ML20086P225

Forwards Nonproprietary Responses to Resolution of Issues Re Advanced BWR Draft SER

ML20086P225
Person / Time
Site:	05000605
Issue date:	12/19/1991
From:	Marriott P GENERAL ELECTRIC CO.
To:	Pierson R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
EEN-9191, NUDOCS 9112260268
Download: ML20086P225 (29)

Text

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GE WcIvar [ntryy l

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December 19,1991 )

MFN No.17191 Docket No. STN 50 605 EEN 9191 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Robert C. Pierson, Director Standardization and Non Power Reactor Project Directorate

Subject:

GC Responses to the Resolution of issues Itclated to AllWR DSER Chapters 1,3,9,10,11 & 13 (SECY 91235)

Rtference: GE Responses to the Resolution of 1ssues Related to ABWR DSER Chapters 1,3,9,10,11 & 13 (SECY.91235) (Proprietary Information), MFN No.172 91 dated December 19,1991 Enclosed are thirty four (34) copies of the GE responses to the subject issues.

Responses to issues pertaining to Section 11.4 contain information that is designated as General Electrie Company proprietory information. These responses are being submitted under separate cover (Reference).

It is intended that GE will arnend the SSAR, where appropriate, with these responses in a future amendment.

Sincerely, P.W. hbdriott, Manager Regulatory and Analysis Services

.MC 382,(408)925 6948 cc: ' F. A. Ross . (DOE)

N. D. Fletcher - (DOE)

C. Poslusny, Jr. (NRC)

R. C. Ber;;lund (GE)

J. F. Quir < (GE)

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SUMMARY

-RESPONSE To OUTSTANDING AND CONFIRMATORY ISSUES'0F [

ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)  !

(CONTINUED) t Outotanding

__ Issue Description Reaglution/ Comment i

27 ASME Code cases N-433 and See outstanding issue 18 of Ref.1 N-451 (5.2.1.2)-

28 TMI-2 Action items related See outstanding issue 19 of Ref.1 to safety / relief valves (5.2.2) .

29 Inservice inspection (ISI) of See outstanding issue 20'of Ref.1 reactor coolant pressure boundary (G.2.4).

30 Cleaning of stainless steel See outstanding issue 21 of Ref.1 components (5.3.1) 31 ' Residual heat removal design See outstanding issue 22 of Ref.1 compliance with BTP RSB 5-1 (5.4.7) 32 Preservice inspection (PSI) See outstanding issue 21 of Ref.1 (5.2.4) _F 33 TMI-2 action items related to See outstanding issue 24 of Ref.1 emergency core cooling systems-(5.4.6) 34 Engineered safety feature- See outstanding-issue 25 of Ref.1 materials (6.1.1) 35 Containment systems _(6.2) See outstanding issue 26 of Ref.1 ,

36 Containment leak testing See outstanding issue 27 of Ref.1 (6.2.6) 4

-37 ADS' issues (6.3.3) See outstanding issue 28 of Ref.1 38; -Control room habitability See outstanding issue 29 of Ref.1 39~ Atmosphere-cleanup systems See outstanding issue 30 of Ref.1 ,

(6.5) 40 ISIJfor class 2 and 3 piping See outstanding issue 31 of Ref.1 '

(6.6)

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SUMMARY

RESPONSE TO OUTSTANDING AND CONFIRMATORY ISSUES OF ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)

(CONTINUED)

Outotanding

__1Enge Descriotion Resolution / Comment 41 Main steam isolation valve See outstanding issue 32 of Ref.1 leakage control (6.7) 42 Instrumentation and controls Subject of another DSER (7) (SECY-91-294) 43 Electrical power systems (8) Subject of another DSER (SECY-91-355) 44 Seismic classification of Response provided in Amendment 18 spent fuel pool liner (9.1.2) (pages 3.2-18 and 9.1-2) 45 Spent fuel pool level and Response provided on page 9.1-4 leakage monitoring (9.1.2) attached 46 Spent fuel pool cooling and Protection acainst moderate _gngIgy cleanup system (FPC) design oloina failures issues (9.1.3) Response provided on page 9.1-5.1 attached Decay heat removal capability Response provided in Amendment 18 (pages 9.1-3 through 9.1-5.1)

SGTS sinale filter desian See outstanding issue 30 of Ref.1

47. Selamic classification of the Response provided on pages 9.1-6.1 new fuel stand (9.1.4) and 9.1-13 attached 48 overhead heavy load handling Response provided by rewrite of system (OHLHS) SSAR clarifi- Subsection 9.1.5 in Amendment 17 cation items (9.1.5) (pages 9.1-7 through 9.1-13)

, 50 Addition of RG 1.72 require- Response provided on page 9.2-13 ments to interface for attached ultimate heat sink (9.2.5) 51 Makeup water condensate Interface requirement; not an out-system flooding analysis standing issue (9.2.8)

SUMMARY

RESPONSE TO OUTSTANDING AND CONFIRMATORY ISSUES OF ADWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)

(CONTINUED)

Outotanding Issue Descriotion Resolution / Comment 52 Makeup water system Failure.of non-safety related *

(purified) design concern oortions durina an seismic event (9.2.10) Response provided on page 9.2-3 ,

attached ScoDe of MUWP system Provided in docket letter dated July 16, 1990

$3 Reactor building cooling RCW syste.m heat exchancer heat water cyutean heat removal desian capacity capacity and performance Response provided in Amendment 18 (9.2.11) (page 9.2.-3.1)

Four hour shutdown with loss of AC Dower GE/NRC agreed that this item was resolved on Amendment 14 at the GE/NRC meeting on plant systems open issues, March 4-6, 1991 Loss of an RWC system division Response provided on page 9.2-6 attached 54 Heating ventilation and air Response provided on Tablee 9.2-6 canditioning (HVAC) normal and 9.2-7 attached cooling water system safety related boundary, number of chillers, pumps (9.2.12) 55 HVAC emergency cooling water Revised P&ID and description of-system pump capacity, chemical the HECW system will be provided feed tank design, P&ID clari- in January 1992. They will resolve fication (9.1.13) 2:he capaciti of the pumps, the method of cooling diesel generator Zone C and the chemical addition tank will be shown to be non cfsty related.

