ML20045E841

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Forwards ABWR Tier 1/ITAAC Matl That Was Not Included in Submittals Dtd 930426,0521,0604 & 18.Matl Preliminary in That Matl Has Not Been Fully Verified Using GE Procedures Governing Compliance W/Qa Requirements
ML20045E841
Person / Time
Site: 05200001
Issue date: 06/29/1993
From: Recasha Mitchell
GENERAL ELECTRIC CO.
To: Borchardt R
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
MFN-102-93, NUDOCS 9307060125
Download: ML20045E841 (63)
v • d • e


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t June 29, 1993 MFN 102-93 i

Docket No. STN 52-001 i

Document Control Desk U. S. Nuclear Regulatory Commission Washington, D. C. 20555 l

Attention:

Mr. Richard W. Borchardt, Acting Director Standardization Project Directorate

Subject:

Submittal Supporting Accelerated ABWR Design Certification Review -

Submittal of Revised ABWR Tier 1/ITAAC Material

Dear Mr. Borchardt:

Enclosed are thirty-four (34) copies of ABWR Tier 1/ITAAC material that was not included in the earlier submittals dated 4/26/93, 5/21/93, 6/4/93, and 6/18/93. The attached table lists the material covered by this i

submittal. With this submittal, CE now has provided the staff with a complete set of revised ABWR Tier 1 entries.

j The updated Tier.1 material in this and the other submittals noted above have been prepared in parallel with closure of open items on the SSAR and may not reflect very recent GE/NRC agreements on SSAR open issues.

However, we do not view this parallel SAR resolution / Tier 1 submittal process as causing major difficulties and plan to update the. Tier 1 i

systems material in the next few weeks to incorporate resolved SSAR design issues.

The scope of potential changes to the submitted Tier 1 naterial was identified in the letter J. F. Quirk to D. M. Crutchfield dated June i

21, 1993; subject, " Assessment of ABWR SSAR Modifications on ABWR Tier 1 Submittals." We do not anticipate any major perturbations and do not i

believe this change process should impede staff review of the ABWR Tier 1 document.

l The material in this and the other phased submittals (4/26/93, 5/21/93, j

6/4/93 and 6/18/93) is preliminary in that it has not been fully verified using CE procedures governing compliance with quality assurance requirements for engineering documentation. This activity will involve 09onnt 0g-9307060125 930629 PDR ADDCK 05200001 A

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Document Control Desk June 29, 1993 U. S. Nuclear Regulatory Commission MFN 102-93 Washington, D. C.

Docket No. STN 52-001 Attn:

Mr. Richard W. Borchardt verifying Tier 1 material against the final version of the SSAR and is scheduled to be completed in August 1993. Our current plan is to resubmit the complete set of verified ABWR Tier 1 material at the end of August 1993.

As always, GE personnel will be happy to provide any support the NRC staff review teams feel they need to complete their review of the attached material.

Sincerely, h.(-

1 R. C. Mitchell, Acting Manager 1

Safety and Licensing (408) 925-6948 enclosures cc:

T. A. Boyce (NRC)

N. D. Fletcher (DOE)

C. Poslusny (NRC)

A. J. James (GE)

R. Louison (GE)

N. D. Hackford (GE)

J. F. Quirk (GE)

J. N. Fox (GE)

J. A. Beard (GE-Rockville) i i

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The ABWR design provides radiation protection features keep exposures for -

l both plant personnel and the general public below allowable limits.

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maintains radiation exposures to plant personnel as low as reasonably achievable. Shielded cubicles, labyrinth access and provisions for temporary shielding are used to reduce exposure. Under accident conditions, plant shielding designs permit operators to perform required safety functions in sital areas of the plant. A vital area is an area which will or may require ocmpancy to j

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low as is reasonably achievable.

y Plant ventilatica systems insure that concentrations of airborn a mdionuclides are maintained at levels consistent with personnel access requirements. In j

addition, airborne radioactivity monitoring is provided for those normally occupied areas of the plant in which there exists a significant potential for airborne contaminadon. The equipment for measurement of airborne particulate concentrations is not in the scope of this design and is specified as r -

interface in Section 4.8.

