. Miller GE H. Careway GE J. Lyons NRC W. Burton NRC

i i

ABWR OPEN ITEMS MEETING April 13 - 15, 1993 i

FLOOD CONTROL 01 3.4.1 RESOLVED CONFIRMATORY CI 3.4.1 RESOLVED CONFIRHATORY MISSILE PROTECTION 01 3.5.1.2 CLOSED by Amendment 26 COL 3.5.1.2 CLOSED by Amendment 26 COL 3.5.1.4 CLOSED by Amendment 26 COL 3.5.2 CLOSED by Amendment 26 PIPE BREAKS CI 3.6.1 RESOLVED - GE to provide by 4/30 EQUIPMENT OVALIFICATION 01 3.11.3 RESOLVED CONFIRMATORY 01 3.11.3 OPEN - GE will revise App 31 to add all radiation zone values (by 4/30).

CI 3.11.2.1 RESOLVED CONFIRMATORY - GE to reference topical instead of non-i proprietary App 3.K.

CI 3.11.3 CLOSED by Amendment 21 CI 3.11.3 OPEN - GE will revise App 3I time based profiles and add all radiation zone values (by 4/30).

COL 3.11.3 OPEN - GE will revise App 3I to add all radiation zone values (by 4/30).

CLOSED - Reference to 50.49 in Sect 3.11.6.2 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE CI 5.2.5 CLOSED by Amendment 24 ALL NON-ITAAC ITEMS IN CHAPTER 5 ARE COMPLETE P

i i

PRIMARY CONTAINMENT 01 6.2.1.6 RESOLVED - (GE to submit changes by 4/30) 01 6.2.1.6 RESOLVED - (GE to submit changes by 4/30) 01 6.2.1.6 RESOLVED - (GE to submit changes by 4/30)

CI 6.2.1.6 RESOLVED - (GE to submit changes by 4/30)

[

01 6.2.1.7 RESOLVED - (GE to submit changes by 4/30)

CI 6.2.1.7 RESOLVED - (GE to submit changes by 4/30) a CONTAINMENT PURGE & VENT i

01 6.2.4.1 OPEN - Both valves outside containment (DC solenoid valves). Reason is for accessibility and EQ considerations. NRC wants additional info of why this arrangement is acceptable, including discussion of common second isolation valves j

for both wetwell and drywell, branch lines between 1st and 2nd isolation valves, qualification of the lines, leakage control and leak tightness, etc. GE to provide additional justification and description.

OI 6.2.4.1 OPEN - Simultaneous venting of wetwell and drywell - GE will add COL l

actinn item to provide procedure to prevent simultaneous venting. NRC will check I

to see what is acceptable for the periods just before and after shutdown.

{

OI 6.2.4.1 OPEN - Common 2nd isolation valves - see 01 6.2.4.1-1, I

CONTAINMENT ISOLATION i

CI 6.2.4 RESOLVED CONFIRMATORY - GE will modify Tb1 6.2.7 to indicate the manual valves are locked close.

K 10STIBLE GAS CONTROL CI 6.2.5 RESOLVED - GE will revise 4/2 submittal CI 6.2.5 RESOLVED CONFIRMAiORY - 4/2 submittal l

CI 6.2.5 RESOLVED - GE provided markup COL 6.2.5 RESOLVED - GE to provide by 4/30/93 COL 6.2.5 RESOLVED - Testing addressed in Tb13.9-8 (IST) GE to revise Chpt 19 to make consistent. NRC will revise SER to address the overpressure portions of purge and vent system.

f CONTAINMENT LEAKAGE TESTING 01 6.2.6 OPEN - GE to check Tech Specs to see if it's there and reference it in Sect 6.2.6.

01 6.2.6 OPEN - Input by 4/30 01 6.2.6 RESOLVED - GE provided update I

o CONTROL ROOM HABITABILITY 016.4 RESOLVED - GE will provide minimum duct sizes for the major lines, normal positions for dampers and how they are positioned for different operating modes. The rest of this open item will be addressed by rewrite of SSAR to include COL action items.

Input by 4/30/93.

ESF ATMOSPHERE CLEANUP 01 6.5.1 OPEN - App 6A OK; but deviations in App 6B are still not adequately

(

justified. GE to revisit, talk to committee member (Mike Fox), and provide revised i

input. Our experts (Jack Hayes & Ron Bellamy) feel the SRP Tb1 is minimum set and i

should be included.

01 6.5.1 RESOLVED - GE will revise to state when they will use syster: and add COL action item to provide analysis if SGTS is used more than 1% of tu f,ime (90 hrs).

CI 6.5.1 RESOLVED CONFIRMATORY i

6.5.1 NOTE GE will provide comparison between R.G.1.52 and Tb16.5-1 By 4/30/93.

l 9.1.1 NEW FUEL STORAGE l

f GE will add to COL action items requirement for procedures to prevent inadvertent i

placement of fuel assemblies as already exists in Section 9.1.1.

j I

RESOLVED CONFIRMATORY - Section 9.1.1.2 discusses rad monitors and refers to Sect 7.1, Should reference Sect 12.3.4.

Response to RAI 430.181 - 183 should be in text of SSAR Sect 9.1.1.

Need COL action item to prevent tipping of fuel rack, j

9.1.2 SPENT FUEL STORAGE e

Response to RAI 430.185 - 189 should be in text of SSAR Sect 9.1.1.

Defective fuel storage - in Chpt 3 (Tb1 3.2-1) but not in Sect 9.1.2.

Discussed' l

in Sect 9.1.4.2.8.

l 9

SPENT FUEL POOL COOLING AND CLEANUP OI 9.1.3 CLOSED - We were given a discussion of the issue by GE in December l

which resolved the issue. NRC will incorporate in FSER.

CI 9.1.3 CLOSED by Amendment 24 Protection of common length of piping to SFP -

t i

i 1

i

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r i

i HEAVY LOADS OI 9.1.5 CLOSED by Amendment 24 i

01 9.1.5 CLOSED by Amendment 25 CLOSED - by Amendment 25 - Hatches in control building - GE revised dwgs i

OI 9.1.5 CLOSED by Amendment 25 l

01 9.1.5-6

- CLOSED 01 9.1.5-7

- CLOSED CI 9.1.5-1

- CLOSED by Amendment 25 COL 9.1.5 CLOSED by Amendment 25 i

POTABLE AND SANITARY WATER 0I 9.2.4 CLOSED by Amendment 21 MAKEUP WATER COL 9.2.10 CLOSED - NRC to rewrite SER to remove requirement.

CHILL WATER OI 9.2.13 OPEN - GE will provide information on how chillers will be affected by SB0 in the next 3 weeks.

t TURBINE SERVICE WATER OI 9.2.16 CLOSED by Amendment 24 CI 9.2.16 CLOSED by Amendment 21

[

E0VIPMENT AND FLOOR ORAINS COL 9.3.3 RESOLVED CONFIRMATORY - GE will revise Tb1 3.2-1.& Sect 1.2 and review rad monitoring requirement.

l RADI0 ACTIVE ORAIN TRANSFER CI 9.3.8 RESOLVED CONFIRMATORY - by Amendment 28 CI 9.3.8 RESOLVED CONFIRMATORY - by Amendment 28 COL 9.3.8 RESOLVED CONFIRMATORY - by Amendment 28 j

CONTROL ROOM HVAC OI 9.4.1.1 CLOSED by Amendment 25 CI 9.4.3.1 RESOLVED CONFIRMATORY.- by Amendment 27 or 28 l

COL 9.4.1.1 CLOSED by Amendment 24 i

i REACTOR BUILDING HVAC l

CUW pump and heat exchanger rooms do not have internal recirc fan coolers like other non-essential equipment. Why?

RADWASTE BLDG HVAC COL 9.4.6 RESOLVED CONFIRMATORY SERVICE BLOG HVAC l

COL 9.4.8 OPEN - Input by 4/30/93 DIESEL GENERATOR i

CI 9.5.4.2 CLOSED by Amendment 25 CI 9.5.4.2 CLOSED by Amendment 25 CI 9.5.4.2 CLOSED by Amendment 25 i

CI 9.5.6 CLOSED by Amendment 25 CI 9.5.7 CLOSED by Amendment 25 L

CI 9.5.7 CLOSED by Amendment 25 01 9.5.8 CLOSED by Amendment 24 CI 9.5.8 CLOSED by Amendment 24 ALL NON-ITAAC ITEMS IN 9.5.2 THRU 8 ARE CLOSED i

FIRE PROTECTION I

DI 9.5.1.3.1 CLOSED by Amendment 21 l

CI 9.5.1.2.2 RESOLVED CONFIRMATORY COL 9.5.1.4.6 RESOLVED CONFIRMATORY - 4/2/93 SUBMITTAL MAIN STEAM CI 10.3.1 CLOSED by Amendment 21 CONDENSATE AND FEEDWATER CI 10.4.7 CLOSED by Amendment 25 (SECTION 5.4.9.3)

ALL NON-ITAAC ITEMS IN CHAPTER 10 ARE COMPLETE

t RADI0 ACTIVE WASTE MANAGEMENT COL 11.0 RESOLVED CONFIRMATORY j

CI 11.2.2 RESOLVED CONFIRMATORY - Submittal was good but need to identify

[

tanks and show vents on P& Ids.

COL 11.2.1 RESOLVED CONFIRMATORY COL 11.2.1 RESOLVED CONFIRMATORY COL 11.2.2 RESOLVED CONFIRMATORY COL 11.2.2 RESOLVED CONFIRMATORY COL 11.2.2 RESOLVED CONFIRMATORY l

^

COL 11.3.2 RESOLVED CONFIRMATORY i

COL 11.4.1 RESOLVED CONFIRMATORY COL 11.4.1 RESOLVED CONFIRMATORY COL 11.4.1 RESOLVED CONFIRMATORY

{

COL 11.4.1 RESOLVED CONFIRMATORY COL 11.4.2 RESOLVED CONFIRMATORY t

COL 11.5.2 CLOSED by Amendment 23 (Section 10.4.10.1) i CI 11.5.2 CLOSED

}

ALL NON-ITAAC ITEMS IN CMAPTER 11 ARE COMPLETE USI/GSIs 01 20.2 Integrate the response in RSW section and include in 9.2.17.4 RSW COL action items.

r 01 20.3 CLOSED - Electric power requirements OK. GE verified that performance requirements are in Sect 9.4.5.

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ABWR OPEN ITEMS l

PLANT SYSTEMS BRANCH i

l

(

TOTALS j

i Total Resolved Resolved Resolved Un-ITAAC l

Confirm Amend.

resolved Open 93 3

8 29 6

47

)

Confirm 59 6

8 43 2

}

COL ACT.

55 2

15 35 3

f i

Y Chapter 3.

l Total Resolved Resolved Resolved Un-ITAAC Confirm Amend.

resolved j

s Open 12 2

1 1

8 l

f Confirm 12 1

2 8

I i

j COL ACT.

16 15 1

.i Chapter 5.

l f

Total Resolved Resolved Resolved Un-ITAAC i

Confirm Amend.

resolved Open 1

1 Confirm 1

1 l

COL ACT.

I 1

{

Chapter 6.

"1 i

Total Resolved Resolved Resolved Un-ITAAC Confirm Amend.

resolved Open 28 3

5 10 4

6 i

Confirm 7

3 3

1 COL ACT.

4 2

2 t

4 4

v

-r. - - -. -

1 I

i l

Chapter 9.

I i

f Total Resolved Resolved Resolved Un-ITAAC i

Confirm Amend.

resolved 4

Open 43 1

18 1

23 i

Confirm 34 4

30 l

COL ACT.

19 4

13 2

i Chapter 10.

Total Resolved Resolved Resolved Un-ITAAC Confirm Amend.

resolved Open 8

8 Confirm 3

3 COL ACT.

3 3

l Chapter 11.

f i

Total Resolved Resolved Resolved Un-ITAAC r

Confirm Amend.

resolved Open 1

1

)

l Confirm 2

1 1

COL ACT.

12 11 1

l l

Resolved Verbal agreement has been reached. GE will submit mark-up or

}

amendment to the SSAR.

Resolved Confirmatory - Agreement has been reached based on mark-ups of the SSAR. GE will submit as amendment.

Resolved Amendment - An amendment to the SSAR resolving the item has been received.

Unresolved - Agreement has not been reached or GE has not provided information that they committed to provide (Confirmatory Items).

ITAAC - Will be reviewed when final ITAAC submittal is received.

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col

(,, 2, b~ ~ 2 23A6100AE

/ #/ 3 nev n Standard Plant l

3.9,7 COL License Information Subsection 3.9.3.1.)

3.9.7.1 Reactor Internals Vibration Analysis, 3.9.73 Pump and Valve Inservice Testing Measurement and Inspection Program Program U L.

The first COL applicant will provide, at COL applicants will provide a plan for the fg* g the time of application, the results of the detailed pump and valve inservice testing _ad vibration assessment program for the ABWR inspection program. This plan will prototype internals. These results will include the following information specified in Regulatory (1) Include baseline pre-service testing to Guide 1.20.

support the periodic in-service testing of the components required by technical R. G.1.20 Subiect specifications. Provisions are included to disassemble and inspect the pump, check C.2.1 Vibration Analysis valves, and MOVs within the Code and Program safety-related classification as necessary, C.2.2 Vibration Measurement depending on test results. (See Program Subsections 3.9.6, 3.9.6.1, 3.9.6.2.1 a n d C.23 Inspection Program 3.9.6.2.2)

C.2.4 Documentation of Results (2) Provide a study to determine the optimal frequency for valve stroking during NRC review and approval of the above inservice testing. (See Subsection information on the first COL applicant's docket 3.9.6.2.2) will complete the vibration assessment program requirements for prototype reactor internals.

(3) Address the concerns and issues identified in Generic Letter 89-10; specifically the In addition to the information tabulated method of assessment of the loads, the above, the first COL applicant will provide the method of sizing the actuators, and the information on the schedules in accordance with setting of the torque and limit switches.

the applicable portions of position C.3 cf (See Subsection 3.9.6.2.2)

Regulatory Guide 1.20 for non-prototype internals.

3.9.7.4 Audit of Design Specification and Design Reports Subsequent COL applicants need only provide the information on the schedules in accordance COL applicants will make available to the with the applicable portions of position C.3 of NRC staff design specification and design Regulatory Guide 1.20 for non-prototype reports required by ASME Code for vessels, I internals. (See Subsection 3.9.2.4),

pumps, valves and piping systems for the purpose of audit. (See Subsection 3.9.3.1) 3.9.7.2 ASME Class 2 or 3 or Quality Group D Components with 60 Year Design Life 3.9.8 References COL applicants will identify ASME Class 2 1.

BIVR Fuel Channel Mechanical Design and or 3 or Quality Group D components that are Dc/lection, NEDE-21354-P, September 1976.

subjected to cyclic loadings, including operating vibration loads and thermal transients cffects, 2.

BHR/6 Fuel Assembly Evaluation of Combined of a magnitude and/or duration so severe the 60 Safe Shutdown Earthquake (SSE) and year design life can not be assured by required Loss-of-Coolant Accident (LOCA) Loadings, Code calculations and, if similar designs have NEDE-21175.P, November 1976.

not already been evaluated, either provide an appropriate analysis to demonstrate the required 3.

NEDE-24057-P (Class III) and NEDE-24057 design Ufe or provide designs to mitigate the (Class I) Assessment of Reactor Internals.

magnitude or duration of the cyclic loads. (See Vibration in BWR/4 and BWR/5 7'lants,

\\

3943 Amendment 23

Co L f', 7. f-L ABWR

2 a3 usamse Standard Plant nev a Table 3.9-8 (Continued)

IN-SERVICE TESTING SAFETY-RELATED PUMPS AND VALVES T22 Standby Gas Treatment System Valves Safety Code Valve Test Test SSAR Class Cat.

Func. Para Freq. Fig.

No.

Qty Descriptien (h)(i)

(a)

(c)

(d)

(e)

(f)

(g)-

F012 2 Filter train DOP sampling line valve 3

B P

El 6.5-1(2,3) downstream of after HEPA F014 2 STGS sample line valve 3

B P

El 6.5-1(2,3)

F015 2 PRM discharge to stack valve 3

B P

El 6.5-1(2,3)

F500 2 Filter unit vent line valve 3

B P

El 6.5-1(2,3)

F501 2 Filter unit drain line valve 3

B P

El 6.5-1(2,3)

F5(M 2 Filter unit vent line valve 3

B P

El 6.5-1(2,3)

F505 2 Exhaust fan vent line valve 3

B P

El 651(2,3)

F506 2 Filter train vent line valve 3

B P

El 6.5-1(2,3)

F507 2 Filter train vent line valve 3

B P

El 6.5-1(2,3)

F508 2 Filter train vent line valve 3

B P

El 651(2,3)-

F509 2 Filter train vent line valve 3

B P

El 651(2,3)

F510 2 Filter train vent line valve 3

B P

El 651(2,3)

F511 2 Exhaust stack drain line valve 3

B P

El 651(2,3)

F700 2 Filter unit demister dp instrument line valve 3

B P

El 651(2,3)

F701 2 Filter unit demister dp instrument line valve 3

B P

El 651(2,3)

F705 2 Filter train prefilter dp instrument line valve 3

B P

El 6.5-1(2,3)

F706 2 Filter train prefilterdp instrument line valve 3

B P

El 651(2,3)

F707 2 Filter train preHEPA dp instrument line valve 3 B

P El 6.5-1(2,3)

F708 2 Filter train preHEPA dp instrument line valve 3 B

P El 651(2,3)

F709 2 Fliter train charcoal adsorber dp inst. line viv 3 B

P El 651(2,3)

F710 2 Filter train charcoal adsorber dp inst line viv 3

B P

El 651(2,3)

F711 2 Filter train after HEPA dp inst line valve 3

B P

El 6.5-1(2,3)

F712 2 Filter train after HEPA dp inst line valve B

P El 651(2,3)

F713 2 Filter train exhaust flow instrument line vahr 3 B

P El 6.5-1(2,3)

F714 2 Filter train exhaust flow instrument line valve 3 B

P El 6.5-1(2,3)

T31 Atmospheric Control System Valves A

F001 1 N2 supply line from Reactor Building HVAC 2 A

I,A LP 2 yrs 6.2-39(1)

S 3mo F002 1 N2 supply line to drywell inboard cont-2 A

1.A L,P 2 yrs 6.2-39(1) ainment isoaltion valve S

3 mo F003 1 N2 supply line to werwell inboard cont-2 A

1,A L,P 2 yrs 6.2-39(1) ainment isoaltion valve S

3 mo F004 1 Containment atmosphere exhaust line from 2

A I,A L,P 2 yrs 6.2-39(1) drywellisoaltion valve S

3mo F005 1 Drywell atmosphere exhaust line vahr 2

A I,A L,P 2 yrs 6.2-39(1)

T31-F004 bypass line S

3mo F006 1 Containment atmosphere exhaust line form 2

A 1,A L,P 2 yrs 6.2-39(1) cO&

wetwellisolation valve S

3 mo I,. 2 J-f F007 1 Wetwell overpressure line valve 2

A P

L,P 2 yrs 6.2-39(1) 39 5827 Amendment 24

cat

6. 2. r-2 ABM 3 of y 23xeioaxe Standard Plant ne.. n Table 3.9-8 (Continued)

IN-SERVICE TESTING SAFETY-RELATED PUMPS AND VALVES T31 Atmospheric Control System Valves Safety Code Valve Test Test SSAR Class Cat.

Func. Para Freq. Fig.

No.