56 SSAR clarification for turbine "esponse provided on page 9.2-10 building cooling water system attached (9.2.14)

- . . . .=_

SUMMARY

RESPONSE TO OUTSTANDING AND CONFIRMATORY ISSUES OF ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)

(CONTINUED)

Outstanding Issue Descriotion ___Faaslution/ Comment 57 Reactor service water perf- Response provided in Amendment 18 ormance ch&racteristics, (pages 9.2-12 and 9.2.13) design and location of pump (9.2.15) 58 Turbine service water system Response provided in Amendment 18 details (9.2.16) (pages 9.2-12.2 and 9.2-13) 59 Standby liquid control system GE has elected to automate SLCS.

compliance with 10CFR50.62 This modification will be made in (9.3.5) January 1992.

60 Control building KVAC design Revised P&ID will be provided in details (9.4.1.1) January 1992 61 Essential-electrical and The RCW pump and heat exchanger room reactor building cooling radioactivity level is alarmed with water equipment HVAC design the control room when it is above details, compliance with RG background level. An insignificant for component and filtration level of radiation is expected via specifications (9.4.1.2) Hx tube leaks over the 60 year life.

Filtration of the RCW rooms is not required to meet any 10CFR limit, and would cause a large increase in filter maintenance and testing for no benefit.

62 Engineered safety feature Revised P&ID will be provided in ventilation system design January 1992 detail (9.4.5.1) 63 Essential equipment design Revised P&ID will be provided in detail (9.4.5.2) January 1992 64 Essential diesel generator Revised P&ID will be provided in HVAC compliance with GDC 4, January 1992 design details (9.4.5.5) 65 Containment purge supply / Revised P&ID will be provided in exhaust system design January 1992 details (9.4.5.5)

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SUMMARY

RESPONSE TO OUTSTANDING AND CONFIRMATORY ISSUES OF ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235) i (CONTI!iUED)

Outstanding ,

Issue Descriotion Resolution / Comment 66 Nitrogen' accumulator and main Nitrocen accumulator seismic steamline seismic classifi- classification  :

cation ' (10.3) The backup nitrogen accumulators are non-seismi" Category I Main steamline seismic classif.

Main steamline classification is being developed-as-part of the ABWR open issues meetings. . Reoponso ,

will be'provided in January 1992. ,

67 Radiation monitoring in- Response provided in Amendment 18 turbine gland sealing (pages 11.5-15 and 11.5-17)'

system (10.4.3) 68 Circulating water system Table 3.4-2 deleted in Amendment 16. a flood protection (10.4.5) Response provided in revised Table 3.4-1-(page 3.4-8) attached.

69. Condensate and'feedwater- Discussed at GE/NRC Plant Systems. .

system power source (10.4.7) Branch meeting 10/29-30/91. Awaiting additional NRC guidance.

70' _ Condensate. storage tank local Response provided on page 9.2-2 alarm.(11.2.2) attached 71 Monitoring service building The service building ventilation ventilation exhaust (11.3.1) exhaust is sent to the plant release point where the off-gases are ,

monitored and sampled during release

-72 Lack of charcoal adsorber The mechanical vacuum pump exhaust ,

and filters (11.3.1) and the turbine building normal ventilation exhaust are not treated by charcoal adsorbers and/or HEPA ,

filters because calculations have shown:that these streams are of very low radioactivity and treatment is required. The calculations of exposure due to gaseous releases _'

consider these streams being released without treatment.

73 Containment secondary exhaust Response provided in Amendment 18 monitoring (11.3.1) (page 11.5-3) e, _

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SUMMARY

RESPONSE TO QUTSTANDING AND CONFIRMATORY ISSUES OF ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)

(CONTINUED)

Outetanding Issue Descriotion Resolution / Comment 74 Wet. Watto solidification GE proprietary information. Response process details (11.4.1) provided under separato cover.(page ,

11.4-4.1) l l

75 High integrity container Response provided in Amendment 18 l information (11.4.2) (page 11.4-4.1) ,

76 Incinerator fire protection Incinerator fire orotection i features, solid waste ship- GE proprietary information. Response ments (11.4.2) provided under separate cover (page 11.4-3)

Solid waste shioments GE proprietary information. Response provided under separate cover.

(revined Table 11.4-2, page 11.4-7) 77 . Release points (11.5.1) Interface requirement; not an out-standing issue 78 Service building exhaust Service buildina exhaust monitorina monitoring, turbine gland Response provided in Amendment 17 sealing system monitoring (page 11.5-8) and procedures, SSAR Table 11.5-5, RG 1.97 requirements Containment HVAC radiation monitor (11.5.2) sensitivity Response provided in Amendment 17 (page 11.5-3)

Gland seal orocess stream samplina and analysis Response provided on Amendment 17 (Table 11.5-7, iten 2) f39 1.d7 ILTViXD2.8!DLfi Response provAded on Attachment A to thin summary table 79 Chapter 12 Subject of another DSER 80 Chapter 13 No outstanding issues in this DSER 81 Chapters 14,15,16,18 & 19 Subjects of other DSERs thru and THI action plan issues 86

SUMMARY

RESPONSE TO OUTSTANDING AND CONFIRMATORY ISSUES OF ABWR DSER CHAPTERS 1,3,9,10,11 & 13 (SECY-91-235)

Confirmatory issuo Descriotion Resolution / Comment 1 Fuel licensing acceptance See confirmatory issue 1 of Ref.1 criteria (4.2)  !

2 BWR stability (4.4) Informally resolved at the GE/NRC Reactor Systems Branch 11/20-21/91 meeting. Information will be placed on docket in January 1992.

3 Pressure-temperature limits See confirreatory issue 3 of Ref.1-(5.3.2)

( Neutron fluence (5.3.2) Sne confirmatory issue 4 of Ref.1 5 Reactor water cleanup systen See confirmatory issue 5 of Ref.1-temperature capability (5.4.8)-

6 KVAC normal cooling water Safety-related portions include safety-related portions isolation valves for primary (9.2.12) containment. There are no isolation valves for secondary containment penetrations.