Inspections, Tests, Analyses and Acceptance Criteria Tables 3.2a and 3.2b provide a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken j

for the ABWR plant shielding, ventilation and airborne monitoring equipment.

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0 0 Table 3.2a Plant Shielding Design e inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 1. The plant design shall provide radiation 1. An analysis of the expected radiation levels 1. Maximum expected radiation dose rates in shielding for rooms, corridors and in each plant area will be performed to each plant area (deep dose equivalent operating areas commensurate with their verify the adequacy of the shielding measured at 30 cm from the source of the occupancy requirements to maintain design. This analysis shall consider the radiation, not contact dose rates) are no - radiation doses to plant personnel as low following: greater than the dose rates specified for as reasonably achievable. the following zones, based on the access a. Confirmatory calculations shall requirements of that area for plant consider significant radiation sources operation and maintenance. (greater than 5% contribution) for an area. Radiation source strength in plant Z_qDn Dose Rate acc_qsa systems and components will be (mrem /hr) Reauirements determined based upon an assumed source term of 100,000 pCurie/second A 50.6 Uncontrolled, offgas release rate (after 30 minutes unlimited access. decay), a 300 pCurie/ gram-steam N-16 4 source term at the vessel exit nozzle, B <1 Controlled, unlimited access. Y and a core inventory commensurate with a 4005 MWt equilibrium core at C <5 Controlled, limited 51.6 kwatt/ liter. Source terms shall be access. 20 hr./ week. adjusted for radiological decay and buildup of activated corrosion and D <25 Controlled, limited wear products. access. 4 hr/ week.

b. Commonly accepted shielding codes, E

<100 Controlled, limited using nuclear properties derived from access.' 1 hr/ week. well known references (such as Vitamin F 2100 Restricted, infrequent C and ANSI /ANS-6.4) shall be used to a ess. Authorizat. ion model and evaluate plant radiation required. environments.

1) For non-complex geometries, point Plant layout such that access to higher kernel shielding codes (such as QAD or zones (areas with higher dose rates) is GGG) shall be used.

from lower zoned areas. Corridors and

2) For complex geometries, more normal traffic areas are Zone C or less.

sophisticated two or three dimensional Control rooms are Zone B or less. transport codes (such as DORT or Radiation zones for the Reactor Building TORT) shall be used. and Control Building are indicated in Figures 3.2a through 3.2u. b

p ) s V 3 Table 3.2a Plant Shielding Design (Continued) Inspections, Tests, Analyses and Acceptance Criteria w Design Commitment inspections, Tests, Analyses Acceptance Criteria c. A safety factor shall be applied based upon benchmark comparisons. 2. The plant design shall provide shielded 2. Using the methods identified in (1) above, 2. Shielding design of a room including any cubicles, labyrinth access, and space for radiation levels present in rooms shall be temporary shielding is such that radiation temporary shielding to reduce radiation evaluated for the contribution from from adjacent rooms shall contribute no exposure from adjacent rooms adjacent rooms. more than a small fraction (10% or less) of the dose rate or less than 0.06 mrem /hr whichever is larger, in the room. For this purpose the drywell shall be considered a room. 3. The plant radiation shielding design shall 3. An analysis of the expected high radiation 3. Under accident conditions, radiation permit plant personnel to perform required levels in each area which will or may shielding design allows access to safety functions in vital areas of the plant require occupancy to permit plant occupancy and egress from areas required (including access and egress of these personnel to aid in the mitigation of or to maintain post accident safety functions 4 areas) under accident conditions. recovery from an accident (vital area) shall such that individual personal radiation ? be performed to verify the adequacy of the doses do not exceed 5 rem to the whole plant shielding design. This analysis shall body, or its equivalent, for the duration of use calculational methods consistent with the accident (based on the required (1.b) above and a radiation source term frequency of access to each vital area). For (adjusted for radioactive decay) based on areas requiring continuous occupancy the following: (such as the main control room, technical support center, and emergency operations a. Liquid containing systems: 100% of the support center), design dose rates shall not core equilibrium noble gas inventory, exceed 15 mrem /hr (averaged over 30 50% of the core equilibrium halogen days). inventory and 1% of all others are assumed to be mixed in the reactor coolant and recirculation liquids recirculated by the residual heat removal system (RHR), the high w bJ .~