Qty Description (hui)

(a)

(c)

(d)

(e)

(f)

(g)

RD8 1 Containment atmosphere exhaust line 2

A I,A I,P 2 yrs 6.2-39(1) to SGTS S

3mo 1

Containment atmosphere exhaust line to 2

A I,A L,P 2 yrs 6.2-39(1)

R/B HVAC S

3mo Drywell overpressure lir,e vahr 2

A P

I,P 2 yrs 6.2-39(1) d 7,5~2 N2 supply line from K-5 outboard cont-2 A

I,A L,P 2 yrs 6.2-39(1) ainment isolation valve S

3mo FU39 1 N2 supplyline from K-5 outboard cont-2 A

I,A L,P 2 yrs 6.2-39(1) ainment isolation vaht S

3mo F040 1 N2 supply line from K-5 to drywell inboard 2

A 1,A 1,P 2 yrs 6.2-39(1) isolation valve S

3 mo F041 1 N2 supply line from K-5 to wetwell inboard 2

A I,A L.P 2 yrs 6.2-39(1) isolation valve S

3mo F044 8 Drywell/wetwell vacuum breater vahr 2

C A

P RO 6.2-39(2)

R E3 F050 1 N2 supply line to drywell test line valve 2

B P

El 6.2-39(1)

F051 1 Containment atmosphere exhaust Ime test 2

B P

El 6.2-39(1) line valve F054 1 Drywell personnel air lock hatch test 2

B P

El 6.2-39(2) line vahr 7

F055 1 N2 supply line from test line valve 2

B P

El 6.2-39(1)

F056 1 Wetwell personnel air lock hatch test 2

B P

El 6.2-39(2) line valve F700 1 N2 supplyline to drywell FE upstream 2

B P

El 6.2-39(1) instrument line F701 1 N2 supply line to drywell FE downstream 2

B P

El 6.2-39(1) instrument line F702 1 N2 supplyline to wetwell FE upstream 2

B P

El 6.2-39(1) instrument line F703 1 N2 supply line to wetwell FE downstream 2

B P

El 6.2-39(1) i instrument line r

F720 8 DW/WW neuum breaker valve N2 supply 2

A I,P L

RO 6.2-39(2) line isolation valve I'/30 1 Drywell pressure instrument line isolation 2

B P

El 6.2-39(2) vahe F731 1 Drywell pressure instrument line solenoid 2

A I,P 1,P RO 6.2-39(2) isolation valve F732 2 Drywell pressure instrument line valve 2

B P

El 6.2-39(2) l i

F733 2 Drywell pressure instrument line solenoid 2

A 1,P 1,P RO 6.2-39(2) isolation valve F734 4 Drywell pressure instrument line for NBS 2

B P

El 6.2-39(2) valve 3.9 58.28 Amendment 23

ABWR 2mimo Standard Plant nry c 62.63 Containment Iselation Valve 12akage a flange), that are connected between isolation Rate Test (Type C) valves and form a part of the primary containment boundary need not be Type-C tested 62.63.1 General due to their infrequent use and multiple barriers as long as the barrier configurations any y

g Type C tests will be performed on all are maintained using an administrative control g,yj.4 containment isolation valves required to be program, p s p 7-6, 2., 6, L /

tested per 10CFR$0 Appendix J. All testing is performed pneumatically, except hydraulic testing For Type C testing of containment penetra-may be performed on isolation valve Type C tests tions, all testing will be done in the correct using water as a sealant provided that the system direction unless it can be shown that testing in line for the valve is not a potential containment the reverse direction is equivalent, or more con-atmosphere leak path, servative. The correct direction for this design is defined as flow from inside the Type C tests (like Type B test) are performed containrcent to outside the containment.

by local pressurization using either pressure decay or flowmeter method. The test pressure is 62.631 Acceptance Criteria applied in the same direction as when the valve is required to perform its safety function, un-The combined leal %c rate of all components less it can be shown that results from tests with subject to Type E and Type C (Subsection pressure applied in a different direction are 6.2.6.3) tests shall not exceed 60% of L. If equivalent or conservative. For the pressure de-repairs are required to meet this limit, t!"e re-cay method, test volume is pressurized with air sults shall be reported in a separate summary to or nitrogen to at least P. The rate of decay the NRC, to include the structural conditions of of pressure of the known te"t volume is monitored the components which contributed to the failure.

to calculate leakage rate. For the flowmeter methed, required pressure is maintained in the 62.6A Scheduling and Reporting of Periodic test volume by making up air, nitrogen or water Tests (if applicable) through a calibrated flowmeter.

The flowmeter fluid flow rate is the isolation The periodic leakage rate test schedules for valve (or Type B test volume) leakage rate.

Type A, B and C tests are described in Chapter 16.

All isolation valve seats which are exposed to containment atmosphere subsequent to a LOCA are Type B and C tests may be conducted at any tested with air or nitrogen at containment peak time during normal plant operations or during accident pressure, P.

shutdown periods, as long as the time interval between tests for any individual Type B or C e

MSIVs and isolation valves isolated from a tests does not exceed 2 years. Each time a Type O

scaling system will use a test pressure of at B or C test is completed, the overall total leak-least P.

age rate for all required Type B and C tests is updated to reflect the most recent test re-Those valves which are in lines designed to sults. In addition to the periodic tests, any be, or remain, filled with a liquid for at least major modification, replacement of component 30 days subsequent to a loss-of-coolant accident which is part of the primary reactor containment are leakage rate tested with that liquid. The boundary, or rescaling a seal welded door, per-liquid leakage measured is not converted to formed after the preopertional leakage rate test equivalent air leakage nor added to the Type B will be followed by either a Type A, Type B, or N

and C test total.

Type C test as applicable for the area effected by the modification. Type A, B and C test All test connections, vent lines, or drain results shall be submitted to the NRC in the lines consisting of double barrier (e.g. 2-valves summary report approximately three months after in series, one valve and a cap, or one valve and each test.

MM T 6,2,6,7.)

nllance msrecteel 3f cou ddalwn 3,a sf Thue Iwes are s wv 3 I d3 y intervel$ (internal and ekel to pna,9 uptew~ent aneno,em n rarekl ) as qusg by Julmfion.s 16 SR 3.4.1.33 y

a n d /6.S A 3. 6. I..? 4.

i I

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nA6tocAS MN Standard PlcpC _..._ 'POPRIETARYlhTORMAT10N Qass it!

Rev A

Response

valves are opened during power operation. Rese are air-operated valves with rapid closure times, present.

This requirement is not applicable to the ABWR.

ing little opportunity for substantial releases from the 1

It applies only to PWR-type reactors.

PCV in the eveat of a transient requuing containment isciation. Note that under the twWeni speciScations, 19A.2.26 Isolation Dependability [ Item (2) containment inerting and purging with the larger (xiv)]

ventilation lines is permitted during power operation above 15% for limited periods at either end of the NRC Position operating cycle. The process of purging the contain-Provide containment isolation systems that:

ment with air also serves to remove any potential

[II.E.4.2]

activity for AIARA considerations prior to actual personnel entry into the PCV.

(A) Ensure all non-essential systems are isolated automatically by the containment isolation The large ventilation valves will be tested regularly and after any valve maintenance to assure that closing

system, times are within the limits assured in the radiological (B) For each non-essential penetration (except it-design basis. (See Subsection 19A33) strument lines) have two isolation barriers in

series, 19A.2.28 Design Evaluator [ Item (2) (xvi)]

(C) Do not result in reopening of the containment NRC Position isolation valves on resetting of the isolation

signal, Establish a design cri:erion for the a!!owable number of actuation cycles of the emergency core (D) Utilize a containment set point pressure for ini-cooling system and reactor protection system consis-tiating containment isolation as low as is com-tent with the expected occurrence rates of severe over patible with normal operation, cooling events (considering both anticipated tran-sients and accidents). (Applicable to B&W designs (E) Include automatic closing on a high radiation only.) [II.E.5.1]

signal for all systems that prodde a path to the endrons.

Response

Response This requirement is not applicable to the ABWR. It applies only to PWR-type (B&W designed) reactors.

This item is addressed in Subsection 1A.2.14 19A.2.29 Additional Accident Monitoring 19A.2.27 Purging [ Item (2) (xy)]

Instrumentation [ Item (2) (xvii)]

NRC Position NRC Position Provide a capability for containment purg-Provide instrumentation to measure, record and ing/ venting designed to minimize the purging time readout in the control room: (A) containment pres-consistent with ALARA principles for occupational sure, (B) containment water level, (C) containment exposure. Provide and demonstrate high assurance hydrogen concentration, (D) containment radiation that the purge system will reliably isolate under acci-intensity (high level), and (E) noble gas effluents at dent conditions. [II.E.4.4j all potential, accident release pcitats. Prodde for con-tinuous sampling of radioactive iodines and particu-

Response

lates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and c n, at The ABWR primasy containment vessel (PCV) op-measure these samples. [II.F.1]

6.2,3 -J crates with an inert atmosphere. During normal op-eration, all large valves in containment ventilacion i

lines are closed Only small,2*, nitrogen makeup g enert % isola 6en rufbre d osk hhe "A"

Ar' evertresure pmfech'on,

^ * " * * *

• f m

23A610 Mil Standard Plant ne, a 9.1-15 and 9.1-26. (The SSE response is tuo times normal and abnormal storage conditions equal to or the OBE response).

less than 0.95 in the new fuel storage racks. To ensure design criteria are met, the following normal Verticalimpact analysis is required because the and abnormal new fuel storage conditions will be fuel assembly is held in the storage rack byits own analyzed:

weight without any mechanical holddcwn devices.

Therefore, when the downward acceleration of the (1) norrnal positioning in the new fuel array, and storage rack exceeds ig, contact between the fuel assembly and the storage rack is lost. Horizontal (2) eccentric positioning in the new fuel array impact analysis is required because a clearance exists between the fuel assembly and the storage rack walls.

The new fuel sgorage area will accommodate fuel (kg < 135 at 20 C in standard core geometry) with See Subsection 9.1.6.2 for COL license information no safety implications.

requirements.

9.1.13.2 Structural Design 9.1.1.1.7 (Deleted)

(1) The new fuel vault contains one or more fuel 9.1.1.2 Facilities Description (New Fuel storage racks which provides storage for fuel a Storage) maximum of 40% of one full core fuelload.

_(1) h !-'N nf-th: a M ap 9 6 'k (2) The new fuel storage racks are designed to be freestandinc i.e., no sukports+above the baseL%s -e n.+ Ms(9.L

r M ? S g-x 4 6 % " '

p s c s hb.t ib.s o en v% %

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(f) The new fuel storage racks are top entry racl's (3) The* racks incluoe individual solid tube storage 4

designed to preclude the possibility of criticalit, compartments which provide lateral restraints under normal and abnormal conditions. The over the entire length of the fuel assembly.

upper tieplate of the fuel element rests against the module to provide lateral support. The (4) The weight of the fuel assembly or bundle is lower tieplate sits in the bottom of the rack, supported axially by the rack lower support.

which supports the weight of the fuel.

(5) The racks are fabricated from materials used for (7) The rack arrangement is designed to prevent construction are specified in accordance with the 5

accidental insertion of fuel assemblies or latest issue of applicable ASTM specifications, bundles between adjacent racks. The storage rack is designed to provide accessibility to the (6) Lead-in guides at the top of the storage spaces fuel bail for grappling purposes.

provide guidance of the fuel during insertion.

(/) The floor of new fuel storage vault is sloped to (7) The racks are designed to withstand, while 6

a drain located at the low point. This drain maintaining the nuclear safety design basis, the removes any water that may be accidentally and impact force generated by the vertical free-fall unknowingly introduced into the vault. The drop of a fuel assembly from a height of 1.8 drain is part of the floor drain subsystem of the meters.

iiquid radwaste system.

g,d, (S) The rack is designed to withstand a pullup force (f) The radiation monitoring equipmen for the of 1814 kg and a horizontal force of 454 kg.

7 new fuel storage areas is described in Section There are no readily definable horizontal forces 24,- 1 z. 3 4 in excess of 454 kg and,in the event a fuel assembly should jam, the maximum lifting force 9.1.13 Safety Evaluat;on of the fuel-handling platform grapple (assumes limit switches fail)is 1351 kg.

l 9.1.13.1 Criticality Control (9) The new fuel storage racks require no periodic The design of the new fuel storage racks provides special testing or inspection for nuclear safety for an effective multiplication factor (k gg) for both purposes.

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(2) live loads - effect of lifting an empty rack The loads in the three orthogonal directions are during installation; considered to be acting simultaneously and are combined using the SRSS method suggested in (3) thermal loads - the uniform thermal expansion Regulatory Guide 1.92. The leads due to the OBE due to pool temperature changes; event are approximately 90% of those due to an SSE event, and allowable stress levels for OBE are 50% of (4) seismic forces cf OBE and SSE; SSE, therefore making the OBE event the limiting load condition except for stability, where SSE (5) accidental drop of fuel assembly from acceptance criteria of 67% of critical buckling maximum possible height 1.8 meters above strength is limiting.

rack; and Under fuel drop loading conditions, the acceptance must remain <0.95. The rack is designed such criterion is that, although deformation may occur, k (6) postulated stuck fuel assembly causing an l

upward force of 1361 kg.

should the drop of a fuel assembly damage the tubes The load combinations considered in the rack and dislodge a plate of poison material, the kdf l design are:

would stillbe <0.95 as required.

(1) live loads The effect of the gap between the fuel and the storage tube is taken into account on a local effect (2) dead loads plus OBE basis. Dynamic response analysis has shown that the fuel contacts the tube over a large portion of its (3) dead loads plus SSE; and length, thus preventing an overloaded condition of both fuel and tube.

(4) dead loads plus fuel drop.

The verticalimpact load of the fuel onto its seat is ]

Thermal loads are not included in the above considered conservatively as being slowly applied combinations because they are negligible due to the without any benefit for strain rate effects. See design of the rack (i.e., the rack is free to Subsection 9.1.6.7 for COL license information expand / contract under pool temperature changes),

requirements.

The loads experienced under a stuck fuel assembly 9.1.2.1.4 Thermal. Hydraulic Design condition are typically less than those calculated for the seismic conditions and, therefore, need not be The fuel storage racks are designed to provide included as a load combination, sufficient natural convection coolant flow to remove decay heat without reaching excessive water i

The storage racks are designed to counteract the temperatures (100 C).

tendency to overturn from horizontalloads and to lift from verti' cal loads. The analysis of the racks assurne6n adequate supporting structure, and loads

( Aeee gbnerated accordingly.

In the spent fuel storage pool, the bundle decay heat

]N-Stress analyses will be performed by the vec tor is removed by recirculation flow to the fuel pool using classical methods based upon shears a.

cooling heat exchanger to maintain the pool tempera-moments developed by the dynamic method. Using ture. Although the design pool exit temperature to the given loads, load conditions and analytical the fuel pool cooling heat exchanger is far below l methods, stresses will be calculated at critical boiling, the coolant temperature within the rack is sections of the rack and compared to acceptance higher depending on the naturally induced bundle criteria referenced in ASME Section III subsection flow which carries away the decay heat generated by l NF. Compressive s' ability will be calculated the spent fuel. The purchase specification for the fuel according to the AISl code for light gage structures.

storage racks require the vendor to perform the thermal-hydraulic analyses to evaluate the rate of naturally circulated flow and the maximum rack water exit temperature. See Subsection 9.1.6.7 for COL license information requirements.

9.1-a Amendment 23

23A6100AH Standard Plant wn compatible with the environment of treated water i provides the required dynamic stabilit

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and provides a design life of 60 years.

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(3) The racks include individual solid tube storage 9.1.2.2 Facilities Description (Spent Fuel compartments, which provide lateral restraints Stnrage) over the entire length of the fuel asse ubiv or

> I ra 33ItT S bundle.

(f) The spent fuel storage racks provide storage in the reactor building spent fuel pool for spent (4) The racks are fabricated from materials used for g

fuel received from the reactor vessel during the construction and are specified in accordance refueling operation. The spent fuel storage with the latest issue of applicable ASTM racks are top entry racks designed to preclude specifications at the time of equipment order.

the possibility of criticality under normal and i 4

abnormal conditions. The upper tieplate of the (5) Lead-in guides at the top of the storage spaces fuel elements rests against the rack to provide provide guidance of the fuel during insertion.

lateral support. The lower tieplate sits in the bottom of the rack, which supports the weight (6) The racks are designed to withstand, while of the fuel.

maintaining the nuclear safety design basis, the impact force generated by the vertical free-fall

(/) The rack arrangement is designed to prevent drop of a fuel assembly from a height of 1.8 6

accidental insertion of fuel assemblies or meters.

bundles between adjacent modules. The i

storage rack is designed to provide accessibility (7)i The rack is designed to withstand a pullup force to the fuel bail for grappling purposes.

f of 1814 kg and a horizontal force of 454 kg. l I

! There are no readily definable horizontal forces (3) The location of the spent fuel poolis shown

in excess of 454 kg and in the event a fuel l Section 1.2

/ assembly should jam, the maximum lifting force

~

/

of the fuel-handling platform grapple (assumes

[

limit switches fail) is 1361 kg-l 9.1.23 Saftty Daluation

[

The fuel storage racks are designed to handle onf(8) irradiated fuel assemblies. The expected 9.1.23.1 Criticality Control j

radiation levels are well below the design levels.

The spent fuel storage racks are purchasyd j The fuel storage facilities will be designed to 1

equipment. The purchase specification for the spept fuel storage racks will require the vendor to provide. Seismic Category I requirements to prevent the information requested in Question 430.190@ ; earthquake damage to the stored fuel.

criticality analysis of the spent fuel storage including -

the uncertainity value and associated probability and The fuel storage pools have adequate water l

l confidence level for the k value. See Subsection shielding for the stored spent fuel. Adequate 9.1.63 for iew4.a require *rkents.Mo ^

shielding for transporting the fuel is also provided.

l gcot b c*ws c. A<

j Liquid level sensors are installed to detect a low pool 9.1.23.2 Structural Design and Material water level, and adequate makeup water is available Compatibility Requirements to assure that the fuel will not be uncovered should a leak occur.

(1) The spent fuel pool racks provide storage for 270% of the reactor core.

Since the fuel storage racks are made of noncombustibic material and are stored under water, (2) The fuel storage racks are designed to be:

there is no potential fire hazard. The large water supported above the pool floor by a support -

volume also protects the spent fuel storage racks from structure. The support structure allows ;

potential pipe breaks and associated jet impingement sufficient pool water flow for natural con '

loads.

vection cooling of the stored fuel /Since IP' Fuel storage racks are made in accordance with the

{m'o'Eles are freestanding (i.e., no supports latest issue of the applicable ASTM specification at y above the base), the support structure also W

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s MM 23A6100All Standard Plant ne n 9.1.6 COL License Inforrnation operator qualifications, training and control pwgram.

9.1.6.1 New Fuel Storsge Racks Criticality Analysis 9.1.6.7 Spent Fuel Racks Structural Evaluation The COL applicant referencing the ABWR The COL applicant will provide the NRC design shall provide the NRC confirmatory critically confirmatory structural evaluation of the spent fuel analysigas required by Subsection 9.1.1.1.L racks as outlined in Subsection 9.1.2.13.

9.1.6.2 Dynamic and impact Analyses of New Fuel 9.L6.8 Spent Fuel Racks 'Rermal-Hydraulic Analysis Storag: Racks The COL applicant will provide the NRC The COL applicant shall provide the NRC confirmatory thermal-hydraulic analysis that evaluates confirmatory dynamic and impact analyses of the the rate of naturally circulated flow and the maximum new fuel storage racks. See Subsection 9.L1.1.6.

rack water exit temperature as required by Subsection 9.1.2.1.4 9.1.63 Spent Fuel Storage Racks Criticality Analysis 9.1.6.9 Spent Fuel Firrwater Makeup Provedures and Training The COL applicant shall provide the NRC confirmatory criticality analysis as required by The COL applicant will develop detailed Subsection 9.1.23.1.

procedures and operator training for providing firewater makeup to the spent fuel pool. (See 9.1.6.4 Spent Fuel Racks IAad Drop Analysis Subsection 9.133).