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ATTACHMENT A RESPONSE TO RG 1.97 PORTION OF OUTSTANDING ISSUE 78

a. The types of process radiation detectors used in the plant and for accident monitoring are all defiled in Tables 11.5-1 & 11.5-2. The calibration frequency cnd techniques of calibration, dependent on the equipment vendor's data for the concors, will be prov>4;d i the COL applicant's plant operation and maintenance manual.

b. The location of' vadircio'n detectors are specified in Table 11.5-1 and are daccribed in Subsectica ?.). 5.2 for the appropriate subsystems. The location of tha sampling points VA42 to extract representative samples are normally cpacified in the re'spet.Lve system P&ID. The sampling frequency for each meccurement is specified An ' tables 11.5-4 through 11.5-7.

c. The location of instruieht readouts and methods of recording are specified in the system descript: ion for each of the subsystems defined in Subsection 11.5.2. Also, thin information is indicated to a great extent in the system IED, Figure 7.6-5.

d. Most measurements and readouts are continuously provided by the monitoring instrumentation at all times for operator review. The use of process computer end chart recorders provides for logging of measurements for historical rccords.

o. The description of procedures or calculation methods for the conversion of instrument readouts to release rates are equipment vendor dependent. This information will be provided in the COL applicant's operation and maintenance manual.

f. The application of the shielding requirements for low radiation exposure as ctipulated in NUREG+0737 Item II.F.1, clarification 2 of Attachment 2 is vendor cquipment dependent and will be included in the equipment specifications.

D :cription of the sampling system design to show compliance with the required regulatory shielding design basis will bo provided in the COL applicant s op3 ration and maintenance manual,

g. The design of the sampling system that monitors for air particulates and lodines does provide for nozzle entry of the gas samples and will maintain icokinetic conditions during and following an accident as stipulated in NUREG-0737 Item II.F.1, clarification 3 of Attachment 2. Two isokinetic probes cro utilized by the plant vent discharge monitoring subsystem, one is a high range monitor for uso during post-accident operation (refer to Subsection 11.5.2.2.4). The sampling flow rate relative to the flow rate through the stack in adjusted to naintain isokinetic conditions to within 20%. Any high or low campling flow will be annunciated in the control room. This sampling technique does meet the regulatory requirements and will be described in detail in the COL applicant's plant operation and maintenance manual.

h. The sampling system design as described above under item (g) does include provisions to permit the collection of representative samples of radioactive iodines and particulates during and following an accident. These measurements cro used to determine the quantitative releases for dose calculations and assessment as required by NUREG-0737, Table II.F.1-2. Description of this technique will be provided in the COL applicant's plant operation and maintenance manual. f

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ABWR se mim pry n Standard Plant _ _ _ , , ,

Table 3.41 STRUCTURES, PENETRATIONS AND ACCESS OPENINGS DESIGNED FOR FLOOD PROTECTION Reactor Senict Control Radwaste Turbine Structurt Buildina Building Buildina Buildine Building

%m4. -!:. ; 2- - : .. L - :;.; 2 N ll.7e e ll'f e e tno il 4%fe- u##,

- e- M,7ee Design Mood Level (mm) 11,700 '7 9Q Referenc: Plant Grade (mm) 12,000 -

Base Stab (mm) - 8,200 4,^ ' . . td%66 W  %)

-c m -t,t.. - s ,5 5,s Actual Plant Grade (mm) 12,000 h'*

• h "

49,700 16,700 19,700 23,000 49,350 g Building Height (mm)

E None

• Penetrations Below Design Refer to None Refer to None Mood Level Table 6.2-9 Table 6.2 9 for RCWlines Access Openings Below Tunnel from 5/B Men Entrance Tunnet from 5/B rye Tunnet Tunnet from s/D Design Mood Level 6 '-.. m Q grace levst $ tenmm.l!X from R/B&T/D 6 F=6vem g 3.500mmTMSL Aru Accus GNm 7.3. O

' from 5/B 6 Note 3 I N $e ia,.r. *

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3. Thrl1Tes* tdt rTf58ugPfhe dwaste building tunnel are not exposed to outside ground flooding.

Amendment 16

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1 MkN tusxxwr ev e Standard Plant sizing of each minimum recirculation flow path is prior to plant startyp. The method of assessing evaluated to assure that its use under all the loads, the method of sizing the actuators, analyzed conditions will not result in and the setting of the torque and limit switches degradation of the pump. The flow rate through will be specifically addressed. (See Subsection minimum recirculation flow paths can also be 3.9.7.3 for interface requirements),

periodically measured to verify that flow is in accordance with the design specification. The in service testing of MOVs will rely on diagnostic tecbiques that are consistent with The safety related pumps are provided with the state of the art and which will permit an instrumentation to verify that the net positive assessment of the performance of the valve under suction head (NPSH)is greater than or equal to actualloading. sJdOVs that fail the acceptance q the NPSH required during all modes of pump criteria, and are ' declared inoperable,' for operation. Ther numps can be disassembled for stroke tests and leakage rate can be evaluation wbH. the Code Section XI testing disassembled for evaluation. The Code provides results in a deviation which f alls within the criteria limits for the test parameters

' required action range.' The Code provides identified in Table 9 9 8. e l'1a criteria limits for the test parameters indentified in Table 3.9 6. 3.9.6.2.3 Isolation Vahe tsak Tests

/

3.9.6.2 Inservice Testlug of Safer).Related The leak. tight integrity will t e verified Yahes for each valve relied upon to provide a leak tight function. These valves include:

3.9.6.2.1 Check Yahes (1) pressure isolation valves valves that

• All ABWR safety related pipir.1 systems provide isolation of pregiure differential incorporate provisions for testing to demonstrate from one part of a system from another or the operability of the check valves under design between systems; conditions. In service testing will incorporate Ibc use of advance non intrusive techniques to (2) temperature isolation valves valves whose periodically assess degradation and the leakage may cause unascc i. table thermal performance characteristics of the check vahes, loading on supports or stratification in the The Code Section XI tests will be performed, and piping ar.d thermal loading on supports or check valves that fail to exhibit the required whose leakage may cause steam binding ci performance can be disassembled for evaluation. pumps; and The Code provides criteria limits for the test g 18 parameters identified in Table 3.9 8. (3) containment isolation valven valv.s tb31 /

1. Perform a containment isolation function /

'- 15/- 3.9.6.2.2 Motor Operated Valves including valves that may be exempted from Appendix J, Type C, testing but whose The motor operated valve (MOV) equipment leakage may cause loss of suppression pool specifications require the incorporation of the results of either in situ or prototype testing water inventory.