O O g Table 3.2a Plant Shielding Design (Continued) inspections, Tests, Analyses and Acceptance Criteria w Design Commitment Inspections, Tests, Analyses Acceptance Criteria

3. (continued) pressure core flooder (HPCF), and the reactor core isolation cooling (RCIC) systems.

b. Gas containing systems:100% of the core equilibrium noble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere. For vapor containing systems (such as the main steam lines) these core inventory fractions are assumed to be contained in the reactor coolant vapor space. ,wY 4. The plant design shall provide radiation 4. Using the methods identified in (1) above, 4. As a result of normal operations, the shielding to maintain radiation dose to the the radiation dose to the maximally radiation dose from direct and scattered general public outside of the controlled exposed member of the general public radiation shine to the maximally exposed area as low as is reasonably achievable, outside of the controlled area from direct member of the public outside of the. and sca'ttered radiation shine shall be -controlled area is equal to or less than 2.5 determined. mrem / year. NU

Table 3.2b Ventilation and Airborne Monitoring is" Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections Tests Analyses Acceptance Criteria 1. Plant design shall provide for containment 1. Expected concentrations of airborne 1. Calculation of radioactive airborne t of airborne radioactive materials and the radioactive material shall be calculated by concentration shall demonstrate that: ventilation system will ensure that radionuclide for normal plant operations concentrations of airborne radionuclides and anticipated operational occurrences a. For normally occupied rooms and are maintained at levels consistent with for each equipment cubicle, corridor, and areas of the plant (i.e. those areas personnel access needs. operating area requiring personnel access. requiring routine access to operate and Calculations shall consider: maintain the plant) equilibrium concentrations of airborne a. Design ventilation flow rates for each radionuclides will be a small fraction

area, (10% or less) of the occupational concentration limits listed in 10 CFR 20
b. Typical leakage characteristics for Appendix B.

equipment located in each area, and b. For rooms that require infrequent c. A radiation source term in each fluid access (such as for non-routine system based upon an assumed offgas equipment maintenance), the g, 9 rate of 100,000 Curie /second (30 ventilation system shall be capable of minute decay) appropriately adjusted reducing radioactive airborne for radiological decay and buildup of concentrations to (and maintaining activated corrosion and wear products, them at) the occupational concentration limits listed in 10CFR20 Appendix B during the periods that occupancy is required. t c. For rooms that seldom require access i (such as the backwash receiving tank room), plant design shall provide containment and ventilation to reduce airborne contamination spread to other areas of lower contamination. W N r .w + e .4 -c, -ww. ..n. v-3e ,mm-w-- .--.-a ~-m>'- we-nv.- -rw% -.- - --- - -- --- = --m---m

O O O Table 3.2b Ventilation and Airborne Monitoring (Continued) inspections, Tests, Analyses and Acceptance Criteria w Design Commitment inspections, Tests, Analyses Acceptance Criteria 2. Airborne radioactivity monitoring shall be 2. An analysis shall be performed to identify 2. Airborne radioactivity monitoring system provided for those normally occupied the plant areas that iaquire airborne shall be installed as defined in this certified areas of the plant in which there exists a radioactivity monitoring. design commitment. significant potential for airborne contamination (greater than 0.1 per year) The airborne radioactivity system shall: a. Have the capability of detecting the time integrated concentrations of the most limiting internal dose particulate and iodine radionuclides in each area equivalent to the occupational concentration limits in 10CFR20, Appendix B for 10 hours. b. Provide a calibrated response, ,y representative of the concentrations within the area (i.e. air sampling monitors in ventilation exhaust streams shall collect an isokinetic sample), c. Provide local audible alarms (visual alarms in high noise areas) with variable alarm set points, and readout / annunciation capability. wN