The COL applicant shall provide the NRC 9.1.7 References confirmatory load drop analysis as required by Subsection 9.L43.

1.

General Electric Standarditpplication for Reactor Fuel, (NEDE-24011-P-A, latest 9.1.6.5 New Fuel Inspection Stand Seismic approved revision).

Capability Tbc COL applicant will install the new fuel b rcl leks. t vodv P *d inspection stand firmly to the wall so that it does not

, p, (y mw m e

1. 'A(k M fallinto or dump personnel mto the spent fuel pool O ** f M S C (*j (

during an SSE. (See Subsection 9.1.4.23.2.)

s 9.1.6.6 Overhead Load IIandling System Information The COL applicant shall provide a list of all I

cranes, hoists, and elevators and their lifting capacities including any limit and safety devices required for automatic and manual operation. In addition, for all such equipment, the COL applicant shall provide: (1) heavy load handling system l operating and equipment maintenance procedures, (2) heavy load handling system and equipment maintenance procedures and/or manuals, (3) heavy load handling system and equipment inspr.ction and test plans; NDE, visual, etc., (4) heavy load handling safe load paths and routing plans,(5) OA program to monitor and assure implementation and compliance of heavy load handling operations and controls,(6) 9.1-13 Aniendment 26

\\

21A6100AT Standard Plant nrv n QUESTION 00.181 How is the new fuel proteced from intern Dy generated miuites and the effects of rnoderate or high energy piping or rotating machinery in the vicinity of the vault housing the new fuel storage rads. (9.1.1)

RESPONSE TO 00.181 R.., g. a s c -+ e A,. a dc e., pr mN '

• e =d h g " *g" g ' i ' i ' o The new fuel vault is located within the reactor building on the refueling floor ( see Figure 1.2-12).

There are no high energy or moderate energy pipes or rotating machinery located in the vicinity of the new fuel vault.

I QUESTION 00.182 Provide information on how the design of the new fuel storage facility complies with GDC 61, " Fuel Storage and Handling and Radioacthity Control." Identify the ventilation system provided to handle possibic release of radioactivity resulting from accidental damage to the fuel (note that ABWR SSAR 7.1 does not describe the radiation monitoring equipment for the new fuel storage area as stated in ABWR SSAR Section 9.1.1.2). (9.1.1)

% b4be S.%b - s s Prow g m e sd sds'd"^3I1.L RESPONSE TO 430.182 The reactor 6 ding HVAC system monitors the building exhausts for radioactivity. If radioactivity '

encountered, the system is isolated and the SGTS system will start operation. This prevents any possible release "j

rahrtWv from any fuel handline accident.

QUESTION 430.183 f(

Provide sufficient information and drawings to determine that the failure of non. seismic systems and structures in the vicinity of the new fuel storage facility can not cause an unacceptable increase in kg (9.1.1) 4 y gAAtbw 9. i. g. ':

RESPONSE TO 430.183 p, r>.t 4.,, +k a p - u p -

We new tuei swrage racery is located on the remciing iloor of tne reactor building (see Figure 1.2-1.

The reactor building is a Seismic Category I building protecting the new fuel from seismic events and externall r nerated missiles. There are no non. seismic systems in the vicinity of the new fuel storage facility.

QUESTION 430.184 Demonstrate that the analyzed impact of a fuel assembly, including its associated handling tool, dropped from a height of 6 feet bounds the range of all possible load drops from all possible heights. For additional guidance on the required bounding analysis, see SRP Section 9.1.2, Item III.2.e.(9.1.2)

RESPONSE TO 430.184 As discussed near the end of Subsection 9.1.4.3, light loads such as the blade guide, fuel support casting, control rod or control rod guide tube weigh considerably less than a fuel bundle and are administrative 1y controlled to eliminate the movement of any light load over the fuel pool above the elevation required for fuel assembly handling. Thus, the kinetic energy of any light load would be less than a fuel bundle and would have less damage induced.

QUESTION 430.185 Provide sufficient information and drawir.gs to determine that the failure of non-seismic systems and structures in the vicinity of the spent fuel storage facility can not cause an unacceptable increase in kg (9.1.2)

Amendment 16 20.3 353

Mhb 23A6100AT Standard Plant vm RESPONSE TO 00.185 pvamJd m e s cd 5d 3 g 9 1. 2. 2-.

% g m b tb ; pcs4% <J i12e spem fues storage poalis located on the refueling floor o1 the reaaor ouliL,;(wc Figuie 1.2-12).

~

e reactor building is a Seismic Category I building protecting the spent fuel from seismic events and externally generated rnissiles. There art no non-seismic systems in the vicinity of the spent fuel storage pool.

QUESTION C0.156 Provide drawbgs and information pertaini,g to spent fuel transfer canal capability of the fuel transfer canal or other provisions to prevent a dropped shipping cask from causing an unacceptable loss of pool water.(9.1.2)

O > ",)t,.m 3.1.L 2.

1 RESPONSE 00.186

% gw_ +. 4-k s ; p u b p v-,v w M e h Ec-v a

Ae shipping cask is placed in a walled off and drained portion of the spent fuel pool. The drained.

volume is flooded, and the Seismic Category I gates removed. The spent fuelis then transferred. This process is 3 reversed to remove the cask. The ratio of the two volumes is such that failure of the gates will not lower the water level enough to be unacceptable. Interlocks on the main crane prevent the shipping cask from being

'ed over any other portion of the spent fuel storage pool.

~

QUESTION 00.187 Clarify whether there is a) an interconnecting fuel transfer canal capable of being isolated from the fuel pool and adjacent cask loading area, and b) any high-energy piping or rotating machinery in the sicinity of the fuel storage pools. Also, clarify whether the racks are designed to preclude inadvertent placement of a fuel assembly in other than prescribed locations.(9.1.2) u s e - u %.r man p g "g %( ' d" *" 1 -

RESPONSE 430.187 3

(a)

As shown in Figure 1.2-12,'the spent fuel pool and adjacent cask loading area are separated by Seismic Category I gates. These gates isolate the cask loading area frors the spent fuel pool.

(b)

No high energy or moderate energy piping or rotating machinery are located in the sicinity of the spen fuel pool or cask loading area on the refueling flo,or..

QUESTION 00.188 Describe the function of the containment pool mentioned in ABWR SSAR Section 9.1.2.1.5. (9.1.2)

RESPONSE 40.188 Subsection 9.1.2.2(3) has been changed to omit mention of a containment pool.

QUESTION 00.189 What is the seismic category of the gates in the pools? (9.1.2) sd du s 9. L A 7 me

. + k.r F Pe gates tietween the spent fuel pool and other pools are all Seismic Category Amendment 16 20.3-154 l

g GE Nuclext Energy 4

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'5 ::,:.u w =s : ss n March 5,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - Resolution -

of Outstanding Items of Section 3.11 yfp A7pp g/gy

Dear Chet:

Enclosed are SSAR markups of selected portions of Section 3.11 supporting resolution of outstanding items. This enclosure replaces that provided in my February 25,1993 letter.

'Ihese rephcement markups incorporate the discussions of the GE/NRC phone call on March 4,1993 and the summary of Chapter 3 issues provided by Mike Janus on March 4, 1993.

It should be noted that this markup includes Pages 31.3-10 and 31.3-16 of an earlier submittal and the proprietary affidavit under which they were originally issued is applicable.

Please provide a copy of this transmittal to Butch Burton.

Sincerely, Y

Jack For Advanced Reactor Programs cc: Hal Careway(GE)

Norman Fletcher (DOE)

Bernie Genetti(GE)

JP93 48

23A61MAE 1

Seandard Pinnt ggy,

1 3.11 ENVIRONMENTALQUMtCW10N 3.11.1 EquipmeritIdentification and OF SAFETLRELATED MECG/ NNAL Envitunmental Conditions AND ELECIRICAL EQUIPMER11 Safety related electrical equipment within This section defines the environmental the scope of this section includes all three conditions with respect to limiting design categories of 10CFR50.49(b) (Reference 1).

conditions for all the safety related mechanical Safety related mechanical equipment (e.g.,

and electrical equipment, and documents the pumps, motor-operated valves, safety-relief qualification methods and procedures employed to valves, and check valves) are as defined and demonstrate the capability of this equipment to identified in Section 3.2.

Electrical and perform safety-related functions when exposed to mechanical equipment safety classifications are the anvironmental conditions in their respective further defined on the system design drawings.

locations. The safety-related equipment within the scope of this section are defined in Sub-Safety related equipment located in a harsh section 311.1. Dynamic qualification is environment must perform its proper safety j

addressed in Sections 3.9 and 3.10 for Seismic function during normal, abnormal, test, design Category I mechanical and electrical equipment, basis accident and post accident environments as applicabic. A list of all safety related elec.

respectively.

trical and mechanical equipment that is located j

Limiting design conditions include the in a harsh environment area will be included in the Environmental Oualification Document (EOD)

]

following:

to be prepared as mentioned in Subsection (1) Normal Operating Conditions - planned, 3.11.6.1. The COL applicant referencing the purposeful, unrestricted reactor operating ABWR design will provide a list of impacted modes including startup, power range, hot non safety-related control systems and the standby (condenser available), shutdown, and design features for preventing the potential adverse consequences identified in IE I

refueling modes; Information Notice 79-22, Qualification of 5

(2) Abnormal Operating Conditions - any Control Systems. The COL applicant will also W g, deviation from normal conditions anticipated address issues related to equipment wetting and f b;i j.

to occur often enough that the design should flooding above the flood level identified in IE f

include a capability to withstand the con-Information Notice 89-63, Possible Submergence '3ffhthe. !

ditions without operational impairment; of Electrical Circuits Located Above the Flood j

Level Because of Water Intrusion and Lack of (3) Test Conditions - planned testing including Drainage, as required in Subsection 3.11.6.

j j

pre operational tests; Environmental conditions for the zones where i

j (4) Accident Conditions - a single event not safety related equipment is located are cal-i reasonably expected during the course of culated for normal, abnormal, test, accident and plant operation that has been hypothesized post accident conditions and are documentel for analysis purposes or postylated from Appendix 31, Equipment Qualification Environ-unlikely but possible situations or that has mental Design Criteria (EQEDC). Environmenta i

conditions are tabulated by zones, contained in the potential to cause a release of radio-the referenced building arrangements. Typical active material (a reactor coolant pressure boundary rupture may quahfy as an accident; equipment in the noted zones is shown in the a fuel cladding defect does not); and referenced system PAID and IED design drawings.

Occurences of anticipated abnormal operating (3) Post-Accident Conditions - during the length conditions are similar to test conditions and of time the equipment must perform its their significant environments are comparable.

safety related function and must remain in a 4

safe mode after the safety related function is performed.

3111 j

Amchdment 24

. drilling holco in all cppropriato jene..l e boxw, terminal boxOo, pull boxas, condulets, and end-use equipment enclornras ins $de the drywell and the c stainment.

Tha Monticello plant found that a junction box for RHR pump motor leads con-tained several inches of water (NRC Inspection Report 50-263/87-013-DRS).

The box did not have a drain hole.

The licensec in'itially determined that the as-sociated conduits were routed through humid areas, which could have resulted N

89-63 September 5, 1989

)

Page 2 of 3 s

in condensation from the conduits accumulating in the box.

However, the licensee later postulatec that hosing down of. equipment in that area may have caused water to enter the box through unsealed openings.

In this instance, the circuits were found wet but not yet submerged in the accumulating water.

Tha licensea drilled weep holes in all appropriate motor-lead junction boxes and other enclosures to correct the problem.

During an inspection performed at Clinton Power Station from August 17 through August 21, 1987, NRC inspectors identified a terminal box without drain holes.

T** box was required to be environmentally qualified in accordance with the Although the box was located above the

.uirements of 10 CFR 50.49. it was subject to possible water and moisture ipostulated plant flood level, intrusion that could submerge the contents of the box in an accident.

Subsequently, the licensee identified 156 terminal boxes without drain holes, which could affect multiple safety systems.

The licensee drilled drain holes in the affected terminal boxes.

During a followup inspection performed from February 6 through February 24, 1989, the NRC identified six additional junction boxes requiring drain holes.

Several of these boxes contained taped olectrical splices which the licensee's environmental qualification program had not demonstrated to be environmentally qualified to perform their required function for the re p ired duration if they became submerged following a loss-of-coolant accident (LOCA).

Following this finding, the licensee identified numerous other enclosures with taped splices that required drain holes.

Discussion:

The NRC regulation pertaining to environmental qualification specifically regarding submergence is addressed in 10 CFR 50.49(e) (6), which states that

9 ) of 3 nA6t:x:AE Standard Plant pg Environmental parameters include temperature, analytical techniques in the derivation of pressure, relative humidity, and neutron dose envirournatal parameters, the number of units rate and integrated dosegadiation dose for} tested, production toli.rances, and test

- gamma and beta data for both normal and accident equipment inaccuracies conditions will be provided by the COL applicant in accordance with the requirements in ciubscen@

The environmental conditions shown in the d2.2.3.17he radiation requirements are site Appendix 31 tables are upper-bound envelopes y

" g, g' specific documentation owing to the need to model used to establish the environmental design and U#

specific equipment which is applicant quali-fication bases of safety-related

1. ~

< l determined. The HVAC detailed modeling and the equipment. The upper bound envelopes indicate evolving considerations in the area of accident that the zone data reflects the worse case source terms are expected to generate ' expected endronment produced by a compendium of l significantiv differing radiation reouirmentsj accident conditions. Estimated chemical Where applicable, these parameters are given in environmental conditions are also reported is terms of a time-based profile.

Appendix 3I.

The magnitude and 60-year frequency of occur-3.11.2 Quali!1 cation Tests and Analyses rence of significant deviations from normal plant environments in the zones have insignificant Safety-related electrical equipment that is effects on equipment total thermal normal aging located in a harsh environment is qualified by or accident aging. Abnormal conditions are test or other methods as described in IEEE 323 overshadowed by the normal or accident conditions (771)

~

in the Appendix 31 tables.

No4e : C o J a. hk u f m M Margin is defined as the diff erence between the W L.9-2.1.T 6 5 g

most severe specified service onditions of the P r.e dao Lb <~awhp' "

plant and the conditions used for qualification.

a.,el c.4wsss b e Margins shall be included in the qualification parameters to account for normal variatioas in (A hh

. c -O J commercial production of equipment and reasonable y

3 o g t, A s.

• "g ^ M. c errors in defining satisfactory performance. The AI M environmental conditions shown in the Appendix 3I S S A R-t tables do not include margins.

^

Some mechanical and electrical equipment may be required by the design to perform an intended safety function (SetwedUninutes of the occurrence - y), /n of the event but less'than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the Such equipment shall be shown to remain event.

functional in the accident environment for a

c7m period of at least I hour in excess of the time yg_j assumed in the accident analysis unless a time margin of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be justified.

Such justification will include for each piece of equipment: (1) consideration of a spectrum of breaks; (2) the potential need for the equipment later in the event or during recovery operations; (3) detemination that failure of the equipment i

after performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) determination that the margin applied to the minimum operability time, when combined with other test margins, will account for the uncertainties accociated with the use of f,

/*

b AfDCfWimCpt M

ABWR DELETE THM PAGF D*#

224sia Standard Plant f@/r/ 4FN#/x 3g Rev n applicable locations. Alternatinly, actual such locations will produce the maximum criti 1 multi-support excitation effects may be taken responses of the components. In-equip nt into account by performing a multi-support response spectra from time history response di be ememon analysis.

generated and be in accordance w' h the requirement specified in Paragraph 3K.5. a)(6).

(b When determining stresses, the effects of l

lative seismic support movements willbe 3K.4.2.42 QualI5 cation Determinati co idered. When these effects are consi-der significant, they may be obtained by Tbc equipment type will be idered qualified perio ing a static structural analysis of the by demonstrating that the equi ent performance j

system, ciuding anchor movements. Such will meet or exceed its specifi values for the most 1

effects ( hich are secondary) will be severe environment or seg nce of environments combined 'th primary (inertial) effects specified during the qual' ed life. An important i

using the SR step in this process will the determination that the qualification to t requirements adequately

{

3K.4.2.4.1.4.6J Time His ry Analysis envelops the equipme applications.

Time history analysis wi be performed when 3K.4.23 Combi Quall5 cation conditions arise invalidating t e response spectrum method of analysis due to nonii ar phenomena, or Equipment my be qualified by type test, when generation of in-equipmen response spectra analysis, pres ous operating experience, or any 1

or a more exact result is desired. o integrate or combination these three methods.

j differentiate, the analysis will done by an applicable numerical integration te nique. The 3K.4.2.6

-going Quall5 cation largest time step used in the analysis w be 1/10 of the period of the highest significan mode of equipment may have a qualified life less vibration of the equipment. The dynamic put will than e design life of a nuclear power generating be the time history modon at the equipment pport stat' n. The qualified life may be extended by location. For equipment supported at s eral i alling additional equipment of the same type in locations, the responses will be determine by

'ons where service conditions equal or exceed simultaneous excitations using appropriate ti e ose of the equipenent to be qualified, removing history input at each support location. The scale them after a planned period less than the previously time interval will be varied as per Paragraph qualified life and subjecting them to a type test 3K.S.2(a)(6). If the equipment frequency is wit '

qualification program. This test would include the range of the supporting structure, then a dditional accelerated age conditioning, dynamic, interval will be chosen such that the peak of e DBE tests. C@t6 of this type test extends response spectrum shall be at the equip at the ualified life of the installed equipment by the e

lengt of time simulated during age conditioning.

resonance frequency. The total time interval will be provided with the time history.

This p

. dure may be repeated until the qualified life equ is the required installed life of the 3 K.4.2.4.1.4.6.4 Generstlos of In-Egal equipme or the equipment is to be replaced Spectra before its q life is exceeded.

As a part of the dynamie qua fication of 3K.4.2.7 gj:,y equipment,in-egnipment response ta may be gg generated to quaEfy components o the equipment Margin is de as the difference between the dynamically. In-equipment C spectra will be most severe specfied rvice conditions of the plant obtained at criticallocations of components from and the coadirman used r qualification. Margins time. history analysis of the e uipment or,where will be included in the q ification parameters to appropriate methods are av ilable, by response account for normal varia 'ons in commercial spectra analysis. The in-equi est qualification plan production of equipment, asonable errors in shall identify the location t which in-equipment defining satisfactory performan response spectra will enerated and will prove etra generated at that the in-equipment te s

2 P

Nor : Gs PRSFsgr ro gsTu% -To N6 pp

( F RO M APPSW S K. ) I F -m N R C wm ygA v ACLC6 PTABt 6 f

4 "I

• S D EL E TE T Hl.$ P'A 6f gg wwAE FRom APPMDIX 3k Standnrd Plant Rev B Margins will be applied to the specified service con ' ions regardless of the qualification method select. The specific (quantified) margins applied will be amented for each phase of the qualifi-cation.

e levels of margia provided in Table 3K.4 2 are -onsidered appropriate for most applications.

r margins may be tued if justified as adequate fo the situation. In all cases the f

margins will be d cumented. Negative factors will be applied when lo cring the value of the service

/

condition increases t e severity. The application of margin to the age-con 'tioning of equipment will only consider, and con atively account for, any l

+

uncertainties in the process acceleration.

i Some mechanical and elec ' cal equipment may be required by design to perf an intended safety function between minutes of the currence of the event but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into he event. Such equipment will be shown to remain ional in the accident environment for period of at ast one hour in excess of the time assumed in the at analysis unless a time margin ofless than one h can be justified. Such justification will include r each piece of equipment:(1) consideration of a of breaks;(2) the potential need for the equip est later in the event or during recovery operations; ( a determination that failure of the equipment afte performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) determination that the margin applied to the minimum operability, time, when combined with-other test margins, will ace'ount for the uncertantics assoc 2ated with the use of~ analytical techniquu in the derivation of environmental parameters, the number of units tested, production tolerances, and test equipment inaccuracies.

b 5

i

/

%I3 '

Amenomem

-~

a

j,C 23AG10CAE Standard Plant g3 and permitted by 10CFR50.49(f) (Reference 1).