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with full flow and pressure or full differential Leakage rate testing of valvesfolher thaj) s pressure to verify tbc proper sizing and correct (coniainment isolation valves 5ill be in g switch settings of the valves.sTbc applicant accordance with the Code Section XI. An example referencing the ABWR design will provide a study is the fusible plug valves that provide a lower to determine the optimal frequency for valve drywell flood for severe accidents described in stroking during in service testing such that S u b s e c tio n 9.5.12. Tbc valves are unnecessary testing and damage is not done to the safety related due to the function of retaining valve as a result of the testing. (See Subsection suppression pool water as shown in Figure 3.9.7.3 for interface requirements). 9.5 3. These specist valves are noted here and not in Table 3.9 8. Tbc fusible plug valve is a The concerns and issues identified in nonte: losing pressure re!ief desice and the Code Generic Letter 89-10 for MOVs will be addressed requires replacement of each at a maximum of 5 year intervals.

Ameedmes 14 3K1 l

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SECTION 3.9.6 Inservice Testing of Pumps & Valves INSERT 14 A program will be developed by the applicant raferencing the ABWR design to establish the frequency and the extent of disassembly and inspection based on suspected degradation of all safety related pumps, including the basis for the frequency and the extent of each disassembly. The program may be revised throughout the plant life to minimize disassembly based on past disassembly experience. (See Subsection 3.9.7. 3 (1) for interface requirements.)

INSERT 15 A program will be developed by the applicant referencing the ABWR design to establish the frequency and=the extent of disassembly and inspection bcsod on suspected degradation of all safety related check valves, including the basis for the frequency and the extent of each disassembly. The program may be revised throughout the plant life to minimize disassembly based on past disassembly experience. (See Subsection 3.9.7. 3 (1) for interface requirements.)

INSERT 16 Guidelines to justify prototype testing are contained in Generic Letter 98-10, Supplement 1, Questions 22 and 24 through 28.

INSERT 17a A program will be developed by the applicant referencing the ABWR design to establish the frequency and the extent of disassembly and inspection based on suspected degradation of all safety related "MOV's", including the basis for the frequency and the ext 9nt of each disassembly. The program may be revised throughout the plant life to minimize disassembly based on past disassembly experience. (See l

Subsection 3.9.7.3 (1) for interface requirements.

INSERT 17b Periodic testing will be conducted under adequate differential pressure and flow conditions that allow a

-justifiable demonstration of continuing Mov capability for design basis conditions, including recovery from inadvertent i valve positioning.

INSERT 18 in accordance with the Evaluation Against criterion 54, Subsection 3.1.2.5.5.2, i

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2sA61%AE 91v n Standard Plant 3.9.7 Interfaces 3.9.7J Pump and Yahe loserdce Testing Program 3.9.7.1 Reactor Internals Vibration Analpls,  ;

Measurement and laspection Program Applicants referncing the ABWR destga will provide a plan for the detailed pump and valve '

The first applicant refereacluit the ABWR inservice testing and inspection program. This design will p: ovide, at the time of application, plan will s.be results of the vibration assessment program ,.

for the ABWR prototype internals. There results (1) include baseline pre service testing to willinclude the following information specified support the periodic in service testing of in Regulatory Guide 1.20. the components required by technics!

specifications. Provisions are included to R.G.120 Subiect disassemble and inspect the pump, check ve'ves, and MOVs within the Code and C.2.1 Vibratlon Analysis safety related classification as necessary, Program depending on test results. (See Subsections . 34 C.2.2 Vibration Measurement 3.9.6 , '), 9, 4, ! , 4. 9. G 2 . I e d i5 Program 1.9.G.1.1 17a C.2.3 Inspection Program (2) Provide a study t determine the optimal C.2.4 Documentation of frequency for valve stroklug during Results inservice testiag. (See Subsection 3.9.6.2.2)

NRC review and approval of the above information on the first applicants dor.ket will (3) Address the concerns and issucs identified complete the vibration assessroent program in Generie 1.etter 8910; specifically the .

requirements for prototype reactor internals, method of assessment of t'ae loads, the method of string the actuators, and the in addition to the information tabulated setiing of the torque and limit switches.

above, the first applicant referencing the ABWR (See Subsection 3.9.6.2.2) ,

design will provide the information on the schedules in accordance with the applicable 3.9.7.4 Audit ot Desigu Specification and portions of poition C.3 of Regulatory Guide 1.20 Design Reports for non prototype internals.

Applicants referencing the ADWR design will Subsequent applicants need only provide the make available to the NRC staff design information on the schedules in accordance with specification and design reports required by the applicable portions of position C.3 of ASME Code for vessels, pumps, valves and piping Regulatory Guide 1.20 for non prototype systems for the purpose of audit. (See internals. (See Subsection 3.9.2.4 for interface Subsection 3.9.3.1) requirements).

3.9.8 References 3.9.7.2 ASME Class 2 or 3 or Quality Group Componenta wlth 60 Year Design Life 1. BHR Fuel Channel Mechanical Drsign and Applicants referencing the ABWR design will identi'v ASME Class 2 or 3 or Quality Group D 2. BHR/6 fuel Assembly Evaluation of Combined components that are subjected to loadings which Safe Shutdown Earthquake (SSE) and could result in thermal or dynamic fatigue and Loss of Coolant Accident (LOCA) Loading, provide the analyses required by the ASME Code, NEDE.21175 P, November 1576.

Subsection NB. These analyses willinclude the appropriate operating vibration loads and for the 3. NEDE 24057 P (Class !!!) and NEDE 24057 effects of mixing hot and cold fluids (See (Class 1) Assessment of Reactor Internah.

Subsection 3.9.3.1 for interface requirements). Vibration in BWR/4 and BWR/5 Plants.

3 9-1$

Amendment 16

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i ABWR ==

Standard Plant Prv n (2) surface dirt dislodged from equipment control rootn and a local panel. Pump low suc.

immersed in the pool; tion pressure automatically turns off the pumps. A pump low uischarge pressure alarm is (3) crud and fission products emanating from the indicated in the control room and on the local reacto or feel bundles during refueling; pant:1. The circulating pump motors can be -

powered from the diesel. generators if normal (4) debris from isspection or disposal opera. power is not available. Circulating p;mp motor tions; and loads are considered nonessentialloads and will be operateo as required under accident (5) residual cleaning chemicals or flush water, eonditions.