Table 12 Generic Safety issues Verifying SSAR Entrv Parameter Value ITAAC 19B.2-1 Quality and Reliability Assurance Quality System Hequirements have been Identified for Ea':h System Covered by Individual Sys. Entries 19 B.2-2 A-1: Water Hammer Steam Supply System Designed to 3.3 Accommodate Steam Hammer MSL Designed for Dynamic Loadings Due 3 Fast Closing of the Turbine Stop Valves 3.3 Feedwater System Variable Speed Pumps none Slow Acting Low Flow Control Valve none HVAC System Elevated Surge Tank none High Point Vents none COL Req. Procedures to Keep System Filled Slow Acting System Valves none i RSW System Low Point Drains none High Point Vents none COL Req. Procedures to Keep System Filled O T

Table 12 Generic Safety issues (Cont.) verifying SSAR Entry Parameter Value ITAAC t 19B.2-2 A-1: Water Hammer (Cont.) TSW System Low Point Drains none r High Point Vents none Procedures to Keep System Filled COL Req. RCIC System Low Point Drains none High Point Vents none MUWC to Keep System Filled 2.4.4 Low Water Fill Alarm none Slow Acting System Valves none HPCF System Low Point Drains none High Point Vents none 2.4.2 MUWC to Keep System Filled Low Water Fill Alarm none Slow Acting System Valves none 4 O 2 f r

Table 12 _ h-), Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2-2 A-1: Water Hammer (Cont.) l RHR System Low Point Drains none High Point Vents none Jockey Pump to Keep System Filled 2.4.1 Low Water Fill Alarm none Slow Acting System Valves none 198.2-3 A-7: MARK I Long-Term Program l Vacuum Breakers 2.14.1 Swing Check Type Valves Open Passively on Negative 2.14.1 Differential Pressure Require No External Power to Actuate 2.14.1 installed Horizontally Through 2.14.1 Pedestal Wall Solid Catwalk Area Structure in Wetwell Designed as Shielding Against Pool Swells --- none i 19B.2-4 A-8: MARK ll Containment Pool Dynamic Loads Long-Term Program (Refer to response to 198.2-3) 3 a

Table 12 ~ Generic Safety issues (Cont.) Verifying SSAR Entrv Parameter Value ITAAC 19B.2-5 A-9: ATWS Afternate Rod insertion Feature Diverse and Independent From RPS 2.2.8 Electric insertion of FMCRD Feature Diverse and independent From RPS 2.2.8 Recirculation Pump Trip on ATWS Signal 2.2.8 Automatic initiation of SLC on ATWS Signal --- 3.4 198.2-6 A-10: BWR Feedwater Nozzle Cracking The Feedwater Nozzle has an Inner Thermal Sleeve Welded to the Nozzle Safe End. none A Secondary Thermal Sleeve is Placed i Concentricailly Between inner Thermal Sleeve and Nozzle Bore. none 19B.2-7 A-13: Snubber Operability Assurance Snubber Design Considers Load Cycles and Travel During Normal Plant Operation none Thermal Growth Rate of System Does Not Exceed Lock-up Velocity none Snubber Mechanical Properties in Structural Analysis none Design Specification and Testing for Engineered, Large Bore Snubbers includes Activation Level, Release Rate, Spring Rate, Dead Band and Orag. none O 4