3.11.4 IAss of Heating, Ventilating, and Air Equipment type test is the preferred method of Conditioning qualification.

To ensure that loss of heating, ventilating, Safety-related mechanical equipment that is and air conditioning (HVAC) system does not located in a harsh environment,is qualified by adversely affect the operability of safety-analysis of materials data which are generally related controls and electrical equipment in based on test and operating experience, buildings and areas served by safety related HVAC systems, the HVAC systems serving these y

The qualification methodology is described invareas meet the single-failure criterion.

detail in thefNRC approved licensing TopicaD Section 9.4 describes the safety-related HVAC O,'

iLteport on GE'sCenvironmental systems including the detailed safety evalu-

!"/'p.J/ prograngReference 23. This% qualification also add-ations. The loss of ventilation calculations

'h-67d resses compliance with the applicaUe portions oft are based on maximum heat loads and consider the General Design Criteria,of 10CFR50, Appendix operation of all operable equipment regardless

/,MM M A, and the Ouality Assurance CriteriAof 10CFR50, of safety classification.

greps 4 Appendix B. Additionally, the(rebor[ describes

,"7 /M.d :7 IIN gg conformance to NUREG-0588 (Reference 3), and 3.11.5 Estimated Chernical and Radiation Regulatory Guidesjnd IEEE Standards referenced Environment in Section 3.11 of 'UREG 0800 (Standard Review Plan).

g f,gg 3.115.1 Chemical Environment Mild environment is that which, during or Equipment located in the containment drywell p n after a desi basis event (DBE, as defined in and wetwell is potentially subject to water 4xss4+(eferen_ce

, would at no time be significantly spray modes of the RHR system. In addition,

-E-+

more severe than that which exists during normal, equipment in the lower portions of the contain-test and abnormal events.

ment is potentially subject to submergence. The chemical composition and resulting pH to which safety-related equipment is exposed during normal operation and design basis accident conditions is reported in Appendix 31.

fw The COL applicant will requirejvenoors ol[ Sampling stations are provided for periodic equipment located in a mild environmentfo submit > analysis of reactor water, refueling and fuel a certificate of compliance certifying that the storage pool water, and suppression pool water equipment has been qualified to assure its to assure compliance with operational limits of required safety-related function in its the plant technical specifications.

applicable environment. This equipment is qualified for dynamic loads as addressed in 3.11.5.2 Radiatica Endronment Sections 3.9 and 3.10. Further, a surveillance and maintenance progr3m will be developed to Safety related systems and components are ensure equipment operability during its designed designed to perform their safety-related l tife. (See Sub<rreias 3.11.6).

function when exposed to the normal operational radiation levels and accident radiation levels.

3.11.3 Qualification Test Results Electronic equipment subject to radiation _

OPEN The results of qualification tests for exposure in excess of 1000 R andd!Fechan

[.//, j-[

safety related equipment will be documented, equipment in excess of 10,000 R will be maintained, and reported as mentioned in qualified in accordance with Reference 1.

S ubsection 3.11.6.

gher electnol Amendment M 3 13'2

1 E

(g & G' a A.B W 2246iocxe Standard Plant are s The normal operational exposure is based on the Non-safety-related control systems subjected radiation sources provided in Chapter 12.

to adverse environments will be evaluated for safety implications to safety related protective Radiation sources associated with the DBA and functions, and equipment wetting and flooding developed in accordance with NUREG-0588 above the flood level will be addressed in (Reference 3) are provided in Chapter 15.

accordance with Subsection 3.11.1.

Integrated doses associated with normal plant 3.11.7 References operation and the design basis accident condition for various plant compartments are described in (1) Code of Federal Regulations, Title 10, Appendix 31.

Chapter I, Part 50, Paragraph 50.49, Environmental Qualification of Electric 3.11.6 COL License Information Equipment Important to Safety for Nuclear Power Plant.

3.11.6.1 Environmental Qualification Document

?;/c h / 6TE T (2) encral Electric t.nvironmental Qualification The EOD shall be prepared summarizing the Program,NEDE-24326-1-P, Proprietary qualification results for all safety-related ocument. January 19R"4 equipment. The EOD shallinclude the following:

(3) Interim Staff Position on Environmental (1) The test environmental parameters and the Qualification of Safety Related Electrical methodology used to qualify the equipment Equipment, NUREG 0588.

l located in mild and harsh environments shall be identified.

(2) A summary of environmental conditions and qualified conditions for the safety-related equipment located in a harsh environment zone c hall be presented in the system com-ponent evaluation work (SCEW) sheets as STE T'

-u described in Table I-1gf GE's environmental ' e

"~~ ~"

qualification programuReferenceif'The SCEW sheets shall be compiled in the EOD.

g t.

(3) Equipment gama and beta radiation dose data for both normal and accident conditions will g3 ;

be provided in accordance with the, requirements ofCSubseerion 12.2.3.1Lg y g,o: J: 2

" q qf;,7,

-2 3.11.6.2 Environmental Qualification Records The results of the qualification tests shal*l be recorded and maintained in6n auditable rupG a ec.ov'c} s os c e toIfh IhG resyo'rbr25) 3.11.6.3 Surveillones, Maintenance and M /f e fuher/.U),

Erperience inforunation The COL applicant will requitecvenoop equipment certificates ot' qualification compliance and will develop a surveillance and maintenance program in accordance with Subsection 3.11.2.

3114 Amendment 24 i

Phff J

NRC Question: The staff noted in the DSER that the integrated gamma accident dose is in primary containment for the ABWR is given as 6 x 107 rads, which is less than the typical value of about 2 x 108 rads quoted in the safety analysis reports of several operating reactors (e.g. Perry: 2.7 x 108 : River Bend : 1.7 x 108 rads; Clinton: 2 x 10 rads' Nine Mile Point: 1.4 x 108 rads). It is not clear 8

why the ABWR integrated gamma accident dose is lower than the corresponding doses quoted for several operating reactors. GE's position which was provided in Section 5.3.2.1.5 of SSAR Amendment 15, did not adequately address this issue.

To resolve this issue GE must fully explain why the ABWR integrated gamma accident dose is lower than the corresponding doses quoted for several operating reactors. This is Open item 3.11.3-3.

,g,g Reply: The value of 6 x 107 rads originally reported in the ABWR SSAR is not the total integrated gamma dose for the primary containment for U.S. application but is the total integrated gamma dose as stipulated for the Kashiwaski 6/7 reactors being built and licensed under Japanese regulations. The difference between this value and the quoted existing U.S. reactors is one of philosophical approach between the two countries. To examine this difference we will compare the above ABWR calculation to that of the original TVA STRIDE design (BWR 6) which is shown in the following table.

TVA Stride Primary Containment integrated Gamma Dose Source Dose (rads cart >on) 100 % Noble Gases + daughters 2.54 x 107 airbome 50% Halogens + daughters airbome 7.55 x 107 25% Halogens + daughters plateout Wall Plateout 7.04x 10e Equipment Plateout 1.20 x 106 Total of Plateout 1.90 x 107 Total Dose 1.20 x 108 Noble Gas Dose in he ABWR, a preliminary calculation for the noble gases has been done and is shown in the attached figures. In all three figures we see that the integrated drywell dose approaches 1.2 x 107 rads in each of the three major primary containment volumes. In Stride, a single volume was used to contain all the fission product release whereas in ABWR there are three separately distinct volumes which are separated by meters of concrete. In ABWR it is possible that for a short time, on the order of hours, all the noble gases would be contained in a single volume, this would certainly not be the case for a 100 day evaluation. This short period containment has not been considered in the attached figures but will be in the final

If 2 ef f

^

+

evaluation. Therefore if a single volume (forcing aN the release into a

'I sngle volume for 100 days) were considered, the ABWR dose would increase an estimated factor of 2 to approximated 2.4 x 107 rads which is similar to Stride.

Halogen All'5orne Dose j

The airbome halogen dose is similar to the noble gas dose and for ABWR is roughly estimated at approximately 4 x 107 rads, again dividing the l

fisson product release between three compartments. In a similar fashion l

if a single compartment were considered the dose would be approximately 8 x 107 rads which is similar to the Stride value of 7.55 x 107 rads.

j Halogen Plateout Dose I

It is at this point that the K-6/7 analysis and the standard U.S. analysis differ. The standard U.S. analysis as is shown in the above table also consdors an additional 25% halogen plated out onto the containment surfaces. The K-6/7 analysis does not. No estimate exists yet as to what i

this factor will be on ABWR since it needs to be determined if the release would be dmded between the three volumes equally or by some mechanistic algorithm or whether all the release will be concentrated in a single volume. Nevertheless, it is not likely that ABWR will vary l

significantly from other analyses.

Conclusion I

The above discussion has attempted to describe the original basis and differences betwoon existing plant analyses and the value originalty found j

in the ABWR SSAR. As is shown, when the ABWR evaluation is compete, j

lt is not expected that the final ABWR will vary significantly from current plants.

(

i I

t l

l i

i

O Lower Drywell Integrated Dose from Noble Gas at various Leakages 1.0E+08

~~J

$ 1.0E+07:-

m'"~

~

- *

• d*"

at m

[

1.2%/ day g

/

/

/

I 1.0E+06 0 5

10 15 20 25 30 35 40 45 Time in days s

"i

..-..~.

Upper Drywell Integrated Dose

,5,from Noble Gas at various Leakages 1.0E+08 note.v.I

-J u

1.0E + 07-

=

-t

/

/

f 1.0E+06 0 5

10 15 20 25 30 35 40 45 Time in days 4

x

$s

.. ~.. - -..

.c t

Wetwell Integrated Dose fnygiNoble Gas at various Leakages 1.0E+08 so tw m

$ 1.0E+07 oc

,-~ g e

lause

/

- l 1.0E+060 5

10 15 20 25 30 35 40 45 Time in days l

q l

i rw

• \\

i I

\\

@ldi 23A6100AE Standard Plant PEV B I

2 Environmental parameters include temperature, analytical techniques in the derivation of I

pressure, relative humidity, and neutron dose environmental parameters, the number of units l

rate and integrated dosejRadiation dose for' tested, production tolerances, and test

~

Famma and beta dataTor both normal and accident ; equipment inaccuracies j

conditions will be provided by the COL applicant l

0462 7S @The radiation requirements are site]t in accordance with the requirements irk 5ubsectsp The environmental conditions shown in the Appendix 31 tables are upper-bound envelopes 1

fahah#' specific documentation owing to the need to model ! used to establish the environmental design and

' ?'ntec#or specific equipment which is applicant quali-fication bases of safety-related j

\\

l determined. The HVAC detailed modeling and the equipment. The upper bound envelopes indicate evolving considerations in the area of accident that the zone data reflects the worse case x.\\

source terms are expected to generate, expected environment produced by a compendium of (pfN i,s,_ignificantly differing radiation requirment) accident conditions. Estima':cd chemical y*3 4 Where applicable, these parameters are given in environmental conditions are also reported is terms of a time. based profile.

Appendix 31.

$EE SMM The magnitude and 60-year frequency of occur-3.11.2 Qualification Tests and Analyses If4 i-@ rence of significant deviations from normal plant environments in the zones have insignificant Safety-related electrical equipment that is effects on equipment total thermal normal aging located in a harsh environment is qualified by

)

1 or accident aging. Abnormal conditions are tes or other methods as described in IEEE 323 overshadowed by the normal or accident conditions y 94 in the Appendix 31 tables.

Margin is defined as the difference between the most severe specified service conditions of the plant and the conditions used for qualification.

Margins shall be included in the qualification j

parameters to account for normal variations in j

commercial production of equipment and reasonable l

errors in defining satisfactory performance. The i

environmental conditions shown in the Appendix 31 l

tables do not include margins.

Some mechanical and electrical equipment may l

be required by the design to perform an intended safety functionfetGe"Sthinutes of the occurrence yg l

of the event but less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into the event. Such equipment shall be shown to remain functional in the accident environment for a period of at least 1 bour in excess of the time assumed in the accident analysis unless a time margin of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be justified.

i Such justification will include for each piece of I

equipment: (1) consideration of a spectrum of i

breaks; (2) the potential need for the equipment later in the event or during recovery operations;

.l (3) detemination that failure of the equipment after performance of its safety function will not be detrimental to plant safety or mislead the operator; and (5) determination that the margin l

applied to the minimum operability time, when combined with other test margins, will account for the uncertainties accociated with the use of l

W 3I l

Amendment 24 i

~

t Gel Dectue company I

MkN PROPRIETC(WFORMATION DA6100AE l

Standard Plant a=m Table 31.315 l

nermodynamic Environment Conditions Inside Reactor Building t

(Secondary Containment)

Plant Accident Conditions j

(a) Pressure, temperature and relative humidity l

l Plant Zonefrypical Equipment 100 100 66 66 i

Control rod drive hydraulic Temperature ( g )

system (scram etc. of hydrau-Pressure (Kg/cm g 0.035 0.035 0.035 0

l lic control unit) [ Fig's.1.2-4 Humidity (%)

Steam Steam 100 90 max

/4.68 Time (2) 1(h) 6(h) 12(b) 100(day) i 171 100 66 66 Temperature ( g )

MS isolation valve (1)

Pressure (Kg/cm g 0.035 0.035 0.035 0

t MS drain isolation valve I

Humidity (%)

Steam Steam 100 90 max l

l Nitrogenlineisolationvalve (1),(4)

Time (2) 1(h) 6(h) 12(h) 100(day)

Process water line isolation valve (1),(4)

I

[ Fig's.1.2-2,1.2 3,1.2-3a, l

5.1-3J t

171 100 66 66 4

Temperature ( g )

Feedwater isoldion valve (1)

Pressure (Kg/cm g 0.035 0.035 0.035 0

[ Fig's.1.2-2,1.2-3,1.2-3a/

i 5.1-3]

Humidity (%)

Steam Steam 100 90 Max.

Tune (2) 1(b) 6(h) 12(h) 100(day) 171 100 66 66 RCIC injection valve (1), check Temperature ( g )

valve (inside MS tunnel), steam Pressure (Kg/cm g 0.035 0.035 0.035 0

line isolation valve [ Fig's.

Humidity (%)

Steam Steam 100 90 Max.

1 1.2 2,1.2-3,1.2-3a/5.4-8]

Tune (2) 1(h) 6(h) 12(h) 100(day) l 100(3) 66 66 RCIC (valve exceptisolation Temperature ( g )

valve, assemblies, cable, Pressure (Kg/cm g 0.035 0.035 0

turbine) [ Fig's.1.2-4/

Humidity (%)

Steam 100 90 Max.

5.4-8]

Tune (2) 6(h) 12(h) 100(day) 100 66 66 Temperature ( g )

RCIC turbine electne control t

Pressure (Kg/cm g 0.035 0.035 0

system (3),(6)[ Fig's.1,2-5/

5.4-8]

Humidity (%)

. Steam 100 90 Max.

l I

l Time (2) 6(h) 12(h) 100(day) i 4*

100 66 66 n

RHR (LPF1, cooling system at Temperature ( g )

t S/D, containment coolmg.Ser-Pressure (Kg/cm g 0.035 0.035 0

vice water system) valve, pump Humidity (%)

Steam 100 90 Max.

l (motor, seal cooler) instrument Tune (2) 6(h).

12(h) 100(day) y control electric equipment (in-ciuding cabic and sources of electricity)[ Fig's.1.2-4/

5.4-10]

f 31.V. 16 Amendment 21 i

i i

s

,a-e

General seeme compny MM PROPRIETARYINFORMATION 23A6100AE Standard Plant cu m

%8 Table 31.3-9 Radiation Environment Conditions Inside Primary Containment Vessel Plant Normal Operating Conditions i

(b) Radiation environment Number Plant Zone / Typical Operating dose rate lategrated dose (2)(3)

I}II Equipment Gamma Beta Neuty and Neutros Duence (R/h) (R/h) (N/cm -sec)

Gamala Beta Nectrop (R)

(R)

(N/ca')

I b-1 Upper drywell area 6x10 1x10 l

[ Fig's.1.2-3/ 5.13) 0 b-2 Upper area oflower 2x10 5x10 t

drywell l

[ Fig's.1.2-3a/ 5.13]

b3 Lower area oflower 1x10 3x10 drywe!!

[ F i g's. 1.2-3 b /

11.2-2]

b.4 Wetwell area (sup-8x10 2x10 pression pool and air space)

[ F i g 's. 1.2-3 c /

6.2-39, 7.6-11]

Notes:

(!) Operating dose nue is at 100% ratedpower asd awayfrom the rehntum sowce.

(2) Integrated dose means the integrated wdue over 60 years.

(3) The gamma and beta doses A beprended - acew=Ne rednanon sowce terms h*r~ne defliTef by the applicant referencing the ABWR design in L

GC

$ th* It9WEments Of Secaun 12.2.M n

y

^

N.

^^'8' J-1 _.L _

s

. _ en

.K.W*,

9

~, r. v a s w 1 < v " T ' ' ' "' * ' = &p '

vo

~

3IN Amendment 21

s 23A6100AE Standard Plant I

Table 31.310 Radiation Environment Conditions Inside Reactor Building (Secondary Containment)

Plant Normal Operating Conditions (b) Radiation environment IIN )

f N3)

Plant Zone / Typical Operating dose rate Integrated done l

Equipment Gamma (R/h) Beta (R/h)

Gamma (R) Beta (R) i General floor area (not otherwise r

noted)/Similar equipment RHR room [ Fig's. 1.2-4/5.4-10]

RCIC room [ Fig's.1.2-4/5.4-8)

HPCF room [ Fig's. 1.2-4/6.3-7]

SGTS room (Fig's. 1.2-10/6.5-1]

MS tunnel room (Fig's.1.2-8/5.1-3]

Divisionalvalve room

[ Fig's.1.2-8/ECCS]

Instrument rack room -

[ Fig's.1.2-6/ECCS}

Notes:

(1) Operating dose rate is at 100% ratedpower and awayfrom the radiation source.

l (2) Integrated dose means the integrated value over 60 years.

y

- ~ -. -.

(3) The gamma and beta doses will be provided by th plicant referencing the ABWR design in accordance with the requuements ofL5sbsection 12.23.

Pll

?,,, ? ?

rvvi i

i l

31.111 i

h Amendment 22 i

/

I ABWR Standard Plant

• '^'C i

Table 31.311 i

Radiation Environment Conditions Inside Reactor Building (Outside Secondary Containment)

Plant Normal Operating Conditions (b) Radiation emironment Plant Zone / Typical Operating dose rate lategrated dose (2M3)

IIN I Equipment Gamma (R/h) Beta (R/h)

Gamma (R) Beta (R)

Clean zone outside secondary cont-ment area (not otherwise noted)

(Fig's. 6.2-26/6.7-11 Monitor room [ Fig's. 1.2-8/6.51]

Notes:

(1) Opemting dose mte is at 100% ratedpower and awayfrom the radiation source.

l (2) Integrated dose mewss the integrated value owr 60 years.

p (3) The gamma and beta doses will be provided by the applicant referencing the ABWR design in acccedance mth the requuements offiubsecaon ILL3.h>2

, n _. a cJ. +- -

C,,.., 3. f g, A

_- -- - e

---< e vi,

i i

I s

i e

31 } 10 Amendment 22 e

I ABM u sioast Standard Plant g,. a l

Table 3I.3-12 Radiation Environment Conditions Inside Control Building t

i Plant Normal Operating Conditions (b) Radiation emironment IIN3I Plant Zone / Typical Operating dose rate Integrated dose (2M3)

Equipment Gamma rays (R/h) Beta (R/h)

Gamma Rays (R) Beta (R) i RCW pump and heat exchanger room

[ Fig's 1.2-15/9.2-la]

Main control, computer, battery and i

HVAC rooms (Fig's 1.2-15/18C.7-1]

Notes:

.}

(1) Operating dose rate is at 100% ratedpower and awayfrom the radiadon source.

