ML20036A084

Forwards SSAR Markup & New App 1B, Comparison of Us ABWR & K-6/7 Difference, Addressing Draft Final SER Confirmatory Item 1.2-1

ML20036A084
Person / Time
Site:	05200001
Issue date:	04/30/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9305100022
Download: ML20036A084 (13)

Text

>

).

GE Nuclear Energy cemet Deanc compan 175 Ctriner henne. Srr Jose. CA 95125 April 30,1993 '

Docket No. STN 52-001 Chet Poslusny, 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 1.2-1

Dear Chet:

Enclosed is a SSAR markup and a new Appendix 1B," Comparison of U.S. ABWR and K-6/7 Difference" addressing DFSER Confirmatory Item 1.2-1.

Please provide a copy of this transmittal to Jerry Wilson.

Sincerely, Ef ek Fox Advanced Reactor Programs l-cc: ' Alan Beard (GE)

Norman Fletcher (DOE) l 060049 06 JIW125 I

9305100022 930430-

~I D

PDR-ADOCK.05200001 A

PDR '

CI

1. 7.

n AMM ussiooxc nry c Standard Plant 1.2 GENERAL PLANT DESCRIPTION (8) Reactor controls, including alarms, are A

arranged to allow the operator to rapidly 1.2.1 Principal Design Criteria assess the condition of the reactor system and locate system malfunctions.

The principal design criteria are presented in two ways. First, they are classified as either a (9) Interlocks or other automatic equipment are

,/

power generation function or a safety function.

provided as backup to procedural control to

[

Second, they are grouped according to system.

avoid conditions requiring the functioning l

Although the distinctions between power genera-of nuclear safety systems or engineered tion or safety functions are not always clear cut safety features.

and are sometimes overlapping, the functional f

classification facilitates safety analyses, while (10) The station is designed for routine the grouping by system facilitates the under-continuous operation whereby steam standing of both the system function and design.

activation products, fission products, corrosion products, and coolant dis-1.2.1.1 General Design Criteria sociation products are processed to remain within acceptable limits.

1.2.1.1.1 Power Generation Design Criteria 1.2.1.1.2 Safety Design Criteria (1) The plant is designed to produce steam for l

direct use in turbine-generator unit.

(1) The station design conforms to applicable codes and standards as described in Sub-l (2) Heat removal systems are provided with section 1.8.2.

sufficient capacity and operational adequacy to remove heat generated in the reactor (2) The station is designed, fabricated, f

core for the full range of normal crected, and operated in such a way that operational conditions and abnormal the release of radioactive material to the

(

operational transients.

environment does not exceed the limits and I

guideline values of applicable government (3) Backup heat removal systems are provided to regulations pertaining to the release of remove decay heat generated in the core radioactive materials for normal opera-under circumstances wherein the normal tions, for abnormal transients; and for operational heat removal systems become accidents.

inoperative. The capacity of such systems is adequate to prevent fuel cladding damage.

(3) The reactor core is designed so its nuclear characteristics do not contribute to a (4) The fuel cladding in conjunction with other divergent power transient.

plant systems is designed to retain integrity so that the consequences of any (4) The reactor is designed so there is no failures are within acceptable limits tendency for divergent oscillation of any j

throughout the range of normal operational operating characteristic considering the conditions and abnormal operational interaction of the reactor with other transients for the design life of the fuel.

appropriate plant systems.

(5) Control equipment is provided to allow the (5) The design provides means by which plant reactor to respond automatically to load operators are alerted when limits on the changes and abnormal operational transients.

release of radioactive material are i

approached.

(6) Reactor power level is manually control-lable.

(6) Sufficient indications are provided to allow determination that the reactor is (7) Control of the reactor is possible from a operating within the envelope of conditions single location.

considered safe by plant analysis.

i s s %d ow ho-+bc Med*^M ag dap %

'The

v. s. Aswe a m n l

4 cb i s e v uev M-)

e q d e yta'-

[

• k JYb k-.G h ),

Amendment 7 ] %.ov%c, N a olc u Po # G *ue

{

p

, m h rA d e cho c P o w ev-Co M "

e 4

g f

K' f "'ON gh e d[a g,

. S. A S% ekR 8 Wp cm Aeeod x i s.

m

~-

1 t

APPENDIX 1B COMPARSION OF U.S. ABWR AND K-6/7 D11+ERENCES l

t l

t l

i.

a

l COMPARISON OF U.S. ABWR AND K-6/7 i

DIFFERENCES U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 1.