Table 12 h Generic Safety issues (Cont.) verifying i SSAR Entry Parameter Value ITAAC 19B.2-8 A-24: Qualification of Class 1E Safety Related Equipment All Class 1E Electrical Equipment is Environmentally, Dynamically and ~ Seismically Qualified Refer to 1.2(3) 19B.2.-9 A-25: Non-Safety Loads on Class 1E Power Sources Non-Class 1E Loads not Connected to Class 1E Loads Except FMCRD Loads 2.12.1 Class 1E Load Breakers in Division i Between Class 1E Power and 2.12.1 Non-Class 1E FMCRD Loads O Zone Selective Interlocking Between \\/ Class 1E Load Breakers and i Upstream Class 1E Bus Feed Breakers none i 19B.2-10 A-31: Residual Heat Removal (RHR) Shutdown Requirements RHR System Composed of 3 Electrically And Mechanically Independent Divisions 2.4.1 Shutdown Cooling Can Be Manually 2.4.1 i Initiated from the Control Room RHR System Can Bring the Reactor to Cold Shutdown Within 36 Hours With Any 2 of the 3 Divisions none RHR System Can Be powered from Both Offsite and Standby Emergency Electrical Power 2.4.1 2.12.1 o O 5

i Table 12 O Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2-11 A-35: Adequacy of Offsite Power Systems Bus Voltage Less than 90% of Rated for a Given Time Delay Alarmed in Control Room none Protective Relay Timer Started to Protect Class 1E Loads none Timer Resets if Bus Voltage Recovers none Trips Feeder Breaker if Times Out none Equipment Qualified for Operation with 2.12.1 Voltage up to 10% Less than Normal 19B.2.12 A-36: Control of Heavy Loads Near Spent Fuel Equipment Handling Components Meet 2.15.3 Single Failure Criteria Redundant Safety Interlocks and Limit Switches Prevent Heavy Loads Over 2.15.3 Spent Fuel Design Safety Factor for Lifting Strongback 10 none i Transportation Routing Study, Operating i instructions, Maintenance Manual, i inspection and Test Program for Heavy COL Req. Load Handing System b O 5 6

Table 12 O o e e e ric s <etv i e e < c e et.) Verifying SSAR Entry Parameter Value ITAAC 19B.2.13 A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits Each S/RV Discharge Pipe Fitted with an X-Ouencher 2.14.1 2.1.2 (Design Des. Only) No Submergered Structures Located Within the Sphere Circumscribed By Quencher Arm none Load Definition Methodology Consistent With Mark 11 and ll1 Containment Designs none 19B.2.14 A-40: Seismic Design Criteria Short-Term Program Seismic Analysis Methods and Acceptance Criteria Conform to Appropriate Sections of Rev. 2 of SRPs none 198.2-15 A-42: Pipe Cracks in Boiling Water Reactors (Refer to response to 19B.2-41) 19B.2-16 A-44: Station Blackout Sources of Electrical Power No. of Standby Turbine Generators 1 2.12.11 No. of Emergency Diesel Generators 3 2.12.13 No. of Offsite Power Sources 2 none (Interface Req. 2.12.1) i O 7

Table 12 O-o e eric s retv i u e s ( c o at.) Verifying SSAR Entrv Parameter Value ITAAC 19B.2-17 A-47: Safety implications of Control Systems Feedwater Controller Trip Feedpumps on High Water Level 2.2.3 Fault Tolerant Through Redundant Micro-processors and Self Diagnostics -- 2.2.3 NRC-approved Technical Specifications COL Req to Verify Operability of Overfill Protection 19B.2-18 A-48: Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment Containment inerted During Noncal 2.14.6 Operation Permanently Installed Hydrogen 2.14.8 Recombiners 19B.2-19 B-10: Behavior of BWR Mark 111 Containments l Condensation Oscillation and Chugging Loads Based on ABWR Horizontal Vent none Tests t Load Definition on Submerged Structures Consistent with Methodology for Mark ll and 111 Containment Designs none 19B.2-20 B-17: Criteria for Safety-Related Operator Actions No Operator Action Required for 30 Minutes for Design Basis Accidents none RHR Heat Exchanger in LPCI Injection Loop 2.4.1 CRTs in the Control Room for Monitoring and Alarming none 8