A post strainer in the effluent stream of the The water level in the spent fuel storage filter.demineralizer limits the migration of pool is maintained at a height which is suffi.

filter material. The filter holding element can cient to provide shielding for normal building withstand a differential pressure greater than occupancy. Radioactive particulates removed the developed pump head for the system. from the fuel pool are collected in filter.de.

mineraliter units which are located in shielded The filter demineralizer uaits are located cells. For these reasons, the exposure of plant separately in shielded cells with enough clear. personnel to radiation from the FPC system is ance to permit removing filter elements from the minimal. Further details of :=diological vestels. considerations for this system are described in Chapter 12.

Each cell contains only the filter.deminera.

liter and piping. All valves (inlet, outlet, The circulation patterns within the reactor recycle, vent, drain, etc.) are located on the well and spent fuel storage pool ne establisued outside of one shiciding wall of the room, by placing the diffusers and skimmers so that together with necessary piping and headers, particles dislodged during refueling operations instrument elements and controls. Penetrations are swept away from the work area and out of the through shielding walls are located so as not to pools, compromise radiation shielding requirements.

. Check valves prevent the pool from alphoning Tbc filter.demineralizers are controlled from in the event of a pipe rupture.

a local panel. A differential pressure and conductivity instruments provided for each Heat from pool evaporation is handled by the filter.demineralizer unit indicate when backwash building ventilation system. Makeup water is is required. Suitable alarms, differential provided through a remote. operated valve.

pressure indicators and flow indicators monitor the condir n of the filter demineralizers. 9.1.3.3 Safety Evaluation System. instrumentation is provided for both The maximum possible heat load for the FPC automatic and remote manual operations. A low. system upon closure of the fuel gates (21 days) low level switch stops the circulating pumps when is the decay heat of the full core load of fuel the fuel pool drain tank reserve capacity is at the end of the fuel cycle plus the remaining reduced to the volume that can be pumped in decay heat of the spent fuel d3scharged r.t approximately orie minute with one pump at rated previous refuelings upoa clost.re of the fuel capacity (250 m 3/ht). A level switch is gates; the maximum capacity of the spent fuel provided in the fuel pool to alarm on high and storage poolis 270% of a core. The teroperature low level. A temperature element is provi e to of the fuel pool water may be permitted to rise display pool temperature in the main control to approximately 140'F under these condi.

room. in addition, leakage flow detectors in the tions. During cold shutdown conditions,if it pool drains and pool liners are provided and appears that the fuel pool temperature will alarmed in the control room. exceed 125'F, the operator can connect the FPC system to the RHR system. Combining the ca.

The circulating pumps are controlled from the pacities enables the two systems to keep the og l o c ct \ \ g )

cow +2lomeluOe@

room.

ABWR zmimui Standard Plant PEV B water supply (yard main, one motor driven pump and water soutec) are seismically designed. The motor driven pump is powered from a bus which nat.

s uf:ty related diesel generator as one of its power sources. A second seismically designed pump, directly driven by a dier.cl engin:is also pro 5ided.

46 Tram'she foregoing analysis, it is concluded that th: JPC system meets its design bases.

91.3.4 luspection and Testing Requirements No special tests are required because, normally, one pump, one heat exchanger and one filter demineralizer are operating while fuel is stored in the pool. The spare unit is operated periodically to handle abnormal heat loads or to replace a unit for servicing. Routine visual icspection of the system components, instrumen-tation and trouble alarms is adequate to verity system operability.

9.1.3.5 Radiological Considerations The water level in the sper;t fuel storage pool is maintained at a height which is suffi-cient to provide shielding for normal building occupancy. Radioactive particulates removed from the fuel pool are collected in tilter-demineralizer units which are located in shielded cells. For these reasons, the exposure of plant personnel to radiation from the FPC system is minimal. Further details of radiological considerations for this and other systen;s ar. des;ribed in Chapteis 11,12, and 15.

Amendment 18 9l-51

INSERT 4G The FPC components, housed in the Seismic Category I reactor building, are Seismic Category I, Quality Group C including all components except the filter demineralizer. Thece momponents are protected from the effects of nature.1 phenomena, +:5 as:

earthquake, external flooding, wind, tornado ans e,t.rnal missiles. Inside the reactor building the FPC sate-y-related components are protected from the effects of pipe whip, internal flooding, internally generated missiles, and the effectn of a moderate. pipe rupture within the vicinity.

- ABWR m 6u m i Standard Plant %n 9.1.4.2.2 Overhead Bridge Cranes 9.1.4.23.2 New FuelInspection Stand 9.1.4.2.2.1 Reactor Building Crane The new fuelinspection stand (Figure 9.14) senes as a support for the new fuel bundles uradergoing re.

The reactor building crane is a seismically analyzed ceiving inspection and provides a wotking platform piece of equipment. The crane consists of two crane for technicians engaged in performing the inspection.

girders and a trolley which carries two hoists. The runway trask, which supports the crane girders, is The new fuelinspection stand consists of a vertical supported from the reactor building walls at eleva- guide column, a lift unit to position the work platform tion 34,600. The trolley tr. 4 Lterally on the crane at any desired level, bearing seats and upper clamps girders carrying the c:ain hoist and auxiliary hoist, to hold the fuel bundles in position.

AT The reactor building crane is used to move all of 9.1.4.233 ChannelBon Wrtoch the major components (reactor vessel head, shroud head and separator, dryer assembly and pool gates) The channel bolt wrench (Figure 9.15) is a manu-as required by plant operations. The reactor build- ally operated device approximately 3.76 meters (12 ft) it.g crane is used for handling new fuel from the re- in overalllength. The wrench is used for removing and installing the channel fastener assembly while the l actor new fuel building inspection stand entry and hatch the spent to fuelnew pool. Itfuelfuelstorage, assembly isthe held in the fuel preparation machine.

also is used for handling spent fuel cask. The The channel bolt wrench has a socket which mates principal design criteria for the reactor building and captures the channel fastence capscrew.

crane are described in Subsection 9.1.5.