General Design 1.1 Single unit plant Dual unit Some facilities shared between dual units and other site units 1.2 Seismic 0.3g SSE all Seismic sF.e specific ALWR soils envelope 1.3 60 year plant life 40 year ALWR 1.4 Ultimate heat sink Maximum temperature U.S. design supports generic site i

maximum temperature of 85*Fassumed of 95'F assumed envelope 1.5 U.S. Codes and MITI Codes and NRC Standards Stanciards 1.6 ABWR Product K-6/7 Product None i

Structure structure 1.7 Grid frequency 60 Hz Grid Frequency 50 Hz None 1.8 Radwaste system Standard None design customized for Hitachi/Toshiba design U.S.

2.

Plot Plan 2.1 Turbine building and Axis perpendicular to turbine axis in-line with ALWR/ Japanese choose to address reactor building reactor building turbine missile issue entirely from a structuralperspective to have a more compact site plot plan 2.2 Control building located Located between dual between reactor Cost minimization effort for single reactor buildings unit plant building and turbine building a.

Control room HVAC Single air intake '

Dualintake design results in less includes dualwidely separated operator dose to operator in U.S. control selectable air room exposure analysis i

intakes b.

RCW HX's located in Dedicated HX building U.S. layout reconfigured to reflect basement of control building different site plot plan c.

RIP MG sets located RIP MG sets located in Individual preference in control building radwaste building :

I COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued)

U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 2.3 Radwaste building Shared facilities on Japanese emphasis on efficiency designed for a single multi-unit site. K-6/7 and compact site layout unit (ABWR) share facilities with K-5 (BWR-5)

I 2.4 Technical support Not required in Japan NRC requirement that TSC be within center located in service 2 minute walk of the main control building room 2.5 Condensate storage Storage poollocated in CST cannot be housed in non-tank (CST)in yard radwaste building seismic Category I structure i

2.6 Dual unit common Common switchgear Single versus Dual unit plant design I

switchgear deleted used i

3.

Power Cycle System 3.1 Power cycle system Japanese emphasis is ALWR design meets U.S. utility on maximum heat rate I

preference, with and thermal efficiency emphasis on simplicity.

a.

FW pumps driven Steam driven pumps ALWR I

by variable speed motor I

b. Condensate has Condensate pumps ALWR 4x33-1/3% pumps; plus booster pumps; no condensate 3x50% pumps at each

]

booster pumps stage

c. Low pressure FW Pumped forward ALWR, high pressure heater drains heater drains pumped forward in both designs cascaded back to condenser d.

Moisture 2 stage reheat ALWR separator / reheaters l

have 1 stage reheat l

l

e. Condenser is Single pressure ALWR l

multiple pressure f.

Condenser tubing Titanium ALWR, requirements allow use of cooling water materials suitable for actual site dependent cooling water conditions i

j. l

t COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued) l U.S. ABWR K-6/7 REQUIREMENT / COMMENTS

g. Turbine gland Dedicated system ALWR sealing steam supplies clean steam extracted from main steam h.

Steam jet air 1100% train plus 1 ALWR ejectors has startup train (driven by 2x100% trains auxiliary steam)

i. Condenser heat Sea water ALWR sink site dependent j, TBCW system has 3x50% pumps and HXs ALWR 2x100% pumps and HXs k.

Condensate Single stage ALWR, meets water quality polishing is two exposure and radwaste burial stage volume goals 3.2 Offgas system is GE H/T design based on Individual preference r

N68 design earlier GE N62 design 3.3 Hydrogen water Not adopted Desirability still under study in Japan chemistry integralwith design 3.4 Provisions for Zinc No Zinc addition Zinc addition is optional addition to Feedwater I

i l l

+

COMPARISON OF U.S. ABWR AND K-6/7 9

DIFFERENCES (Continued)

U.S. ABWR K-6/7 R EQQ)R E M E NT/COM M E NTS 4.

Electrical Design 4.1 Offsite/onsite AC power 7 unit site with multiple U.S. design reflects ALWR sources are the low offsite AC power sources requirements (both designs voltage generator include normal compliment of output breaker plus one emergency diesel generators) independent offsite source plus non-safety onsite gas turbine 4.2 Onsite power No generator output AC network interface designed distribution network has breaker or gas turbine; for respective site conditions generator output startup transformers used (switching logic also modified breaker and feed from to provide feed in accordingly) gas turbine added; conventional way l

startup transformers deleted 4.3 Isolation of 1E from non-Circuit Breakers are used Circuit breakers are not accepted 1E loads on low voltage between 1E supplies and by NRC for electricalisolation of ac/dc circuits non-1E loads 1E and non-1E loads i