Table 12 fs() Generic Safety issues (Cont.) Verifying SSAR Entrv Parameter Value ITAAC 19B.2-21 B-36: Develop Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Absorption Units for Engineered Safety Features Systems and For Normal Ventilation Systems Design Addresses Requirements in Regulatory Guide 1.52 SGTS none Control Room Habitability Area HVAC System none 19B.2-22 B-55: Improved Reliability of Target Rock Safety / Relief Valves D1 ABWR Uses a Direct Acting S/RV Design 2.1.2 ,L / (Design Des. Only) 19B.2 23 B-56: Diesel Reliability Independent Diesel Generators 3 2.12.13 Starting Reliability Per Demand .95 none Combustion Turbine Generator 1 2.12.11 19B.2-24 B-61: Allowable ECCS Equipment Outage Periods ECCS Capable of Being Tested During Plant Operation RCIC 2.4.4 2.4.2 HPCF 2.4.1. RHR Technical Specification LCOs in Accordance with Industry Practice none O 9

I s Table 12 qg Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 198.2-24 B-61: Allowable ECCS Equipment Outage Periods (Cont.) LCOs Accounted forin PRA none ABWR PRA Meets Goal of 1.0 x E-5 CDF Per Reactor Year none 19B.2-25 B-63: Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary Boundary Valves Designed, Fabricated and Tested According to ASME B&PV Code, Section ill Design Des. Identifies ASME Code Class for O System Components 2.4.1 RHR System 2.4.2 HPCF System 2.4.4 RCIC System 2.2.2 CRD System 2.2.4 SLC System CUW System 2.6.1 Nuclear Boiler System 2.1.2 Reactor Recirculation System 2.1.3 in-Service Testing of Boundary Valves Will Be Performed in Accordance with ASME B&PV Code, Section XI COL Req. 19B.2-26 B-66: Control Room Infiltration Measurements Normal AC Filtration Units Number of Divisions 2 2.15.5a Mechanically and Electrically Separate ---- 2.15.5a Number of Outdoor Air intakes 2 2.15.5a O i l 10 j l

Table 12 hp Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 198.2-26 B-66: Control Room infiltration Measurements (Cont.) Automatic Switch-over to Emergency Units on High Radiation in Air intake 2.15.5a Emergency Filtration Units Number of Units 2 2.15.5a Mechanically and Electrically Separate ---- 2.15.5a With HEPA and Charcoal Filters none Provisions for Site Unique Toxic Gas Sensors none (Interface O Req 2.15.5) Provisions to Detection Smoke 2.15.5a Airborne Radioactive Material 2.3.1 Provisions to Remove Smoke and Airborne Radioactive Material 2.15.5a 19B.2-27 C-1: Assurance of Continuous Long Term Capability of Hermetic Seals on instrumentation and Electrical Equipment Safety-related Electrical Equipment is Environmentally Oualified in Accordance with NRC Guidance including NUREG-0588 ---- Refer to 1.2(3) O 11

Table 12 O-c e # eric s eteiv i= = # e - ( c e "i-) verifying SSAR Entrv Parameter Value ITAAC 19B.2-28 C-10: Effective Operation of Containment Sprays in a LOCA SGTS Redundant 2.14.4 Filters Gaseous Effluent from Primary and Secaondary Containmnent 2.14.4 No. of RHR Subsystems Which Provide Containment Spray 2 2.4.1 Sprays Manually Initiated by Operator 2.4.1 Sprays Automatically Terminated When LPFL Injection Valve Opens 2.4.1 High Drywell Pressure Interlock On Drywell Spray Operation 2.4.1 19B.2-29 C-17: Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes COL Req. Will Meet Requirements of 10CFR61 19B.2-30 15: Radiation Effects on Reactor Vessel Supports Vessel Support Skirt Located Below Core Beltline 2.1.1 Wide Water Flow Region Between Shroud and Vessel Wall 2.1.1 19B.2-31 23: Reactor Coolant Pump Seal Failures N/A l (Not Applicable to ABWR) O l 1 12