(2) Integrated dose means the integrated value over 60)<ars.

_?d f). The gamma and beta doses will be provided by the applicant referencing the ABWR design inl accordance with the requirements ofFubsecnon 12.2.3)', b,4'i $i; W 3-? L, $':E ')

",. /

~ %

t I

s I

i Amendment 22

ABWR

• • '[7 Standard Plant 3

Table 31.3.II Radiation Environment Conditions Inside Turbine Building Plant Normal Operating Conditions (b) Radiation environment Plant Zone / Typical Operating dose rate lategrated dose (2H3)

{

I"I Equipment Gamma rays (R/h) Beta (R/h)

Gamma Rays (R) Beta (R) 2 Main steam stop valve area

[ Fig's 1.2 25/7.2 9) l Notes:

(1) Operating dose rate is at 100% ratedpower and awayfrom the radiation source.

(2) Integrated dose means the integmted vahte owr 60 years.

l(3) The gamma and beta doses will beprovided by the applicant referencing the ABWR design in accordance with the requirements ofrauosecuan 12 z < 12

-,/

e D,.'r L,_A A

.n...

I

)

l l

31118 Amendment 22 1

l ABM uAMMAE l

Standard Plant we Table 3L319 Radiation Environment Conditions Inside Primary Containment Vessel Plant Accident Conditions (b) Radiation environment I II I 14CA( }(

1stegrated dose l

Plant Zooc/ Typical Equipment Gamana rays (R/h) Beta (R/h)

Gamma rsys (R) Beta (R)

Primary containment vessel

[ Fig's. 6.2-26/5.1-3, 6.2-39,7.6-11, 11.2 2]

b t

\\

Notes:

(1) Assumes that 100% of the inert gases 50% of Halogen and 1% of the solidfission products are releasedfrom she cont 6mng LOC 4.

(2) Theinterated dateisfor 6 mands.

p :

(3) The gamma and bets doses will be_ provided by the applicant referencing the ABWR design in accandance mesh she - * -J C"~ 122IDL p,4 C ? ?L, ? A :No,#, W K h' i

i 1

I r

3OU Amendment 22

i i

MM 1

4 Standard Plant

• • ^'U j

Table 31.3 20 j

Radiation Environment Conditions Inside Reactor Building (Secondary Containment) Plant Accident Conditions 1

i I

i (b) Radiation environment NO WG j

Plant Zone / Typical 14CA Inesysted dose Equipment Gamma rays (R/h) Beta (R/h)

Gamma rays (R) Beta (R)'

i l

General floor area (not otherwise noted)/Similar equipecat j

RHR room [Fgs.1.2-4/5.4-10]

li RCIC room (Fig's.1.2-4/5.44]

HPCF room [ Fig's. 1.2-4/6.3-7]

SGTS room [Fgs.1.210/6.5-1]

t f

MS tunnel room

[ Fig's.1.24/5.1-3]

l t

Dmssonalvalve room

[ Fig's.1.24/ECCS]

Instrument rock room

[Fgs.1.24/ECCS]

l l

Notes:

(1) Assumts that 100% of she inert gases,.50% of Helops end 1% of she solidfission producu are rel*==*dpom she core enowng LOC 4.

e t

(2) Theinnerened doseisfor6 monAs.

(,,,

geninse end btSC doses anill bt prOVided by tht eftent referencing the ABW}t dt2ign iN h

(.ccordane,i.m she reprimaner er ~

, ?":O': f - t h &%. "~~ '

b 13^

D,%

a t

i k

f f

31.3 2.3 Amendment 22 i

b

- l

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t ABMR

• • [^E Standard Plant Table 3I.3 21 Radiation Environment Conditions Inside Reactor Br.ilding (Outside Secondary Containment) Plant Accident Conditions (b) Radiation environment LOCA('(

lategrated dm (2)(3)

Plant Zone / Typical Equipment Gamana rays (R/h) Beta (R/h)

Gamma rays (R) Beta (R) l Ocan zone outside secondary containment area (not otherwise l

noted) [ Fig's. 6.2-26/6.7-1]

l Monitor room [ Fig's.1.2-8/

6.5-1]

Notes:

(1) Assumes that 100% of the inert gases, 50% of Halogen and 1% of the solidpssion products are releasedfrom the cae Ausng LOCA.

(2) neinsepaneddoneisfor 6 monAt p

T

,[ne gamme and bets doses will be provided by the appliennt referencing the ABWR design n' sceorduce with de requuements of Subssenos 12.1.UF2-

\\

L A l L %

t" h 2.

y ~n

,m. w.

i

[

I l

l l

3I M '

Amtnoment 22

4 ABM -

n1e Standard Plant l

n,..

i Table 31.3 22 i

Radiation Environment Conditions Inside Control Building Plant Aulaent Conditions 1

(b) Radiation environment Plant Zooe/ Typical LOCA(

I II3I lategrated dose Equipenest h=== rays (R/h) Beta (R/h)

Gamans rays (R) Beta (R) j RCW pump and beat exchanger rooms l

[ Fig's.1.2 15/9.2-la]

Main control, computr.r, battery l

and HVAC rooms [ Fig's.1.2-15/

18C.71]

1 0

Notes:

(1) Assumes that 100% of the inerr gases,30% of Halogen and 1% of the solidfission products are releasedpom the cars durung LOCA.

(2) The inseynted dose isfor 6 monks.

i

= - _,.

~..

'~(3)

The pnmms and bets doses wi!! be provided by the sp licant referencing the ABW}t design m

/

h accordence with the requuwments of.5kbsecaon 12.2-w D;*: L n,.A' ?e Rv, hA h.s

~ ~ _ _ _

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3IM i

Amendment 22 f

ABWR 2mioac Standard Plant nev e for plant personnel, limit offsite releases of airborne 1.2.2.16.6 Fire Protectian System contaminants.

The fire protection system is designed to pro ide The following environmental systems are an adequate supply of water or chemicals to points provided:

throughout the plant where fire protection is required. Diversified fire-alarm and fire-(1)the control room air conditioning system suppression types are selected to suit the particular consisting of supply, recirculation / exhaust and areas or hazards being protected. Chemical fire makeup air cleanup units to ensure the fighting systems are also provided as additions to habitability of the control room under normal or in lieu of the water fire fighting systems.

and abnormal conditions of plant operation; Appropriale instrumentation and controls are provided for the proper operation of the fire (2) the reactor building secondary containment detection, annunciation and fire fighting systems.

HVAC system maintains a negative pressure in the secondary containment under normal and 1.2.2.16.7 Floor Leakage Detection System abnormal operating conditions thereby isolating the environs from potentialleak sources. This The drainage system is also used to detect system removes heat generated during normal abnormalleakage in safety related equipment plant operation, shutdown, and refueling periods; rooms and the fuel transfer area.

i (3) the drywell cooling system to remove heat from 1.2.2.16.8 Vacuum Sweep System the drywell generated during normal plant operations including startup, reactor scrams, hot A portable, submersible. type, underwater standby, shutdown, and refueling periods; vacuum cleaner is provided to assist in removing crud and miscellaneous particulate matter from the (4) the power block pressure control supply and pool floors or reactor vessel. The pump and the exhaust system to distribute air so that a negative filter unit are completely submersible for extended pressure is maintained in the emergency core periods. The filter ' package" is capable of being cooling equipment rooms, thereby isolating the remotely changed, and the filters will fit into a potential airborne contamination in these rooms; standard shipping container for offsite burial.

(5) the electrical equipment supply and exhaust 1.2.2.16.9 Decontamination System system to pressurize the electrical rooms allowing exfiltration of air to the battery rooms for exhaust The decontamination system provides areas, to the outside atmosphere; equipment and services to support low radiation level decontamination activities. The services may (6) the power block exhaust system to maintain the include electrical power, service air, demineralized refueling floor at a negative pressure with respect water, condensate water, radioactive and nonrad.

to the outside atmosphere to prevent the ioactive drains, HVAC and portable shielding.

potential release of airborne contamination; 1.2.2.16.10 Reactor Building (7) the diesel generator area air exhaust system to provide cooling during o;ieration of the diesel The reactor building includes the containment, generators. A tempered air supply system drywell, and major portions of the nuelcar steam controls the thermal emironment when the diesel supply system, steam tunnel, refueling area, diesel generators are not operating; and generators, essential power, non-essential power, emergency core cooling systems, HVAC and (8) coolers in the steam tunnel and ECCS rooms to supporting systems; remove heat generated during operation of the equipment in these rooms.

1.2.2.16.11 Turbine Building i

m DP 7 /, 2. 2. / 6. f The turbine building houses all equipment associated with the main turbine generator. Other auxiliary equipment is also located in this building.

l Amendment 20 1.2 16.6

i s

II I

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lWSen7" L2 2,/6 b' i

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ABM 234stooxs f

Standard Plant REV.B i

TABLE 3.2-1 j

CLASSIFICATION

SUMMARY

(Continued)

Table Table t

3.2-1 MPL 3.2 1 MPL

[

Item No.

Number **

Tilt item No.

Number **

M.

R Station Electrical Systems (Continued)

T5 T25 PCV Pressure and Leak Testing l

Facility R11 R40 Combustion Turbine Gncrerator T6 T31 Atmospheric Control System l

R12 R42 Direct Current Power Supply

• T7 T41 Drywell Cooling System R13 R43 Emeregncy Diesel Generator System

• T8 T49 Flammability Control System l

r R14 R46 Vital AC Power Supply 19 T53 Suppression PoolTemperature l

Monitoring System

• i l R15 R47 Instrument and Control Power Supply U

Structures and Servicine Snterns j

l R16 R51 Communication System UI U21 Foundation Work j

R17 RS2 Lighting and Servicing Power U2 U24 Turbine Pedestal Supply l

U3 U31 Cranes and Hoists S

Power Transmission Systems i

U4 U32 Elevator i

S1 S12 Reserve Transformer US U41 Heating, Ventilating and Air '

Conditionin

• 5"'Nff NdIWS T

Containment and Environmental Control fd$4 bit *'

I USl 00-Systems O'#7 COMeM > rid /f55"'

U6 U43 Fire Protection System l

70,f rfo T1 T11 Primary Contammen vs Vt/ J' U7 U46 Floor Leakage Detection T2 T12 Containment Internal System Structures 1

U8 U47 Vacuum Sweep System T3 T13 Reactor Pressure Vessel Pedestal l

l T4 T22 Standby Gas Treatment System

• These systems or subsystems thereof, have a primary function that is safety-related. As shown j

in the balance of this Table, some of these systems contain non-safety related components and, i

conversely, some systems whose primary functions are non-safety related contain components that have been designated safety-related.

~

t Master Pans List Number designatedfor the system i

I Amendment 23 32'D i

l l

i

?

e ABWR 234.ioase Standard Plant REv s TABLE 3.21 CLASSIFICATION

SUMMARY

(Continued).

Quality i

Group Quality Safet4 Loca-Classi-Assurance Seismic Principal Component Class llan fication Requirement

• Catemory Npica US Heating, Ventilating, and Air Conditioning Systems * (Continued)

{

g.

Valves and Dampers-2 SC,RZ B

I

-l secondary containment isolation B

I j

h.

Other safety-related 3

H,Z valves and dampers

i. Electrical modules with 3

SC,RZ B

1 safety related function H,X

)

j. Cable with safety-related 3

SC.RZ B

I j

function H,X f

2. Non-safety related equipment **

i i

a.

HVAC mechanicalor N

SC,RZ,H -

E electrical components X,W,T 4

with non-safety related 5

functions (t)(u) b.

Non-safety related fire N

SC,RZ,H, -

E protection valves and X,W,T dampers j

7 U6 Fire Protection System

1. Piping including supports and 2

C B

B' I

l valves forming part of the primary wntainment boundary l

I (t) (u) l

2. Other piping including supports N

SC,C,X D

E and valves RZ,H,T, i

W,0 i

1

. t) (u)

(

3. Pumps N

F D

E (t) (u)

4. Pump motors N

F E

• Includes thennaland radiological environmental controlfunctions within the ABHR Standard Plant scope.

" Controls emironment in rooms or areas containing non-safety related equipment within the ABKR Standard Plant.