9.1.4.23.4 Channel liandling Tool 9.1.4.23 Fuel Servicing Equipment The channel handling tool (Figure 9.1-6) is used in The fuel senicing equipment described below has conjunction with the fuel prrparation rnachine to been designed in accordance with the criteria listed remove, install and transport fuel channels in the in Table 9.12. Items not listed as Seismic Category spent fuel pool. l 1, sucL as hoists, tools and other equipment used for servicing shall either be removed during operation, The toolis composed of a handling bail, a moved to a location where they are not a potential lock / release knob, extension shaft, angle guides and Lazard to safety related equipment, or seismically re- clamp arms which engage the fuel channel. The strained to prevent them from becoming missiles. clamps are actuated (enended or retractedj by manu-ally rotatinglock/ release knob.

9.1.4.2J.1 Fuel Prep Machine The channel handling toolis suspended by its bail Two fuel preparation machines (Figure 9.13) are from a spring balancer en the channel handling boom l mounted on the wall of the spent fuel pool and are located on the spent fuel pool periphery. l used for stripping reusable channels from the spent fuel and for rechanneling of the new fuel. The ma- 9.1.4.23.5 Fuel Pool Vacuum Sipper chines are also used with the fuelinspection fixture to provide an underwater inspection capability. The fuel pool vacuum sipper (Figure 9.17) pro-vides a means of identifying fuel suspected of having Ea6 c e! preparation machine consists of a work cladding failures. The fuel pool Vacuum sipper con-platform, a frame, and a movable carriage. The sists of a fuelisolation container, fluid console, moni-frame and movable carriage are located below the toring console with program controller and beta de-l normal water level in the spent fuel pool thus provid. tector and the inter-connecting tubing and cables.

ing a water shield for the fuel assemblies being han. The suspected fuel assembly is placed in the isolation died. The fuel preparation machine carriage has a container. A partial Vacuum is established in the gas permanently installed up-travel stop to prevent rais- volume above the fuel assembly. The fission product ing fuel above the safe water shield level. gas leakage ' sensed by the beta detector and moni-loring console.

1141 l Amendment 16

INSERT 47 The new fuel inspection stand will be firmly attached to the wall so that it does not fall into or dump personnel into the spent fuel pool during an SSE. (See Subsection 9.1.6.5 for interface requirements).

MlN .

33A6100All Standard Plant nev e 9.1.6 Interfaces 9.1.6.1 New Fuel Storage Racks Criticality Analysis The applicant referencing the ABWR design shall provide the NRC confirmatory criticality analysis as required by Subsection 9.1.1.1.1.

9.1.6.2 Dynamic and Impact Analyses of New Fuel Storage Racks The applicant retrencing the ABWR design shall provide the NRC confirmatory dynamic and imp ct analyses of the new fuel storage racks. See Subseetion 9.1.1.1.6.

9.1.63 Spent Fuel Storage Ra,cks Criticality Analysis The applicant referencing the ABWR design shall provide the NRC confirmatory critically analysis as required by Subsection 9.1.23.1.

9.1.6A Spent Fuel Racks Load Drop Analysis The applicant referencing the ABWR design shall provide the NRC confirmatory load drop analysis as required by Subsection 9.1.43.

9.1.7 References

1. General Electric Standard Application for Reactor Fuel, (NEDE-24011-P-A, latest approved revision).

[ 9.1. G. S N ew Nei T w r pa c.h o r w J S e.s 3% c. Cc,.p6 k d) ha ts ad v&+c., w 3 M ABWR. dt s g w ud gw ske \\ k VN 4 w w4.,k t W T 9 4 cbow Sh* Wh [trvek 40 4 W A kl So k c sk dOAA Wok (cd ( gwko o v- wm p {-ev-t ow n d bsk o

-A. s g 4 t cLm q ~ sse.rsu sh A-A 9. l. 4 . 7 3. 2. )

9.1 13 Amendment 17 E

ABM nwooxn REV B Standard Plant _

(e) MUkC runsfer pumps (see Table 9.2 3) boundary; (three SN gpm at 141 psi head.)

(2) capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) Water can be seni to the CST from the following sources:

(3) ability to prevent or mitigate the conse-(a) MUWP pumps quences of events that could result in po-tential offsite exposures.

(b) CRD system The MUWC system is not safety-related.

(c) radwaste disposalsystem However, the systems incorporate features that assure reliable operation over the full range of (d) condensate demineralizer system effluent normal plant operations.

(main condenser high level relief) 9.2.9.4 Tests and Inspections (4) Associated receiving and distribution piping valves, instruments, and controls shall be The MUWC system is proved operable by its use provided, during normal plant operation. Portions of the system normally closed to flow can be tested to ensure operability and the integrity of the system.

(5) Overflow and drain from the CST shall be sent to the radwaste system for treatment. The air operated isolation valves are capable of being t-sted to assure their operating (6) Any outdoor piping shall be protected from integrity by manual actuation of a switch freezing. located in the control room and by observation of associated position indication lights.

Flow to the various systems is balanced by (7) All surfaces coming in contact with the means of manual valves at the individual takeoff condensate shall be made of corrosion- points. Divisional isolation valves are instal-resistant materials, led at the primary containment boundaries.

(8) All of the pumps mentioned in (2) abow 9.2.10 Makeup Water System (Purified) shall be located at an elevation such that Distribution System adequate suction head is present at all water levels in the CST. 9.2.10.1 Design Ilases (9) Instrumentation shall be provided to indi. (1) The makeup water-purified (MUWP) cate CST water level in the main control distribution system shall provide makeup room, s water purified for makeup to the reactor

' coolant system and plant auxiliary systems. _

(10) Potential flooding is discussed in Subsection 3.4. Potential flooding from (2) The MUWP system shall provide purified water lines within the reactor building and the to the uses shown in Table 9.2-2.

control building are evaluated in Subsection ,

3.4.1.1.1. (3) The MUWP system shall provide water of the  ;~

quality shown in Table 9.2-2a. If these 9.2.9.3 Safety Evaluation water quality requirements are not met, the water shall not be used in any safety-Operation of the MUWC system is not required related system. The out-of spec water shall to assure any of the following conditions: be reprocessed or discharged.