4.4 DG fuel storage is 2x200% divisionally cross-K-6/7 design emphasizes 3x100% divisionally tied tanks (per reactor compact site plot plan; cross ties separated tanks located unit) located above allowed by less rigorous divisional underground ground separation requirements 4.5 DG start capability Normalcapability (AC ALWR incorporates manual (no power required)

AC) start capability 4.6 DG fire suppression is CO2 system ALWR foam system 4.7 No PVC electrical Use of PVC OK ALWR i

insulation allowed l

4.8 Non-safety chillers and Gas turbine is not required ALWR coolers connectable to j

on-site gas turbine 4.9 Separation of 1E Separation of 1E divisions NRC divisions is done with 3 may utilize distance l

hour fire barriers where without intervening j

practically achievable.

barriers.

l t

J l

l- !

l

COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued)

U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 5.

Primary Containment 5.1 Severe accident design Not part of design Subject of severe accident -

features mitigation is still under study in Japan

a. Containment Not part of design Passive venting of wetwell overpressure airspace through two rupture protection discs in series in hardened path; i

containment integrity recoverable by closing normally open AOVs

b. Strengthened Not part of design Drywell head thickness increased drywell head from 1" to 1.25"; Pressure capability increased to near ultimate strength of balance of the containment structure l

c. Limestone concrete Not part of design Reduces non-condensable gas prohibited in lower generation from potential core-drywellarea concrete interaction

d. Lower drywell Not part of design Utilizes fusible plugs on pipes flooder connecting suppression pool to lower drywell

e. AC independent Not part of design Fire water system cross-tied into i

water addition RHR with manually operated capability valves f.

Onsite combustion Not part of desgn ALWR turbine generator 5.2 Wetweil/Drywell vacuum Vacuum breakers are air Testability removed based on breakersare not testable testable check valves PRA insight that additional failures are introduced.

5.3 SRV discharge piping in Specified as MITI Class 4 NRC wetwell region specified so no ISI required as ASME Class 2 (MITI Class 3 equivalent) therefore, ISI is required 5.4 RPV metal temperature K-6/7 to have extra ALWR/ Extra monitoring capability sensor reduction monitoring capability not needed for follow-on plants l

l ;

i

COMPARISON OF U.S. ABWR AND K-6/7 e

DIFFERENCES (Continued)

U.S. ABWR K 6/7 REQUIREMENT / COMMENTS 5.5 Bottom head drain line Manualisolation valve on NRC concem, limits break path to has AOV to provide bottom head drain line isolation of drainage above the top of active fuel path in the event of an unisolable break in the CUW system outside containment.

6.

Secondary Containment 6.1 Redundant flammability Portable skids -one skid For K-6/7 redundancy is provided control system normally installed in by portability of skid in other unit's i

(hydrogen recombiners) reactor building of each reactor building i

permanently installed unit 6.2 SGTS has 4000 scfm 1200 scfm capacity Less prescriptive requirements capacity with auto negative pressure for SGTS sizingin Japant i

controlcapability.

increased capacity of U.S. system Redundant trains necessitates capability to control separated by 3-hour fire negative pressure to prevent

barriers, excessive differential pressure on reactor building 6.3 Steam and FWlines Seismic out to turbine; no Seismically qualified turbine classified non-seismic seismic interface restraint outboard of seismic building is standard Japanese interface restraint practice

a. Leak-before-break Conventionally analyzed and Leak-before-break methodology l

methodology used to supported eliminate pipe whip still under study in Japan restraints 6.4 HPCF pumps discharge None check valve added NRC/High pressure isolation 6.5 ECCS injection valve None haadwheel and improved sabotage resistance improved position monitoring added i

i 6.6 CRD pump motor 20%

U.S. Codes and Standards overspeed 25%

6-

COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued)

U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 6.7 Wallsof upper 2 levels Tomado not a design NRC and the roof of the requirement in Japan reactor building have been increased for tornado missile protection.

i i

7.

Control Room 7.1 ARBM logic enforces Logic does not enforce ARBM enforcement of OLMCPR OLMCPR, even in OLMCPR in manual in all modes eliminates RWE as Manual mode, to mode; RWE transient credible transient in U.S.; thus, prevent Rod Withdrawal analyzed as acceptable analysis is not required Error transient 7.2 Automatic boron Manual NRC/ Recirculation run back and' injection ARl/FMCRD run in initiated from I

scram 7.3 Automatic suppression Manual ALWR, No operator action pool cooling for 72 required for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following hours transient 7.4 Automatic ADS after Manual NRC additional 8 minutes i

without high drywell pressure i

a. ADS includes Inhibit switch not provided ADS inhibit switch required in manualinhibit U.S. to help mitigate ATWS switch on main control panel

b. Monitor solenoid Monitor solenoid improved sabotage resistance continuity for ADS continuity not provided SRVs 7.5 RPS seismic trip is not Trip on high ground Seismic scramtripis standard an RPS input acceleration Japan practice l

f 7.6 No RPS trip on TCV RPS trip on TCV solenoid Standard Japan practice solenoid position position switch input t