yy 7,Toble andfuihry vaktfprNur I

[Y 3.2-29

~'

~~~

Amendment 24 h

Snwf A frwff. 3.ul.2

I ABM 23xamre Standard Plant arv A l

TABLE 3.2-1 CLASSIFICATION

SUMMARY

(Continued)

Quality Group Quality Safetg Loca-Classi-Assurance Seismic Princinal Comnonent' Class 112a fication Requirement

• Cateson Notes K1 Radweste System 1.

Drain piping including supports N ALL D

E (p) and valves - radioactive (exce yt Peie5 2$.. _.. _.

N g

2.

rain pipingincluding supports N ALL D

E and valves - nonradioactive (Wee //' C,/n) 3.

Piping and valves -

2 C,SC B

B I

N,%

P containment isolation p0" y b y

4 Piping including supports N

C,SC B

B 1

Y

'F and valves forming part of containment boundary (p)

E 5.

Pressure vessels including N

W supports E

(p)-

6.

Atmospheric tanks including N

C,SC,H, supports T,W (p)

E 7.

0-15 PSIG Tanks and supports N

W E

(p) 8.

Heat exchangers and supports N

C,SC,W (p)

E 9.

Pipingincluding supports N

C,SC,H, and valves T,W l

10. Other mechanical and N

ALL E

E (p) electrical modules l

11. ECCS equipment room N

SC C

B I

sump backflow protect-ion check valves NI Turbine Main Steam System 1.

Deleted (See B2.5) 3.2-21.2 Amendment 26

ABMT 23^um^u Standard Plant Rev n 9.2 WATER SYSTEMS fixtures located in areas with no sources of potentially radioactive wastes and conveys them 9.2.1 Station Service Water System to a sewage treatment facility.

The functions normally performed by the (4) The PSW includes a sewage treatment system station service water system are performed by the which treats sanitary waste using the activated systems discussed in Subsection 9.2.11.

sludge biological treatment process. The aeration tanks are capable of receiving waste at 9.2.2 Closed Cooling Water System a rate between 12,000 gpd and 48,800 gpd.

The functions normally performed by the closed (5) The PSW system shall be designed with no cooling water system are performed by the systems interconnections with systems having the discussed in Subsections 9.2.11, 9.2.12, 9.2.13, potential for containing radioactive materials.

and 9.2.14.

Protection shall be prodded through the use of air gaps, where necessary.

9.2.3 Demineralized Water Makeup System 9.2.43 System Description (Conceptual Design) wledn The functions normally performed by the demin-The PSW system 6composedNpotable water eralized water makeup system are performed by system, a sanitary drainage system and a sewage the systems discussed in Subsections 9.2.S,9.2.9 and treatment system.

9.2.10.

9.2.43.1 Potable Water System 9.2.4 Potable and Sanitaty Water System (PSW)

Filtered water flows by gravity from the filtered water storage tank of the MWP system into a potable This subsection provides a conceptual design of water storage tank. A hypochlorite addition pump the potable and sanitary water (PSW) as required and tank are provided which adds sodium by 10CFR52. The interface requirements for this hypochlorite to the water entering the potable water system are part of the design certification. A storage tank. Two potable water pumps send water separate portion of the PSW, the non radioactive from the potable water storage tank to a drains is described in Subsection 933.

hydropneumatic pressure tank. A hydropneumatic pressure tank and air compressor are provided to 9.2.4.1 Safety Design Bases (Interface maintain adequate pressure within a potable water Requirements) distribution piping system. Potable water is sent to a heater where it is heated and distributed throughout The PSW system has no safety-related function.

the plant.

Failure of the system does not compromise any safety-related system or component, nor does it 9.2.43.2 Sanitary Drainage System prevent a safe shutdown of the plant.

The sanitary drainage system collects liquid wastes 9.2.4.2 Power Generation Design Bases (Interface and conveys them to the sewage treatment system.

Requirements)

This system is installed in accordance with ANSI A40.8, National Plumbing Code, and applicable local (1)The PSW system is designed to provide a or state codes.

minimum of 200 gpm of potable water during peak demand periods.

9.2.433 Sewage Treatment System (2) Potable water is filtered and treated to prevent The sewage treatment system is a concrete harmful physiological effects on plant structure containing several compartments. The personnel.

sewage treatment systems uses the activated sludge biological treatment process. The system includes a (3)The PSW includes a sanitary drainage system comminutor with a bypass screen channel, two which is designed to collect liquid wastes and aeration tanks, three final clarifiers, one chlorine entrained solids discharged by all plumbing contact tank, two aerobic digesters, three air blowers, 921 Amendment

~

ABWR nuuma Standard Plant wn SECTION 9.3 CONTENTS Ssetion Title Pace 933 Eouipment and Floor Drainare Systems 93-2 933.1 Non-Radioactive Drains 93-2 933.1.1 Safety Design Bases 93-2 933.1.2 Power Generation Design Bases 93-2 933.13

System Description

93-2 933.1.4 System Operation and Component Description 93-2 933.1.5 Safety Evaluation 93-2.1 933.2 Non-Radioactive Drain Interface 93-2.1 933.2.1 Safety Design Bases (Interface Requirement) 93-2.1 933.2.2 Power Generation Design Bases (Interface Requirement) 93-2.1 933.23 System Description (Conceptual) 93-2.1 933.2.4 Safety Evaluation (Interface Requirement) 93-2.1 u

933.2.5 InstrumentationQd Alar]m (Interface Requirement) 93-2.1 93.4 Chemical and Volume Control System MVR) 93-2.1 93.5 Standby Llauld Control Svstem 93-2.1 93.5.1 Design Bases 93-2.1 93.5.1.1 Safety Design Bases 93-2.1 93.5.2

System Description

93-2.2 93.53 Safety Evaluation 93-4 93.5.4 Testing and Inspection Requirements 93-6 93-iii Amendment

ABWR muun Standard Plant nev n SECTION 9.3 CONTENTS (Continued)

Section Title Page 933 Eautoment and Floor Drainace Svstems 93-2 93.53 Instrumentation Rcquirements 93-6 93.6 Instrument Air System 93 7 93.6.1 Design Bases 93-7 93.6.1.1 Safety Design Bases 93-7 93.6.2

System Description

93-7 93.63 Safety Evaluation 93-7.1 93.6.4 Inspection and Testing Requirements 93-7.1 93.6.5 Instrumentation Application 93-7.1 93.7 Service Air System 93-8 93.7.1 Design Bases 93-8 93.7.1.1 Safety Design Bases 93-8 93.7.1.2 Power Generation Bases 93-8 93.7.2

System Description

93-8 93.73 Safety Evaluation 93-8 93.7.4 Inspection and Testing Requirements 93-8 93.7.5 Instrumentation Application 93-8a 93.8 Radioactive Drain Transrer System 9 3-11 93-11 933.1 Design Bases 93.8.1.1 Safety Design Bases 9 3-11 93.8.1.2 Power Generation Design Bases 9 3-11 93.8.2

System Description

9 3-11 93-iiia Amendment I

ABWR mamn Standard Plant un n SECTION 9.3 CONTENTS (Continued)

Section Title Page 93.8.2.1 General Description 9 3-11 93.8.2.2 System Operation 9 3-11.1 93.8.23 Component Description 9 3-11.1 93.8.2.4 Safety Evaluation 9 3-11.2 93.8.2.5 Tests and Inspection 9 3-11.2 93.9 Hydrocen Water Chemistrv Svstem 9 3-12 93.9.1 Design Bases 9 3-12 93.9.1.2 Safety Design Basis 93-12 93.9.13 Power Generation Design Basis 9 3-12 93.9.2

System Description

9 3-12 93.93 Safety Evaluation 93.12 93.9.4 Inspection and Testing Requirements 9 3-12 93.9.5 Instrumentation and Controls 9 3-13 93.10 Oxveen Inlection System 9 3-13 93.10.1 Design Bases 9 3-13 93.10.2

System Description

9 3-13 93.103 Safety Evaluation 9 3-13.1 9 3.10.4 Tests and Inspections 93 13.1 9 3.10.5 Instrumentation Application 9 3-13.1 93.11 ZIne Iniection System 9 3-13.1 93.11.1 Design Bases 9 3-13.1 9 3-13.1 9 3.11.2 Safety Evaluation 93-iiib Amendment

ABWR amm^n Standard Plant un n SECTION 93 CONTENTS (Continued)

Section Title Eagt 93.113 Test and laspections 9 3-13.1 9 3.11.4 Instrumentation 9 3-13.1 93.12 COL License Information 9 3-13.2 9 3.12.1 Non-radioactive Drains (Interface Requirements) 9 3-13.2 9 3.12.2 Storage Tank Discharge Valve Reliability 93-13.2 i

l i

93-iiic Amendment

ABWR

-n Standard Plant uvn 9.3.3 Equipment and Floor (3) Open drainage lines from areas that are re-Drainage System quired to maintain an air pressure f

differentialfbut drain to a rndig;tstj@

The system which collects and transfers all EQ are provided with a water seal.

radioactive liquid wastes is discussed in subsection 9.3.8. The non-radioactive drains are (4) All drainage lines into each sump shall be discussed in this subsection. The non-radio-turned down and terminated below the lowest active drains consist of equipment inside the fluid level to which the sump pump can draw.

standard plant buildings and COL interface requirements for that portion outside the 933.13 System Description buildings. The drains release effluent to the site specific discharge structure. The potable The non-radioactive drain system is designed and sanitary water systems (Subsection 9.2.4) to assure that waste liquids, valve and pump includes the non-radioactive drains.

leakoffs and component drains and vents are directed to the proper area for processing. The 933.1 Non-Radioactise Drains process portion of the systems consists of sump pumps, valves and instrumentation. Sumps are 933.1.1 Safet,* Design Bases provided as shown in the arrangement drawings in Section 1.2.

(1) There shall be no interconnection between any portion of the radioactive drain transfer All drainage systems are essentially passive system and any non-radioactive waste system.

systems down to the sumps or yard pipe connections. This is, flow is by gravity with (2) Effluent from non-radioactive systems shall no valves, pumps, and the like in the lines such

~

bediionitore) prior to discharge to assure that failure could cause a system not to drain.

F M '/' E, that there are no unacceptable discharges.

All exposed drainage piping is seismically analyzed to remain intact following an SSE, and (3) Non-radioactive drains piping shall be thus will drain the area as required. See non-nuclear safety class and quality group D Subsection 3.4.1 for further details.

and shall not have any effect on the operation of safety-related equipment.

Unacceptable flooding consequences are precluded by the capacity of the drain and the (4) The floor drain piping system shall be placement of safety-related equipment on raised arranged with a separate piping system for pads or grating. Also, check valves in sump each quadrant. The piping shall be arranged pump discharge lines prevent reverse flow from so that flooding or backflow in one quadrant other sumps that have piping to common cannot adversely affect the other quadrants.

collection tanks.

(5) Any valves that are relied upon to prevent The design of the drain system precludes backflow shall be inspectable and testable release to the environs of radioactive liquid, and withstand SSE.

drain systems areM, all non-radioactive As a backup, however or radiation prior 933.1.2 Power Generation I)esign Bases (1) The drains shall be designed to collect and remove effluent from their point of origin to 933.1.4 System Operation and Component the site discharge structure.

Description (2) The sump level switches shall serve as The drain system in similar in operation and leakage monitors for equipment or systems component descriptions as discussed in Subsec-served by each sump. Leakage detection is tions 9.3.8.2.2 and 9.3.8.2.3 excepting radia-also discussed in Subsection 5.2.5.

tion effects and the interfacing discharge process in lieu of discharge to radwaste.

9M Amendment

ABWR mmu Standard Plant prv n 933.1.5 Safety Evaluation nstrumentatioAind alarms aNe~ quired to

, monitor the non. radioactive effluent discharge The non-radioactive drains are not (to assure there is no unacceptable release,to safety related. The sumps may be instrumented the cnvironment.

and alarmed as required to assure there is no effect on safety-related equipment.

9.3.4 Chemical and Volume Control System (PWR) 933.2 Non-Radioacthe Drain Interface (Not applicable to a BWR)

The COL applicant shall provide the continuation of the drain system (Subsection 9.3.5 Standby Liquid Control System 9.3.3.1) from the standard plant buildings to the site discharge structure. A conceptual design 93.5.1 Design Bases continuation is discussed in this subsection.

93.5.1.1 Safety Design Bases 933.2.1 Safety Design Bases (Interface Requirement)

The standby liquid control system (SLCS) has a safety-related function and is designed as a The safety design bases is the same as listed Seismic Category I system. It shall meet the in Subsection 9.33.1.1.

following safety design bases:

93.3.2.2 Power Generation Design Bases (1) Backup capability for reactivity control (Interface Requirement) shall be provided, independent of normal re.

activity control provisions in the nuclear The power generation design bases is the same reactor, to be able to shut down the reactor as listed in Subsection 933.1.2.

if normal control ever becomes inoperative.

I 933.23 System Description (Conceptual)

(2) The backup system shall have the capacity for controlling the reactivity difference The non-radioactive drain system collects between the steady-state rated operating waste water from the following sources: plant condition of the reactor with voids and the buildings (reactor, turbine, radwaste, service cold shutdown condition, including shutdown and other buildings), precipitation and other margin, to assure complete shutdown from the surface runoff. A system composed of collection most reactive conditions at any time in core piping, curb and gutter inlets, manholes and life.

pumps is provided. Waste water is sent to dual settling basins where suspended solids are (3) The time required for actuation and settled and oil is collected on the surface.

effectiveness of the backup control shall be Means are provided to perform any required tests consistent with the nuclear reactivity rate or analyses required by the discharge permit.

of change predicted between rated operating Periodically, one of the basins is taken out of and cold shutdown conditions. A fast scram service and the suspended solids and oil are of the reactor or operational control of removed.

fast reactivity transients is not specified to be accomplished by this system.

933.2.4 Safety Evaluation (Interface Requirement)

(4) Means shall be provided by which the functional performance capability of the The safety evaluation is the same as backup control system components can be S u b s e c t io n 9.3.3.1.5.

verified periodically under conditions

_ n approaching actual use requirements.

933.2.5 Instrumentation (nd Alar]d)(Interface Demineralized water, rather than the actual Requirement) neutron absorber solution, can be injected into the reactor to test the-operation of Provisions for obtaining water samples from the non-radioactive drain system shall be provided.

A sampling vui and analysis program shall be provided to show that radicac-tive liquids are not being discharged from the non-radioactive drain system.

ABM usamii Standard Plant n,y n all components of the redundant control system.

(5) The neutron absorber shall be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage or imperfect mixing.

(6) The system shall be reliable to a degree consistent with its role as a special safety system; the possibility of unintentional or accidental shutdown of the reactor by this system shall be minimized.

93.5.2 System Description The SLCS (Figure 9.3-1) is automatically initiated or can be manually initiated through the keyboard switches in the main control room to pump a boron neutron absorber solution into the reactor if the operator determines the reactor cannot be shut down or kept shut down with the control rods. Once the operator decision for initiation of the SLCS is made, the design intent is to simplify the manual process by providing dual keylocked switches. This prevents inadvertent injection of neutron absorber by the SLCS. However, the insertion of the control rods is expected to assure prompt shutdown of the reactor should it be required.

The ke3 ocked control room switch is provided l

to assure positive action from the main control room should the need arise. Procedural controls are applied to the operation of the keylocked control room switch.

The SLCS is required only to shut down the reactor and keep the reactor from going critical again as it cools.

The SLCS is needed only in the improbable event that not enough control rods can be inserted in the reactor core to accomplish shutdown and cooldown in the normal manner.

The boron solution tank, the test water tank, the two positive displacement pumps, the two motor-driven injection valves, the two motor-9C Amendment

ABWR m,w t Standard Plant sn 9.3.8 Radioactive Drain Transfer System (1) The drain transfer system shall be designed to collect and remove waste liquids from their 93.d.1 Design Bases point of origin to the radwaste system for j

further processing.

The rajoactise drains are part of the radwaste 4

system in Subsection 11.2.

(2) The sump level switches shall serve as leakage monitors for equipment or systems sersed by 9.3.8.1.1 Safety Design Bases each sump. Leakage detection is also discussed in Subsection 5.2.5.

l (1) The drain transfer system drains equipment and Door areas where required for structuralloading (3) Open drainage lines from areas that are re-reasons and to protect systems required for a quired to maintain an air pressure differential.

safe shutdown.

but drain to a radioactise sump, are provided with a water seal.

(2) All potentially radioactive drains are pipe directly to the radwaste system and shall not affect safety-(4) All drainage lines into each sump shall be related equipment operation.

turned down and terminated below the lowest fluid level to which the sump pump can draw.

(3) Containment and drywell penetrations shall be designed and fabricated in accordance with the 93.8.2 System Description ASME Code, Sectio,111, Class 2. Secondary Containment penetrations shall be in accordance The system P&lD showing the sumps with their with the ASME Code, Section Ill, Class 3.

pumps, piping, instruments and controls are provided in Section 11.2.

l (4) Effluent from the radioactive drains shall be monitored prior to discharge to assure that there 9.3.8.2.1 General Description are no unacceptable discharges.

The drain transfer system is designed to assure (5) The radioactive drain transfer collection piping that waste liquids, valve and pump leakoffs and shall be provided with the following features:

component drains and vents and system are directed to the proper area for processing. The process (a) These piping systems shall be non-nuclear pc tion of the systems consists of sump pumps.

safety class and quality group D with thep stunp coolers (if necessary) tanks, valves and cxception of the containment penetrations 6;rd instrumentation. Sumps are provided as shown in piping within the drywell which shall be the arrangement drawings in Section 1.2.

feismic fategory I and quality)idroup B.

JrlerN/O The following ECCS loops are located in separate (b) The floor drain piping system shall be watertight areas:

arranged with a separate piping system for each quadrant. The piping shall be arranged (1) RHR A RCIC so that flooding or backflow in one quadrant cannot adversely affect the other quadrants.

(2) RHR B and HPCF B (c) There shall be no interconnection between (3) RHR C and HPCF C any portion of the radioactive drain transfer system and any non-radioactive waste Each area contains all of the power. operated valves and associated instrumentation outside the system.

containment for the respective ECCS loop. There-(d) Any valves that are relied upon to prevent fore, a pipe break or major leak in one area could backflow shall be inspectable and testable not Dood any adjoining area and, consequently.

and withstand SSE.

would not render the loops inoperable. The consequences of internal flooding are discussed 9.3.8.1.2 Power Generation Design Bases further in Subsection 3.4.1.

Trdit T M8 ! I

? NJ]riw l dceph wr ] v % b a ck flo % Nfck VBhu epywept y-nu ! wjU w/+ + slall M

Amenomen In tie Eccs be.W5 ntic * :/q E HJ p aM y cjr s y C.

ABWR m m.,,,

Standard Plant m.< n All drainage systems are essentially passive systems down to the sumps or yard pipe connections. This is, Ilow is by gravity with no vah es, pumps, and the like in the lines such that failure could cause a system not to drain. All exposed drainage piping is seismically analyzed to remain intact following an SSE, and thus will drain the area as required. See Subsection 3.4.1 for further details.

Unacceptable flootiing consequences are precluded by the capacity of the drain transfer system and the placement of safety-related equipment on raised pads or grating. Also, check valves in sump pump discharge lines prevent reverse flow from other sumps that have piping to the radwaste collection tank.

The design of the drain transfer system precludes release to the environs of radioactive liquid.

Potentially radioactive systems (equipment, floor, and detergent drains) are routed directly to the radwaste system, with no cross connections to uncontrolled (storm drain, sanitary and normal waste) systeras. As a backup, however, all nonradio-actise drain systems are monitored for radiation prior to release to the emirons.

93411 Amendmen

[.-

t@N ammu Standard Plant un n finished floor. Floor drains in areas of potential radioactivity are welded directly to the collection piping and are prosided with threaded.T handle plugs of the same material. The T-handle plugs are used to seal the floor drains during i

hydrostatic testing of the drainage systems, system startup and during all required leakage rate testing. All drainage piping. except carbon steel and suspended stainless steel piping, is hydrostatically tested during system startup. It is also installed, as required, to preserve the integrity of the drainage systems. Floor drains in areas not restricted because of potential radioac-thity are prosided with caulked or threaded con-nections.

(5) Cleanouts - In collection system piping from ar-I cas of potential radioactivity, cleanouts are provided, when practicable, at the base of each vertical riser where the change of direction in horizontal runs is 90", at offsets where the i

aggregate change is 135" or greater and at maximum intervals of 50 feet. Equipment hubs and floor drains are also used as cleanout points.

Cleanouts are welded directly to the piping and located with their access covers flush with the finished floor or wa!!.

93.8.2..I Safety Evaluation The Drain Transfer System is not safety-related.

Sumps designated as containing radioactive wastes are equipped with charcoal filters in the vents. In the event of a LOCA signal, all drywell sur2ps are automatically isolated to preclude the uncontrolled release of primary coolant outside the PCV.

93.8.2.5 Tests and Inspections Drywell and reactor building floor and equipment drain sumps are provided with the following instru-ments and controls:

(1) High and low level switches are provided on each sump pump to start and stop the sump pump automatically. A separate high-high level switch set at a higher level starts the second pump and simultaneously actuates an alarm in the main control room.

(2) Leak detection is effected by monitoring the fre-quency and duration of pump runs.

9 M L2 Amendmer.t l:

4 ABWR no,mt, Standard Plant ym. n 9.3.12 COL License Information 93.12.1 Non-radioacthe Drains (Interface Requirements)