(1) integrity of the reactor coolant pressure (4) The MUWP system is not safety.related.

Amendment 17 . 922

l lNSG4T 7D High water _ level shall be alarmed both locally and in the main control room.

ABM 2mimsii RM B Standard Plant ,

(5) All tanks, pumps, piping, and other equip- purified water storage tank level in the ment shall be .nade of corrosion resistant main control room, materials.

(8) Continuous analyzers are located st the (6) The system shall be designed to prevent any demineralized water makeup system and at any l radioactive contamination of the purlfied demineralized water storage tank. These are water. supplemented as needed by grab samples.

Allowance is made in the water quality (7) The interfaces between the MUWP system and specifications for some pickup of carbon all safety related systems are located in dioxide and air in any demineralized water l l storage tank. The pickup of corrision the control building or reator building which are Seismic Category I, tornado- products should be minimal because the MUWP missile resistant and flood protected piping is stainless e eel.

1. structures. The interfaces with safety-

%u related systems are safety related valves which are part of the safety related (9) Intrusions of radioactivity into the MUWP system from ether potentially radioactive fi V systems, systems are prevented by one or more of th:

foHowing:

52.

(8) Safey related equipment located by portions of the MUWP system are in Seismic Category I (a) check vans in the MUWP line; structures and protected from all system impact. (b) air (or syphon) breaks m the MUWP lines 9.2.10.2 System Description (c) the MUWP system lines are pressurized while the receiving system is at The MUWP system P&lD is shown in Figure 9.2-5. essentially atmospheric pressure.

7 This system includes the following:

(d) piping to the user is dead ended.

l(1) Any purified water storage tank shall be provided outdoors with adequate freeze (10) There are no automatic valves in the protection and adequate diking and other MUWP system. During a LOCA, the means to control spill and leakage. safety related systems are isolated from the MUWP system by automatic valves in (2) Two MUWP forwarding pumps shall take suction the safety related system.

l from any purified water storage tanks. They shall have a capacity of 308 gpm and a 9.2.10.3 Safety Evaluation discharge head of 114 psi.

Operation of the MUWP system is not required (3) Distribution piping, valves, instruments and to assure any of the following conditions; controls shall be provided.

(1) integrity of the reactor coolant pressure (4) Any outdoor piping shall be protected from boundary; freczing.

(2) capability to shut down the reactor and (5) All surfaces coming in contact with the maintain it in a safe shutdown condition; or purified water shall be made of corro-sion resistant materials. (3) ability to prevent or mitigate the conse-quences of events which could result in (6) All pumps shall be located at an elevation potential offsite exposures.

such that adequate suction head is present l at all levels in a purified water storage The MUWP system is not safety-related.

tanks. Ilowever, the systems incorporate features that assure reliable operation over the full range of (7) Instruments shall be provided to indicate normal plant operations.

Amendment 17 9.2-3

IN S G RT 57_

The portions of the MUWP system, which upon their failure 1 during'a seismic event can adversely impact structures, sys-tems, or components important to safety, shall be designed to assure their integrity under seismic loading resulting from a safe shutdown earthquake 4

AB R utsworn Standard Plant Rev n but not to a temperature that would damage ,

System components and piping materials are equipment or require an immediate shutdown.

selected where required to be compatible with the available site cooling water in order to minimize 9.2.11.4 Testing and Inspection Requirements corrosion. Cathodic protection of the tubing side of the heat exchanger shall be provided. . The RCW system is designed to permit periodic Adequate' corrosion safety factors are used to in service inspection of all system components assure the integrity of the system during the to assure the integrity and capability of the life of the plant. ' ,

system.

Duriq all plant operating modes, all The RCW system is designed for periodic pres. ,

divisic- ave at least one RCW cooling water sure and functional testing to assure; (1) the pump q,.2 ting. Therefore, if a LOCA occurs, the structural and leaktight integrity by visible RCW cooling water system required to shut down inspection of_ the components; (2) the-the plant safely is already in_ operation. If a operability and the performance of the active loss of offsite power occurs during a LOCA, the components of the system; and (3) the pumps momentarily stop until transfer to standby operability of the system as a whole, diesel ge':erator power is completed. The pumps are restarted automatically according to the The tests shall assure, under conditions as diesel loading sequence. If a LOCA occurs, most close to design as practical, the performance of nonsafety related components are automatically the full operational sequence that brings the isolated from the RCW system. Censequently, no system into operation for reactor shutdown and operator action is required, following a LOCA, to for LOCA, including operating of applicable start the RCW system in its LOCA operating mode, portions of the Reactor Protection System an.!

the transfer between normal and standby power All heat exchanprs and pumps will be required sources. These tests shall include periodic during the following plant operating conditions, testing of the heat removal capability of each in addition to LOCA: shutdown at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, RCW heat exchanger. Each of these heat shutdown at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and hot standby with loss of exchangers has been designed to provide 20%.

AC power. margin above the heat removal capability required for LOCA in Tables 9.2 4a, b and c.

Loss of one RCW oivision will result in loss- The revised heat removal capacity of the heat of RCW cooling to every other RIP (five total) as exchangers is shown in Table 9.2 4d. This 20%

shown on RR5 P&ID (Figure 5.4 4) and will cause margin is provided to compensate for the those five RIPS to runback to minimum speed.- The ~ combined effects of fouling and tube plugging.

RIP M G set in the same electrical division, When this snargin is no longer present, the heat which is cooled by the sarne RCW division which exchanger heat removal capacity will be failed and powers two more RIPS, would stop by increased by either cleaning or retubing.

M G set cooling water prptection. This would completely shutdown three RIPS and would have the The RCW system is supplied with a chemical

. resulting total of seven RIPS either at minimum addition tan! to add chemicals to each speed or stopped.l Assuming the event began at division. The RCW system is initially filled full power on the 100% Control Rod Line, the with demineralized water. A corrosion inhibitor resulting temporary reactor power would be can be added if desired. These measures are ad.

approximately 60% power. The operator would then equate to protect the RCW system from the ill correct the RCW problem or initiate a normal effects of corrosion or organic fouling.

plant shutdown.