7.7 RPV waterlevel Reference zero at TAF for in Japan,it was decided that least i

instrumentation fuel zone range only; all confusing solution is to retain reference zero at TAF others use bottom of past BWR practice (U.S.

for allinstruments separator skirt for designed dictated by TMI Action l

reference zero Plan item) l I

y

s COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued) l U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 7.8 Safety related RHR HXs Non 1E NRC outlet temperature monitor 7.9 Keylock switch on RHR No keylock ALWR discharge valve to radwaste 7.10 RPS trip on high Manual ALWR, No operator action suppression pool required for 30 minutes following temperature.

a transient.

f 7.11 Auto power reduction Manual ALWR, No operator action onloss of feedwater required for 30 minutes following heating a transient.

l

(

7.12 Non-Class 1E Non-safety uninteruptible NRC i

uninteruptible power power supplies powered supplies provided.

by 1E power source.

7.13 RPS trip provided for Manual ALWR concem arising from core core power oscillations.

stability incidents l

e F

a 8.

Water /Alr 8.1 RCW has 3x50% vertical 2x100% horizontal HXs

. Differing configurations reflective HXs (per division)

(per division) of locational space constraints

a. Corrosion Not included ALWR i

monitoring i

subsystem included 8.2 Essential HVAC has Division C uses forced air Division C has less heat load and cooling coils in all 3 only for reactor building cooling coils not needed at actual divisions; division C loads and does not serve conditions of K-site; U.S. design serves control room control room must support generic site envelope a.

HVAC essential Divisions A and B only Generic site envelope cooling water l

divisions A, B and C i

b. Drain collection to Storm drains ALWR radwaste or recycle to RCW l w.--

I COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued)

U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 8.3 HVAC normalcooling Smaller size U.S. system has larger capacity to water system has accommodate generic site increased size envelope -

8.4 RCIC room dedicated Shared sump with RHR 'A' Dedicated RCIC sump provides sump considerable PRA benefit from flooding evaluation 8.5 Instrument air system Auto-transfer to back up There is a cross-tie between K-5, has manual cross-tie nitrogen supply mode K-6 and K-7 back-up to nitrogen supply 8.6 Breathing airis Supplied by service air ALWR dedicated system T

8.7 Service air filters and No filters and dryers ALWR dryers added i

8.8 CRD purge water not CRD purge water is Japanese practice not used by heated.

heated.

US plants.

8.9 Reactor Service Water Reactor Service Water US ABWR design utilizes a major components major components separate ultimate heat sink.

relocated to Ultimate located in heat exchanger l

heat sink intake building.

structure and basement of control building.

9.

Fire Protection v

9.1 Physicalfire barriers with Some interdivisional Japanese practice allows some l

3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ratings used at all equipment located in areas that contain safety related divisional boundaries common areas equipment (including of ditferent outside containment designated as "non-fire divisions) to be subject to less high energy piping zone". Penetrations do strict fire protection requirements penetrations also not require 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ratings if supported by analysis showing I

require 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire probability or size of fire to be low ratings (or appropriate justification otherwise) 9.2 U.S. design has No such mode is required U.S. requires capability to dedicated smoke exhaust smoke and prevent I

removal mode migration to other divisions j

consisting of dampers and logic 1

.g.

l

COMPARISON OF U.S. ABWR AND K-6/7 DIFFERENCES (Continued)

~

U.S. ABWR K-6/7 REQUIREMENT / COMMENTS 9.3 Four SRVs controllable 3 SRVs controllable per Addition of 4th SRV at RSP i

at Remote Shutdown onginal design; U.S.

improves results of fire PRA by i

Panel (RSP) design change still under factor of 10 study 10.

Radiation l

v 10.1 Containment leakage 0.4%/ day assumed Japanese data shows 0.5%/ day assumed in consistently less leakage than in dose analysis U.S.; U.S. assumption reflects utility desire to retain margin for test 10.2 MStV leakage 140 scfh 45 scfh total assumed Historic Japanese data shows totalfor alllines consistently less leakage than in assumed in dose U.S.: US. assumption reflects analysis

- utility desire to retain margin for test 10.3 Reconfigure ARM and Site specific Accommodate plant arrangement PRM systems to U.S.

and processes design a

F l

1 1

J 1

-