The COL applicant shall provide the continuation of the non-radioactise drain system outside of the standard plant buildings to process and monitor to the site discharge structure in accordance with Subsection 933.2.

t 93.12.2 Storage Tank Discharge Valse Reliability The COL applicant will confirm that the SLCS storage tank discharge valves will have adequate reliability requirements and that the valves be incorporated into the Operational Reliability Assurance Program. (See Subsection 93.5.4) 9 3-33.2 Amendment

r General Dectric Company ABWR PaoPRirTARv isroRmTios

==

Standard Plant am m nev n Il.2 LIQUID WASTE because of interlocks on the two valves leading to the MANAGEMENT SYSTEM discharge line.

I1.2.1 Design Basis in addition to providing a means for a controlled (i.e., batch) discharge, the sample tanks also function 11.2.1.1 Design objecthe as surge tanks to minimize or delay the offsite i

discharge of liquid volume for which there is no The liquid radwaste equipment is designed to immediate room available in condensate storage. By segregate, collect, store, and process potentially administrative control, the discharge from this single radioactive liquids generated during various modes discharge line to the canalis adjusted so that it can be of typical plant operation: startup, normal operation, shown that the discharge will meet the requirements I

hot standby, shutdown, and refueling. The system is of 10CFR20 on concentration limits and Appendix !

l designed such that it may be operated to maximize of 10CFR50 recycle of water within the plant, v.hich would minimize the releases of liquid to the environment.

Per General Design Criteria 64, means are Maximizing recycle serves to minimize the potential provided for monitoring effluent discharge paths that for exposure of persons in unrestricted areas from may be released from normal operations, including the liquid release pathway. The radioactive drain anticipated operational occurrences and from transfer portion of the liquid radwaste system is postulated accidents. The monitoring of liquid described in Subsection 9.3.8.

release as required by General Design Criterion 64 is accomplished in two steps. First, the sources of 11.2.1.2 Design Criteria release are only from either the HSD sample tank or the sample tanks. These tanks have the necessary The criteria considered in the design of this connections to the sampling system to allow analysis system include minimization of solid waste shipped prior to discharge, for burial, reduction in personnel exposure, minimization of offsite releases and maximizing the The system is designed to treat process liquids quality of water returned to the primary system.

with radionuclide concentrations associated with the design-basis fuelleakage and produce water suitable Per General Design Criteria 60 of 10 CFR 50, for recycle to condensate storage. Plant water Appendix A, the radwaste system design includes balance considerations may require the discharge of means to suitably control the release of radioactive processed liquids to the environs, in which case materials in gaseous and liquid effluents and to concentrations of radionuclides in the effluent will handle radioactise solid wastes produced during meet the requirements of 10CFR20. Radiation normal reactor operation, including anticipated exposure to persons in unrestricted areas resulting operational occurrences. These operational oc-cut-from liquid waste discharged during normal operation rences include condenser leakage, maintenance and anticipated operational occurrences will be less activities, and process equipment down time. Tbc than the values specified in 10CFR50, Appendix 1.

liquid radwaste system provides one discharge line to Liquid discharge to the canal may be initiated from the canal for the release of liquid waste, with the only one sample tank at a time through a locked-flow rate of this effluent stream controlled by a flow dosed valve that is under administrative control. The control valve together with the necessary flow discharge sequence is initiated manually. No single instrumentation. Radiation monitoring equipment is error or failure will result in discharge. The design placed on this line to measure the activity discharged will maintain occupational exposure as low as and to assure specified limits are not exceeded. A practicable in accordance with Nuclear Regulatory high radiation signal from this monitor will close the Commission Regulatory Guide 8.8 while operating discharge valve. The sing!c discharge line is fed by with the design-basis fuelleakage.

the hot shower drain (HSD) receiver tank, a very low level radioactivity source, or one of the two sample The design pressure and temperature of the tanks which usually contain condensate quality water.

pressure-retaining process,cquipment in the liquid These two sources may not discharge simultaneously radwaste system is 11 kg/cm' and 100"C, respectively.

The concentrators, however, are designed for 178 C.

l 112 1 Amendmeru

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vendor should modify the core spray and LPCI system logic so that these fm systems will restart, if required, to ensure adequate core cooling. Before (j

modifying the design, the staff indicated also that the vendor should submit a preliminary design to the staff for approval because it affects several core-cooling modes under accident conditions.

The GE ABWR endorsed the conclusions of the study performed by the BWR Owners' Group which was forwarded to the staff in a letter of December 29,19B0, from D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC). The BWR Owners' Group concluded that the current BWR ECCS design is adequate and that the proposed changes would decrease the overall safety of the plant.

For example, the Owners' Group stated that the modification would significantly escalate the control system complexity, restrict the operator's flexibility when dealing with anticipated events, and reduce system reliability. The Owners' Group concluded that the current ECCS design is adequate because the BWR operator training is comprehensive and thoroughly addresses reactor water level control, the emergency operating procedures address this issue, the operator has sufficient time to correct errors, and the low reactor water level conditions are clearly displayed and alarmed in the control room.

The staff reviewed the results of the Owners' Group study and considered the emphasis placed on water level, control in BWR operator training. The staff finds GE's response acceptable and agrees that the ABWR design need not be modified to provide an automatic restart of the low-pressure ECCS. Therefore, the ABWR meets the requirements of this TMI Item. See the b'elow discussion for 50.34(f)(2)(xviii) for additional water level measurement concerns.

6.

10 CFR 50.34(f)(1)(ix): NUREG-0737 Item II.K.3.24, Confirm Adequacy of Space Cooling for HPCS and RCIC Systems The TMI Issue 11.3.24 of NUREG-0737 has been,1r$orporated into the construc-tion permit (CP)/ manufacturing license (ML) Rule as 10 CFR 50.34(f) Para-graph 1.ix and it addresses the adequacy of space cooling for the high pressure coolant injection systems at a BWR.

Long-term operation of these systems during a complete loss of offsite power, for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, may require space cooling to maintain the high pressure coolant injection pump rooms within allowable temperature limits. The high pressure coolant injec-tion systems and their respective support systems should be designed to withstand the consequences of a complete loss of offsite ac power for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The ABWR design contains two systems to previde room cooling to the HPCF and RCIC systems. During normal operation, the two HPCF pump rooms and the single RCIC pump room are cooled by the non-safety-related secondary containment heating, ventilation, and air conditioning (HVAC) system. During accident conditions, the room cooling function is transferred to the safety grade essential equipment HVAC system and the secondary containment HVAC system is isolated. The essential equipment HVAC system consists of a fan coil unit in each of the pump rooms. Cooling for each fan coil is provided by the safety-related portion of the applicable train of the reactor building cooling water (RBCW) system. The fan coil unit in the applicable pump room, the HPCF sub-system or the RCIC system which the fan coil unit is serving and the asso-ciated RBCW train are all on the essential electrical power divisions Jefv &

ABWR DFSER 20-15 October 1992

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V g y' r

which include $ t,he divisional onsite ac power sourcek that 1s", emergency onsite diesel generator! The fan coil unit is necessary to keep the tempera-O ture of the associated pump room within its design limits.

It is automati-i cally initiated upon start-up of the respective HPCF or RCIC pump.

KC & TD W 6 %' BUB ?3 M< D ~+" W % od'f*" L

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Based on the above, the staff concludes that space cooling for the HPCF and the RCIC systems will be available as required following a complete loss of offsite ac power to the plant for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. However, the design characteristics for the fan coil units were not specifically identified in ABWR SSAR Section 9.4.5.

The ability of the fan coil units to remove suffi-cient heat to maintain the pump room temperatures within design limits was not confirmed. Also, SSAR Tables 8.3-1 and 8.3-2 which list diesel generator loads, do not include the fans of the pump rooms fan coil units. This is Open Item 20.3-1.

A e +er 11 Tde f.4 -+

The HPCF and RCIC pump room cooling systems will be in conformance with the requirements specified in TMI Issue II.K.3.24 and 10 CFR 50.34(f) Para-graph 1(ix) subject to GE's demonstration that the room coolers are adequately sized to maintain the rooms within design environmental conditions following a complete loss of offsite ac power to the plant for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and inclu-sion of the applicable fans of the fan coil units for the HPCF and RCIC pump rooms in the SSAR Tables 8.3-1 and 8.3-2.

7.

10 CFR 50.34(f)(1)(x): NUREG-0737 Item II.K.3.28, Verify Performance Capability of ADS, Valves, Accumulators, and Associated ' Equipment and Instrumentation during and following an Accident Situation O-Paragraph (1)(x) of 10 CFR 50.34(f) requires the following:

" Perform a study to ensure that the Automatic Depressurization System, valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended functions during and following an accident situation, taking no credit for non-safety-related equipment or instrumentation, and accounting for normal expected air (or nitrogen) leakage through valves" for BWRs. The above regulation incorporates the requirements of TMI Action II.K.3(28) of NUREG-0737.

The ABWR has 18 quality Group A, seismic Category I SRVs of which 8 can additionally perform the ADS function. One 189-liter (50-gallon) capacity nitrogen accumulator with a design pressure of 1379 kPa (200 psig) is provided for each ADS SRV to support its ADS function during and following an accident situation. The accumulator supplies compressed nitrogen gas to the valve for its actuations. The staff's safety evaluation on GE's compliance for ABWR with the above regulation is essentially based en ABWR SSAR Sections IA.2.31 and 6.7 and is limited to the adequacy of the accumulators and associated equipment to perform their intended functions during and following an accident situation. The staff's safety evaluation on GE's compliance for ABWR with other requirements of the regulation such as the adequacy of the ADS SRV valves and its associated instrumentation and controls to perform their intended functions during and following an accident situation are provided in ABWR FSER Section 5.2.2, " Overpressure Protection," 6.1, " Emergency Core Cooling Systems," and 7.3, " Engineered Safety Features Systems."

O l

ABWR DFSER 20-16 October 1992

Table 8.3-4 k

D/G LOAD SEQUENCE DIAGRAM it MAJOR LOADS

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ABM n461 ii Standard Plant RIr n TABLE 9.2-4a REACTOR BUILDING COOLING WATER DIVISION A Emergency Normal (LOCA) (Sup.

Operating Mode /

Operating Shutdown at 4 Shutdown at 20 Ilot Standtsy llot Standby pression Pool Compcments Conditions hours hours (no loss of AC)

(km of AC) at97 C ESSINDAL Ileat now fleat now liest now fleat now Ileat now Ileat now (Note 1)

Emergency Dec.

123 1.010 123 1,010 sel Generator A RIIR IIca:

102.4 5.280 32.8 5.280 24.0 5.280 84.7 5.280 Exchanger A FPC licat 66 1.230 6.6 1.230 6.6 1.230 6.6 1.230 6.6 1.230 9.1 1.230 fachanger A Others (essen-3.1 640 34 (A0 3.6 640 3.2 640 3.9 640 4.0 640 tal)(Note 2)

NON-ESSINDAL RWCU lleat 19.1 700 700 700 19.1 700 19.7 700 Exchanger inside Drywell 5.7 1.410 5.7 1,410 5.7 1.410 5.7 1,410 3.2 1.410 (Note 3)

Others (non-23 440 23 440 23 440 23 440 0.8 260 0.7 2(o essential)

(Note 4)

Totalload 37.0 4.420 120.6 9,700

$1.2 9 *KO 37.1 4,420 70.7 10330 111.0 8.420 NOTES:

6 (1) Heatx 10 Bru/h;flowx g/m, sums may not be equaldue to rounding.

l (2) HECWrefrigerator, room coolers (FPCpump, RHR, y SGTS, CAAfS), RHR motor and sealcoolers.

(3) Drywell(A & C) and RIP coolers.

(4) instruments and senice air coolers:R WCUpump cooler, CRD pump oil, and RIP Afg sets.

O 9 2-17 Amendment 26 I

/

23A6100All Standard Plant REV B TABLE 9.2-4b w

REAGOR BUILDING COOLING WATER DIVISION B t

Emergency Nonna!

(LOCA) (Sup.

Operatmg Mode /

Operatmg Shutdown at 4 Shutdown at 20 110 Standby llot Standby prenton Pool Components Conditions hours hours (no Icu of AC)

(lou of AC) at 97 C ESSENDAL lleat Dow Ileat Dow fleat Dow fleat Dow IIcat Dow licat Dow (Note 1)

Emergency Die-123 1.010 12.5 1.013 sel Generator B RI1R IIcat 102.4 5280 32.8 5280 24.0 5.280 84.7 5280 Exchanger B ITC ileat 6,6 1.230 6.6 1230 66 1.230 66 1.230 66 1.230 9.1 1.2%

t Exchanger B Others (essen-4.9 1860 5.4 1860 5.4 1860 4.9 1860 5.5 1860 63

-1560 taal)(Note 2)

NON-ESSENTIAL RWCU Ileat 19.1 700 700 700 19.1 700 19.1 700 Exchanger Inside Drywell 5.1 1.230 5.8 1.230 5.1 L230 5.1 1.230 23 1.230 (Note 3)

Others (non-2.6 700 14 700 1.4 700 L4 700 03 40 40 essential)

(Note 4)

Total lmad 383 5."'20 121.6 11.000 513 11.000 37.1 5.720 70 3 11,350 112.6 9.530 NOTES:

6 (1) Heat x 10 Bru/h;flowxg/m, sums may not be equaldue to rounding.

(2) HECWrefrigerator,roomcoolers(FPCpump,RHR,HPCF,SGTS,FCS, CAMS),HPCFandRHRmotor andmechanicalsealcoolers.

(3) Drywell(B)andRIPcoolers.

(4) Reactor Building sampling coolers; LCW sump coolers (in drysell and reactor building), RIP MG sets and RWCU pump coolers.

(

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Amendment 24 9.2-18

23A6100Ali Standard Plant nrv a TABLE 9.2-4e g

t\\

REACTOR BUILDING COOLING WATER DIVISION C Emergency Normal (IDCA) (Sop-Operating Mode /

Operating Shuidown at 4 Shutdown at 20 110t Standby llot Standby pression Pool Components Conditions hours hours (no loss of AC)

(loss of AC). at 97 C ESEvnAL IIcat now IIcat now IIcat now lient now

!!ces now fleat now (Note 1)

Emergency Die-123 1,010 12,5 1,010 sel Generator B R}IR liest 102.4 5,280 32.8 5,:30 24.0 5,280 84.7 5,280 Exchanger B Others (essen-5.8 2,780 6.3 2.780 6.3 2,780 5.8 2,780 58 2,780 6.9 2,780 taal)(Note 2)

NON-ESSENTIAL Others (non-19 4 1,560 18.2 1,860 7.0 1,860 19.4 1,860 03 220 0.7 220 essential)

(Note 4)

Totalload 25.2 4,640 126.9 9.920 46.1 9,920 25.2 4.640 42.8 9.290 104.8 9.290 NOTES:

(1) Heat x 10 Bru/h;flowx g/m, sums may not be equal due to rounding.

(2) HECW refrigerator, room coolers motor coolers, and mechanical seal coolers for RHR and HKF.

(3) Instrument and service air coolers, CRD pump oil cooler, radwaste components, HSCR condenser, and turbine building sampling coolers.

(s Amendment 21 9.2-19

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ABWR ux61oosu i

Standard Plant ws

}

Table 9.4-4 O

HVAC SYSTEM COMPONENT DESCRIPTIONS (Continued)

(Response to Question 430.243) i ESSENTIAL EQUIPMENT LIST Essential AirConditioners Capacity j

i (Btu /hr)

I HPCp Pump Room AC Div B 436,500 C

HPCS Pump Room AC Div C 436,500 e

RHR Pamp Room AC Div A 291,700 I

RHR Pump Room AC Div B 291,700 RHR Pump Room AC Div C 291,700 t

RCIC Pump Room AC Div A 65,500 4.

j FCS Room AC Div B 52,000 t

FCS Room AC Div C 52,000

. i FPC Pump Room AC Div A 27,000 FPC Pump Roon AC Div B 27,000

-t C AMS Room AC Div A 79,400 CAMS Room AC Div B 79,400 SGTS Room AC Div A 16,000 SGTS Room AC Div B 16,000

}

l NON - ESSENTIAL EQUIPMENT LIST l

l Heating / Cooling Coils Quantity Cooling Heating i

(Bru/hr)

(Btu /hr)

..j RB Secondary Containment HVAC 4 (1 on standby) 6,100,000 9,100,000

. RIP Panel Room HVAC Dhhion A 1

2,000,000 2,700,000 i

4 RIP Panel Room HVAC Division B 1

2,000,000 2,700,000 k

i Fans Quantity Capacity t

(cmb) f RB Secondary Containment Intake Fans 4 (1 on standby) 57,500 RB Secondary Containment Exhaust Fans 4 (1 on standby) 57,500 i

Purge Air Exhaust Fan 1

22,000 RIP Panel Room Dhhion A Fans 2 (1 on standby) 50,000 RIP Panel Room Division B Fans 2 (1 on standby).50,000 i

l J

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9.4-7.4 f

Amendment 22 i

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23A6103AB Standard Plant

%c I

refueling floor ventilation exhaust, both SGTS trains efficiencies are outlined in Table 6.5-1. Dose are automatically operated. When the operation of analyses of events requiring SGTS operation, both the trains is assured, one train is placed in described in Subsections 15.6.5 and 15.7.4 standby mode. In the event a malfunction disables indicate that offsite doses are within the limits established by 10 CFR 100.

an operating train, the standby train is automatically b s

,g Jd imtiated.

g (3) The SGTS is designated as an engineered 1

Igbe d

safety feature since it mitigates the 651.23.2 Manual consequences of a postulated accident by The S

's on standby during no #ptint controlling and reducing the release of operation and m. (e manuall isted before or radioactivity to the environment. The SGTS, during primary contai.

purging (de-inerting) except for the deluge, is designed and built to wben required to lpi etfe de e of contaminants the requirements for Safety Class 3 equipment to the enviro 3dnt within 10CFR.

'mits. It may as defined in Section 3.2, and 10 CFR 50, be manualrfinitiated for testing or when r its use Appendix B.

may needed to avoid exceeding radiation manif or points.

The SGTS has independent, redundant active trains. Should any active train fail, SGTS 6.5.1.2.33 Decay Heat Removal functions can be performed by the redundant train. The electrical devices of independent Cooling of the SGTS filters may be required to components are powered from separate Class prevent the gradual accumulation of decay heat in IE electrical busses.

the charcoal. This heat is generated by the decay of radioactive iodine adsorbed on the SGTS charcoal.

(4) The SGTS is designed to Seismic Category I The charcoal is typically cooled by the air from the requirements as specified in Section 3.2. The process fan.

SGTS is housed in a Category I structure. All surrounding equipment, components, and A water deluge capability is also provided, but supports are designed to appropriate safety primarily for fire protection since redundant process class and seismic requirements.

fans are provided for air cooling. Since the deluge is available, it may also be used to remove decay heat (5) A secondary containment draw-down analysis for sequences outside the normal design basis.

will be performed to demonstrate the capability Temperature instrumentation is provided for control of the SGTS to maintain the design negative of the SGTS process and space electric heaters. This pressure following a LOCA including inleakage instrurcentation may also be used by the operator to from the open, non-isolated penetration lines

[re.] establish a cooling air flow post. accident, if identified during construction engineering and required.

the event of the worst single failure of a l secondary isolation valve to close. (See Water is supplied from the fire protection system Subsection 6.5.5.1 for COL license information and is connected to the SGTS via a spool piece.

requirements).

6.5.13 Design Evaluation 6.5.13.2 Sizing Basis 6.5.13.1 General Figure 6.5-2 provides an assessment of the secondary containment pressure after the (1) A slight negative pressure is notmally design-bag LOCA assuming an SGTS fan capacity maintained in the secondary containment by of 6800 m hr (21 C,1 atmosphere) per fan. Credit the reactor building HVAC system (Subsection for secondary containment as a fission product 9.4.5). On SGTS initiation per Subsection control system is only taken if the secondary 6.5.1.23.1, the secondary containment HVAC containment is actually at a negative pressure by is automatically isolated.

considering the potential effect of wind on the ambient pressure in the vicinity of the reactor (2) The SGTS filter particulate and charcoal building. For the ABWR dose analysis, direct transport of containment leakage to the environment 6.%2 Amendment 26

SVf L c} (

ABWR m,.

Standard Plant nrv c 6.5 FISSION PRODUCTS RE510 VAL AND (4) Remain intact and functional in the esent of CONTROL SYSTEN1S a safe shutdown earthquake (SSE).

6.5.1 Engineered Safety Features Filter (5) M e e t e nvir on m e n t al q ualification Systerns requirements established for system operation.

The filter systems required to perform safety.related functions following a design basis (6) Filter airborne radioactivity (halogens and accident are:

particulates) in the effluent to reduce offsite doses during normal and upset (1) Standby gas treatment system (T22.SGTS).

operations to within the limits of 10CFR20.

(2) Control room portion ol' the HVAC system.

6.5.1.2 System Design (U41.HVAC) 6.5.1.2.1. General i

The control room portion of the HVAC system is 3

discussed in Section 6.4 and Subsection 9.4.1.

The SGTS P&lD is provided as Figure 6.5-1.

rNN/

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The SGTS is discussed in this Subsection (6.5.1).

d 9

6.5.1.2.2 Component Description M

C-- O,

6.5.1.1 Design Basis Table 6.51 provides a summary of the major 6.5.1.1.1 Power Generation Design Basis SGTS components. The SGTS consists of two j parallel and redundant filter trains %E The SGTS has the capability to filter the is taken from above the refueling area or from gaseous effluent from the primary containment or the primary containment via the atmospheric from the secondary containment when required to control system (T31-ACS). The treated discharge l limit the discharge of radioactivity to the goes to the main plant stack.

environment to meet 10CFR100 requirements.

The SGTS consists of the following principal 6.5.1.1.2 Safety Design Basis components:

The SGTS is designed to accomplish the (1) Two filter trains each consisting of a following:

moisture seprator, an electric process heater, a prefilter, a high efficiency (1) Maintain a negative pressure in the particulate air (HEPA) filter, a charcoal secondary containment, relative to the adsorber, a second HEPA filter, and space outdoor atmosphere, to control the release heaters.

of fission products to the environment.

(2) Two independent process fans located downstream of the each filter train and two independent cooling fans for the removal of decay heat from charcoal.