I-S3  : The RCW system is designed to conform with

j. The drywell cooling system can perform its the foregoing requirements. Initial tests shall l function after the loss of any RCW division, be made as described in Subsection 14.2.12.

l With only one RCW division and one drywell cooler i

operating, the drywell temperature will increase l

Amendment 18 916

, + - -. -- , w.-

l I N S ERT 53 Complete failure of any RCW division will reduce drywmil cooling, but, not enough to require plant shutdown or power level reduction. Failure of RCW division A would have only one drywell cooler using RCW cooling and the normal HNCW cooling. Drywell temperatures would not increase enough to adversely affect any drywell components.

k 2 --

" ~ s"t :AB M . z w ioa m

- Standard Plant _ arv s evaporator, if th temperature of the chilled heat exchanger.

water' drops below'a specified level, the controller automatically adjusts the position of 9.2.14.2 System Description the compressor inlet guide vanes. Flow switches prohibit the chiller from operating unlen there 9.2.14.2.1 Gereral Description is water flow through both evcporator and condenser. The TCW system is illustrated on Figure 9.2 6. The system is a single loop system and 9.2.14 Turbine Building CoolineWater System consists of one surge ank, one chemical addition tank,' two pumps with a capacity of a 9.2.14.1 Design Bases 29,000 gpm each, two heat exchangers with heat l $

removal capacity of 130 x 106 Blu/h each.  ?

9.2.14.1.1 Safety Design Bases -(connected in paralle!), and associated coolers, piping, valves, controls, and instrumentation.

The turbine building cooling water (TCW) Heat is removed from the.TCW system and system serves no safety function and has no transferred to the non. safety related turbine safety design basis, service water system (Subsection 9.2.16),

2 There are no connections between the TCW A TCW system sample is periodically taken 3 system and any other safety.related systems. for analysis to assure that the water quality meets the chemical specifications.

92.14.12 Power Generation Design Bases .

9.2.14.22 Component Description (1) = The TCW system provides corrosion-inhibited, demineralized cooling water to all turbine Codes and standards applicable to the TCW island auxiliary equipment listed in Table system are listed in Table 3.21. The system is

-9.2 11. designed in accordance with quality group D  !

specifications. i (2) During power operation, the TCW system operates to provide a continuous supply of The chemical addition tank is located in the cooling water, at a maximum ternperature of turbine building in close proximity to the TCW 1050 F, to the turbine island auxiliary system surge tank. -!

equipment,~ with a service water inlet temperature not exceeding 950F. The TCW pumps are 100% capacity each and are constant speed electric motor driven,13orizontal 'g.

-(3)._ The TCW system is designed to permit the _ centrifugal pumps. The etaeQumps are maintenance of any single active component connected in parallel with common suction and without interrt ption of the cooling disch. 'ines, 56 function.

The TCW heat exchangers are 100% capacity (4) Mneup to the TCW system is designed to each and are designed to have the TCW water permit continuou'ssystem operation with circulated on the shell side and the power cycle design failure leakage and to permit heat sink water circulated on the tube side, expeditious post. maintenance system refill. The surface area is based on normal heat load.

(5)' The TCW system is designed to have an The surge tank, which is shared between the l atmospheric surge tant located at the HNCW and TCW systems, is an atmospheric carbon

-highest point _in the system. steel tank located at the highest point in the TCW system. The surge tank is provided with a (6) The TCW system is designed to have a higher level control valve that controls makeup water pressure than the power cycle heat sink addition.

water to ensure leakage is from the TCW system to the power cycle heat sink in the The surge tank is located above the TCW pumps esent a tube leak occurs in the TCW system and heat exchangers in the turbine building in a

' Amendment 1s 9.2 10

ABM Standard Plant 23ssixxn nry n 9.2.17 laterfaces any failure in the system,.locluding any that cause flooding, shall not result in the failure 9.2.17.1 Ultimate Heat Sink Capability of any safety.related structure, system or component.

Interface requirements pertaining to ultimate heat sink capability ate delineated in Subsection 9.2.17.3 Potable and Sanitary Water System 9.2.5 as follows:

Tbc potable and sanitary water system shall Subsection Illlt be designed with no interconnections with systems having the potential for.containing 9.2.5.1 Safety Design Bases radioactive materials. Protection shall be provided through the use of air gaps, where 9.2.5.2 Power Generation Design Bases . necessary. (See Subsection 9.2.4).

E 9.2.5.6 Enluation of UHS Performance 9.2.17.4 Reactor Senice Water System L9.2.5.7 Safery Evaluation The RSW pumps are described in Table 9.2 13. Tbc applicant shall provide the 9.2.5.8 Conformance to Regulatory . following additional information which is site .

Guide 1.27 dependent: (See Subsection 9.2.15.2 and 9.2.15.3 ) .

9.2.5.9 Instrumentation and Alarms (1) temperature increase and pressure drop

.9.2.5.10 Tests and inspections across the heat exchangers; 9.2.17.2 Makeup Water System Capabilltf -(2) the required and available net positive 50 suction head for the RSW pumps at pump y The raw water treatraent and preparation of suction locations considering anticipated =

,. the demineralized water is sent to the makeup low water levels; -

L water system (purified) described in Subsection 9.2.10. . (3) the location of the RSW pump house; -

If any spray pond piping is made fromh(4) the design features to assure that the i fiberglass reinforced thermosetting resin, the I requirements in Subsection 9.2.15,1,1(3) are applicant shall provide information to show that met; an'd all applicable requirements of Regulatory Guide

, 1.72 are -met. (5) an analysis of a pipeline break and a single l k -

active component failure shall show that The demineralized water preparation system flooding shall not affect the main control shall consis~t of at least two divisions capable room or more than one division of the RSW L L of producin's at least 200 gpm of demineralized ' syst e m.

i water each; Storage ci demineralized water shall bc ai least 200,000 galloas. If addi 1onal 9.2.17.5 Turbine Service Water System demineralized water is needed during peak usage -

periods, rented portable d:mineralizers shall be The applicant shall demonstrate.that all

- used as required.

safety related components, systems, and structures are protected from flooding in the

- The makeup water preparation system shall be event of a pipeline break in the TSW system.

, located in a building which does not contain any (See Subsection 9.2.16.3)

-E safety related structures, systems or L & components. If the system is not available, demineralized water can be obtained from mobile equipmenti The system shall be designed so that

~~Ameedment 18 9.2 13

.- .,