(2) Filter airborne radioactivity (halogen and air particulates) in the effluent to reduce 6.5.1.23 SGTS Operation offsite doses to within the limits specified in 10CFR100.

6.5.1.23.1 Automatic Upon the receipt of a high primary containment pressure signal or a low reactor (3) Ensure that failure of any active component, water level signal, or when high radioactisity assuming loss of offsite power, cannot is detected in the secondary containment or impair the ability of the system to perform its safety function.

OI Amendment 24

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4

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6.S.1.2.3.2 Manual N

The SGTS is in standby mode during j

normal plant operation and may be manually s

1 initiated before or during the primary containment inerting, de-inerting, pressure (Tl d

control or containment purge when required to limit the discharge of contaminants to the s

environment within 10CFR20 limits, it may be I

manually initiated for testing or whenever its use may be needed to avoid exceeding radiation monitor setpoints.

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Definitions 1.1 Table 1.1-1 (page 1 of 1) i MODES l

REACTOR MODE AVERAGE REACTOR I

Mord TITLE SWITCH POSITION COOLANT TEMPERATURE i

t (*F)

-l 1

Power Operation Run NA 2

Startup Refuel (a) or Startup/ Hot NA Standby 3

HotShutdown(a)

Shutdown

> 200 T3 d i

4 ColdShutdown(a)

Shutdown s 200 73 d S

Refueling (b)

Shutdown or Refuel NA I

(a) All reactor vessel head closure bolts fully tensioned.

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(b) One or more reactor vessel head closure bolts less than fully tensioned.

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5W 6 o{ g SGT syste 3.5.4.3

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4 3.6 CONTAINMENT SYSTEMS 1

3.5.4.3 Standby Gas Treatment (SGT) System reu;ns MTwo[SGTOdy:t::: shall be OPERABLE.

LCO 3.6.4.3 t

i h

APPLICABILITY:

MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the M secondary % ntainment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

{

CONDITI$N REQUIRED ACTION COMPLETION TIME re~nin frate l

A.

One SGT :d:yst=

A.1 Restore SGT : d:y: tem 7 days inoperable.

to OPERABLE status.

l B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion l

Time of Condition A A_N,Q i

not met in MODE 1, 2, or 3.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.

Required Action and

---

NOTE-------------

associated Completion LCO 3.0.3 is not applicable.

l Time of Condition A not met during movement of irradiated C.1 Place OPERABLE SGT Immediately f " i"- ;pe:3;t;; in d

fuel assemblies in the o ration.

r{secondaryM 4

1 containment, during CORE ALTERATIONS, or 98

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during OPDRVs.

(continued)

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3.5.4.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.

(continued)

C.2.1 Suspend movement of Immediately irradiated fuel assemblies in

,3-[seconda ry}1r-containment.

152 C.2.2 Suspend CORE Inmediately ALTERATIONS.

AME i

C.2.3 Initiate action to Immediately suspend OPDRVs.

i t r a ina D.

Two SGT :. :y;tems D.1

---

NOTE---------

inoperable during LCO 3.0.3 is not movement of irradiated applicable.

fuel assemblies in the ev{ secondary}'~~

containment, during Suspend movement of Immediately CORE ALTERATIONS, or irradiated fuel during OPORVs.

assemblies in

,s{ secondary}v-

~

containment.

ANE D.2 Suspend CORE Immediately i

ALTERATIONS.

bnB D.3 Initiate action to Immediately suspend OPDRVs.

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SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

i'

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SR 3.6.4.3.1 Operate each SGT ::b:y:::: for n#{10T'~

31 days l

continuoushours,1withheaters operating f 4

I SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP TestingProgram(VFTP).

l 1 rsin SR 3.6.4.3.3 Verify each SGT ::t:y::;; actuates on an r{1gf' months actual or simulated initiation signal.

r{181"Eonths

-v--

l b[3.6.4.3.4 Verify each SGT filter cooler bypass damper can be opened and the fan started.

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GE Nuclear Energy Docket No. STN 52-001 April 2.1993 1

Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - DFSER Confirmatory Item 6.2.5-1 and 6.2.5-2 and Open Item 203-10

Dear Chet:

Enclosed are SSAR markups addressing DFSER Confirmatory Items 6.2.5-1 and 6.2.5-2 and Open Item 20.310.

Please provide a copy of this transmittal to Gene Gou.

Sincerely, s

i.

$4 Jack Fox Advanced Reactor Programs cc: Bill Fitzsimmons (GE)

Norman Fletcher (DOE)

Bernie Genetti(GE) i Jf B 76 I

l' q jof 5 cnm L.S.s'. s gy met:ma nrv c Standard Plant (16) The primary containment purge system will required, through a pathway from the werwell aid in the long-term post-accident cleanup airspace to the stack. The pathway is isolated operation. The primary containment during normal operation with two rupture disks.

atmosphere will be purged through the SGTS to the outside environment. Nitrogen makeup The following modes of operation are provided:

will be available during the purging operation.

(1) Startup - Inerting. Liquid nitrogen is vaporized with steam or electric heaters to (17) The system is also designed to release a temperature greater than 200F and is containment pressure before uncontrolled injected into the wetwell and the drywell.

containment failure could occur.

The nitrogen will be mixed with the primary containment atmosphere by the drywell coolers in the drywell and, if necessary, 6.2.5.2 System Design by the sprays in the wetwell.

6.2.S.2.1 General MMA T _ 6_. ?. I 2./

A (2) Normal - Maintenance of Inert Condition. A The ACS provides control over hydrogen ancD nitrogen makeup system automatically sup-

{ex"veen generated following a LOCA[In an inerted plies nitrogen to the wetwell and upper contamment, mixing of any hydrogen generated is drywell to maintain a slightly positive not required. Any oxygen evolution from pressure in the drywell and wetwell to pre-radiolysis is very slow such that natural clude air leakage from the secondary to the convection and molecular diffusion is sufficient primary containment. An increase in con-i to provide mixing. Spray operation will provide tainment pressure is controlled by venting l further assurance that the drywell or wetwellis through the drywell bleed line.

uniformly mixed. The ste consists of the following features:

e (3) Shutdown - Deinerting. Air is provided to the drywell and wetwell by the primary n/IERT 6 7,5,2-2 (1) Atmospheric mixing is achieved by natural containment HVAC purge supply fan. Exhaust processes. Mixing will be enhanced by is through the drywell exhaust lines and operation of the containment sprays, which wetwell to the plant vent, through the HVAC gg are used to control pressure in the primary or SGTS, as required.

[#*U containment.

/MFA T 4 2. 5. 2-3 g,(4) Overpressure Protection. If the wetwell (2) rrhe pumary containment nitrogen purgg pressure increases to about 5.6 establishes and maintains an oxygen -

kg/cm g, the rupture disks will open.

2 deficient atmosphere (.s3.5 volume percent) f The overall containment pressure decreases in the primary containment during normag as venting continues. Later, the operators I

can close the two 350A air-operated operatioV butterfly valves to re-establish (3) The redundant oxygen analyzer system (CAMS) containment isolation as required.

measures oxygen in the drywell and suppression chamber. Oxygen concentration The following interfaces with other systems are displayed in the main control room.

are provided:

Description of safety related display Residual Heat Removal System (RHR-Ell).

instrumentation for containment monitoring (1) is provided in Chapter 7.

Electrical The RHR provides post-accident suppression requirements for equipment associated with pool cooling as necessary following heat the combustible gas control system are in dumps to the pool, including the exothermic accordance with the appropriate IEEE heat of reaction released by the design basis metal water reaction. This heat of j

standards as referenced in Chapter 7.

reaction is very small and has no real affect on pool temperature or RHR heat in addition, the ACS provides overpressure exchanger sizing. The werwell spray protection to relieve containment pressure, as 6233 Ammendment 17

i CNFM b. 2 5-l

()) 2 o 0 o gs E n r G 2. 5 2'~ l The FCS and ACS are systems designed to control the environment within the primary containment.

The FCS provides control over I

hydrogen and oxygen generated following a LOCA.

INS E R T 4. 2. 5. 2 -2 (1)

The FCS has two recombiners installed in the secondary containment.

The recombiners process the combustible gases drawn from the primary containment drywell.

(2)

The FCS is activated when a LOCA occurs.

The oxygen and hydrogen remaining in the recombiners after having been processed are transmitted to the suppression pool.

The ACS provides for inerting the primary containment during normal or on-line operation.

This system is not designed as an on-line containment purging system.

The ACS exhaust line isolation valves are closed when ABWR is on-line.

The nitrogen supply lines, compensating for leakage, provide a continuous flow of nitrogen to the containment.

If a LOCA signal is received by the ACS the nitrogen supply valves close.

Nitrogen purge from the containment occurs at SAvfdown.

"i'- ;rn gurging is accomplished as the containment exhaust isolation valves are opened and the nitrogen supply valves are closed.

Nitrogen is replaced by air in the containment (See Item (3) Shutdown - Deinerting below in this Subsection).

The system has the following features:

IMEERr 6.1.F.1-3

-(2) The ACS primary containment nitrogen makeup maintains an oxygen'-deficient atmosphere (s3.5 volume percent) in the primary containment during normal operation.

ad +

ABM C^'N M, f-I F7 M 3 maan Rev D Standard Plant 9.4.5 Reactor Building Ventilation System 9.4.5.1.2 System Description dn The reactor building HVAC system is composed The reactor building secondary containtnent V

of the following subsystems:

HVAC system P&lD is shown Figure 9.4-3. The system flow rates are given in Table 9.4-3, and the (1) Secondary Containment HVAC System system component descriptions are given in Figure 9.4 3. The HVAC system is a once-through type.

(2) Essential Equipment HVAC System (14)

Outdoor air if filtered, tempered and delivered to the secondary containment. The supply air system (3) Non. Essential Equipment HVAC System (8) consists of a medium grade filter, a beating coil, a cooling coil, and three 50% supply fans located in the (4) Essential Electrical Equipment HVAC System turbine building. Two are normally operating and the other is on standby. The supply fan furnished (3) conditioned air through ductwork and registers to (5) Essential Diesel Generator HVAC System (3) the equipment rooms and passages. The exhaust air

<*onferime,.T c/cmer//~dy system pulls the air from the rooms through ductwork, filters and monitors the air for (6)4Lrywed PurgpSupply/ Exhaust System radioactivity and exhausts out the plant stack. fyJEA7 (7) Mainsteam/Feedwater Tunnel HVAC System q,y,ff,g (8) Reactor Internal Pump Control Panet Room Operation of the secondary containment HVAC 9.4.5.1 Secondary Containment HVAC System system is not a prerequisite to assurance of either of the following-9.4.5.1.1 Design Bases (1) integrity of the reactor coolant pressure 9.4.5.1.1.1 Safety Design Bases boundary, or The secondary containment HVAC system has no (2) capability to safely shut down the reactor and to maintain a safe shutdown condition.

safety.related function as defined in Section 3.2.

Failure of the system does not compromise any safety.related equipment or component and does not However, the system does incorporate features prevent safe reactor shutdown. Provisions are incor.

that prodde reliability over the full range of normal porated to minimize release of radioactive plant operation. The following signals automatically substances to atmosphere and to prevent operator isolate the secondary containment HVAC system:

exposure.

(1) secondary containment high radiation signal, 9.4.5.1.1.2 Power Generation Design Bases (2) refueling Door high radiation signal, The secondary containment HVAC system is de-signed to provide an environment with controlled (3) drywell pressure high signal, temperature and airuow patterns to insure both the comfort and safety of plant personnel and the integ-(4) reactor water levellow signal, and rity of equipment and components.

(5) secondary containment HVAC supply / exhaust The secondary containment is maintained at a neg-fans stop.

ative pressure with respect to atmosphere.

On a smoke alarm in a division of the secondary containment HVAC system, the HVAC system shall The system design is based on outdoor summer conditions of 115 F, outdoor winter conditions of be put into smoke removal mode. To remove smoke

-40 F. Space temperature is maintained as speciSed from the secondary containment, the standby exhaust and supply fans are started to provide an increase in in Appendix 3I.

air flow through the secondary containment. The wsea r 9M, 5~. I. I.2 9 4.ge and exloud used by -Me AC.S Amendment 22 HVAC. N sgpfaly b nmen+ de'.ney b is

+or pr. n y con t wS d,scussed in S A ser.O on (a 2. 5,2. )M cRction k. d 1 @ -

the. shutdown mode, o4 cpotica in

C4/fM S 2. f-L Py I of f, '

QQ S// # #AM#

Standard Plant elevation which would be covered by post-LOCA the same time and made from the same sheet flooding for unloading the fuel.

to provide uniformity of relief pressure.

6.2.5.2.5 Pressure Control (6) The rupture disks are capable of withstanding full vacuum in the wetwell (1) In general, during startup, normal, and vapor space without leakage.

abnormal operation, the wetwell and drywell pressures is maintained greater than 0 psig (7) The piping material is carbon steel. The s

to prevent leakage of air (oxygen) into the design pressure is 10.5 kg/cm'g (150 primary containment from secondary psi),, and the design temperature is containment but less than the nominal 2 psig 171 C.

scram set point. Sufficient margin is provided such that normal containment 6.2.5.2.7 Recombiner"W I 0'U* A l.

temperature and pressure fluctuations do not Two permanently installed safety-related li cause either of the two limits to be reached (1) considering variations in initial recombiners are located in secondary [

containment conditions, instrumentation containment. Each recombiner, as shown in i errors, operator and equipment response Figure 6.2-40, takes suction from the i time, and equipment performance.

drywell, passes the process flow through a g heating section, a reactor chamber, and a (2) Nitrogen rgakeup automatically maintains a spray cooler. The gas is returned to the wetwell.

530 kg/m' (0.75 psig) positive pressure to avoid leakage of air from the secondary into the primary containm:nt.

(2) The recombiners are normally initiated on high levels as determined by CAMS (if (3) The drywell bleed sizing is capable of hydrogen is not present, oxygen concentrations are controlled by nitrogen,

maintaining the primary, containment pressure j

less than 880 kg/m' (1.25 psig) during makeup).

the maximum containment atmospher c heating which could occur during plant startup.

6.2.5.3 Design Evaluation The ACS is designed to maintain the 6.2.5.2.6 Overpressure Protection containment in an inert condition except for (1) The system is designed to passively relieve nitrogen makeup needed to maintain a positive containtnent pressure and prevent air (0 )

the weJwell vapor space pressure at 5.6 leakage from the secondary into the primary 2

kg/cm g. The system valves are capable of being closed from the main control room containment, using AC power and pneumatic air.

The primary containment atmosphere will be (2) The vent system is sized so that residual inerted with nitrogen during normal operation of core thermal power in the form of steam can the plant. Oxygen concentration in the primary containment wiil be maintained below 3.5 volume be passed through the relief piping to the percent measured on a dry basis.

st ack.

Following an accident, hydrogen concentration (3) The initial driving force for pressure will increase due to the addition of hydrogen relief is assumed to be the expected pressure setpoint of the rupture disks.

from the specified design-basis metal-water reaction. Hydrogen concentration will also (4) The rupture disks are constructed of increase due to radiolysis. _ Any increase in stainless steel or a material of similar hydrogen concentration is of lesser concern because the containment is inerted. Due to corrision resistance, dilution, additional hydrogen moves the i

(5) A number of rupture disks are procured at operating point of the containtnent atmosnhere farther from the envelope of flammabiliS.

6.2-16 j

Amendment 26

cynn

(,.9. C - 2 pyygg Ol'EM Jo,3 10 6.2.5.2.7 Recombiners Tale &reYLO (1)

The FCS consists of two permanently installed, 6- ~1-thermal hydrogen recombiners, with associated piping, valves, controls and instrumentation. The recombiner units are located in the secondary containment and controlled from the main control room. Each recombiner, as shown in Figure 6.2-40, removes gas from the drywell, recombines the oxygen with hydrogen, and returns the gas mixture, along with the condensate to the suppression chamber.Each recombiner unit is an integral package consisting of a blower, electic heater, reaction chamber, water spray cooler, a water separator, piping, valves, controls and instrumentation.

(2)

During operation of the system, gas is drawn from the drywell by the blower, and heated. Hydrogen and oxygen in the gas will be recombined into steam in the reaction chamber and condensed in the spray cooler.

The condensate and spray water, along with some of the gas, are returned to the wetwell. The rest of the gas is recycled through the blower. Cooling water required for operation of the system after a LOCA is taken f rom the RER system. The cooling water is used to cool the water vapor and the residual gases leaving the recombiner prior to returning them to the containment.

(3)

All pressure containing equipment, including piping between components is considered an extension of the containment, and therefore is designed to ASME Section III, Safety Class 2 requirements. Independent drywell and supression chamber penetrations are provided for the two recombiners. Each penetration has two normally closed isolation valves; one pneumatically operated and one motor operated. The system is designed to meet Seismic Category I requirements. The recombiners are in separate rooms in the secondary containment and are protected from damage by flood, fire, tornadoes and pipe whip.

(4)

Af ter a LOCA, the system is manually actuated from the control room when high oxygen levels are indicated by the containment atmospheric monitoring system (CAMS). (If hydrogen is not present, oxygen concentrations are controlled by nitrogen makeup). Operation of either recombiner will provide ef f ective control over the buildup of oxygen generated by radiolysis after a design-basis LOCA. Once placed in operation the system continues to operate until it is manually shut down when an adequate margin below the oxygen concentration design limit is reached.

e i

Task Type: FIXED Start Status: Done Start Date: 2/10/93

[

End Date: 3/1/93 Lead:

f Predecessors j

f 004 + PART IV - Program Assessment (E-*S) i Task Name 005.10 - WITS 9200199 - Conn.0!G !ssues i

Notes

}

Task Ccuplete. Update of Action Plan scheduled for completion by October 30, 1992. To T. Murley by 10/27.

t The November 10, 1992 memo to the Commission revised the Action Plan to include the three issues raised in the j

August 17,1992 memo from Selin to Taylor. This action was requested by the September 21, 1992, memo f rom Chi t k to Taylor.

~

Task Type: FIXED Start Status: Done Start Date: 10/5/92 End Date: 10/27/92 Lead: Widmann l

C Task Name 005.11 - WITS 9200200 - Ass.Due 6/30/93

.l Notes No fornal date assigned in WITS. "End of spring" date came

[

from 8/21/92 Murley memo. WRR will assess the broader

.{

aspects of the deficiencies in the review and response r

process from a global perspective (outside fire protection

}

program).

(

PMAS has the action on this WITS. Completion will be reflected in the Action Plan. Due to Murley on 6/25/93 i

Task Type: FIKED l

Start Status: Future Start Date: 3/1/93 End Date: 6/30/93 i

Lead-i t

Predecessors j

005.20 - Response to Commission's os (E-.S) 004.05 - Report on Findings (E-.5) l e

Task Name 005.12 - Comm. Briefing (Nov.13 1992)

Notes t

Brief Commission on progress of resolution w/T-Lag issues on Novencer 13, 1992.

1 Task Type:

SUMMARY

Start Status: Done l

TIME LIhE Detail Report Page 6 I

r

3 ABM ussioaxa nry e l

Standard Plant Table 6.2 10 Potential Bypass Leakage Paths 1 d

Termination Ie.akage Potential item Name Diameter Region 0)

Barriers (2)

Bypass Path (mm)

,, <,. r ')

, )?

X-1 U/D Equipment Hatch 2600 S

N

~.0 r

X-2 U/D Personnel Hatch 2600 S

No

,s*

f ',

X-3 ISI Hatch 200 S

No I DY' X-4 Wetwell Access Hatch

'2000 S

No r##7

,-h X-5 L/D PersonnelHatch 2400 S

No,,

o i

No y

X-6 L/D Equipment Hatch 2400 S

b No X-10A Mainsteam Line 1100 E

CO,G 3'

X-10B Mainsteam Line 1100 E

Cig 4 'No, *,

h X-10C Mainsteam Line 1100 E

C'- A Ye, %

X-10D Mainsteam Line 1100 E

Q{G X-11 Mainsteam Drain 600 E

C g, q.

X-12A Feedwater Line 950 E#

C T,' #

%h X-12B Feedwater Line 950 E i' CE i-

%h X-22 Borated Water Injection 40 S

No X-30A Drywell Spray 200 S

No X-30B Drywell Spary 200 S

No

. PCF (B) 600 S

No H

X-31A X-31B HPCF (C) 600 S

No X-32A LPFL(B) 650 S

No X-32B LPFL (C) 650 S

No X-33A RHR Suction (A) 750 S

No X-33B RHR Suction (B) 750 S

No X-33C RHR Suction (C) 750 S

No X-37 RCIC Turbine Steam 550 S

No X-38 RPV Head Spray 550 S

No X-50 CUW Pump Feed 600 S

No X-60 MUWP Suction 50 S

No X-61 RCW Suction (A) 200 E

C No X-62 RCW Return (A) 200 E

C No X-63 RCW Suction (B) 200 E

C No X-64 RCW Return (B) 200 E

C No X-65 HNCW Suction 150 E

C No X-66 HNCW Return 150 E

C No X-69 SA 25 E

C No X-70 1A 50 E

C No X-71A ADS Accumulator (A) 50 S

No X-71B ADS Accumulator (B) 50 S

No X-72 Relief Valve Accum.

150 S

No X-80 DrywellPurge Suction 550 E

C No X-81 Drywell Purge Exhaust 550 E

C No X-82 FCS Suction 100 S

No X-90 Spare 400 P

A No X-91 Spare 400 P

A No X-92 Spare 400 P

A No X-93 Spare 400 P

A No X-100A IP Power 450 S

No X-100B IP Power 450 S

No X 100C IP Power 450 S

No 6.2-5037 Amendment 17

r MM

3A61ocAn Rn' c Standard Plant Table 6.2-10 Potential Bypass leakage Paths 1 (Continued)

Termination Leakage Potential Item Name Diameter RegionC3)

Barriers (2)

Bypass Path l (mm)

X-143A C&I 100 S

No X-143B C&I 100 S

No X-143C C&I 100 S

No X-143D C&1 100 S

No X-144A C&I 100 S

No X-144B C&I 100 S

No X-144C C&I 100 S

No X-144D C&I 100 S

No X-146A C&I 300 S

No Notes:

1.

This Table provided in response to Question 430.52b.

2.

A) Penetration is napped

-l B) Terminates at Primary Containment Wall C) Terminates inside Secondary Containment D) Terminates outside Secondary Containment E) Redundant Containment Isolation Valves

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