Table 12 Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 198.2-32 25: Automatic Air Header Dump on BWR Scram System Scram Air Header has a Low Pressure Alarm none Scram initiated by Low Pressure in the Common Header Supplying the Charging 2.2.7 Water to the Scram Accumulators 19B.2-33 40: Safety Concerns Associated with Pipe Breaks in the BWR Scram System Ball-check Valve in the FMCRD Flange Housing at Connection of the Insert Line with the Drive Scram Port 2.2.2 0 19B.2-34 45: Inoperability of instrumentation Due to Extreme Cold Weather instrument Sensing Lines Are Located in Temperature Controlled Environments none 19B.2-35 51: Proposed Requirements for improving the Reliability of Open Cycle Service Water Systems A Closed Cooling Water System Will Be 'Jtilized which Transfers Heat Loads Via Heat Exchanger to Service Water System 2.11.3 Service Water Pump Discharge Piping Shall Be Equiped with a Strainer and/or Silt Removal Capability none (Not in Certified Des.) Materials for Piping, Pumps and Heat Exchangers Which Resist Water Chemistry Conditions Will Be Specified 'none l 13

l Table 12 qQ Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2-35 51: Proposed Requirements for improving the Reliability of Open Cycle Service Water Systems (Cont.) Provisions Will Be Made to Facilitate inspection of Service Water Piping COL Req. The Safety-Related Portions of the RCW and RSW Will Operate as Designed Assuming Loss of All Offsite Power 2.11.3 2.11.9 2.12.1 (Based on Redundancy) Assuming Any Single Failure 2.11.3 2.11.9 O 19B.2-36 057: Effects of Fire Protection Systems Actuation on Safety-Related Equipment l Fusible Link Sprinkler Heads none A Means of Fire Detection is Provided 2.15.6 Capability to isolate Flow Locally by i Manual Isolation Valve none All Rooms in the Reactor and Control Buildings with a Potential for Flooding Are Supplied With Floor Drains -2.9.2 Safety-Related Equipment Raised Off the Floor 2.15.10 Safety-Related Divisions i Number 3 2.15.10 2.15.12 i Mechanically and Electrically independent Covered by Individual Sys O Entries 14 ) +

Table 12 O ce#eric s teiv i - (co#t.) verifying SSAR Entry Parameter Value ITAAC 198.2-37 67.3.3: Improved Accident Monitoring Plant Post Accident Monitoring Variables Neutron Flux 2.7.1 Control Rod Position 2.7.1 2.11.20 Boron Concentration (Sampling Only) Reactor Coolant System Pressure 2.7.1 Drywell Pressure 2.7.1 2.7.1 Drywell Sump Level 2.7.1 Coolant Levelin Reactor Suppression Pool Water Level 2.7.1 Containment Area Radiation 2.7.1 2.7.1 Primary Containment Pressure Primary Containment isolation Valve Position 2.7.1 2.11.20 Coolant Gama (Sampling Only) Coolant Radiation 2.3.1 2.7.1 RHR Flow 2.7.1 HPCF Flow RHR Heat Exchanger Outlet Temp 2.4.1 RCIC Flow 2.7.1 2.7.1 SLC Pressure 2.7.1 SLCS Storage Tank Level 2.7.1 SRV Position 2.2.3 Feedwater Flow High Radioactivity Liquid Tank Level none. 2.7.1 Standby Energy Status Suppression Pool Water Temp 2.7.1 2.7.1 Drywell Air Temperature Drywell/ Containment Hydrogen Concentration 2.7.1 Drywell/ Containment 2.7.1 Oxygen Concentration 0 15

s Table 12 Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2-37 67.3.3; improved Accident Monitoring (Cont.)' Plant Post Accident Monitoring Variables Primary Containment Air Temp 2.7.1 Secondary Containment Airspace (effluent) Radiation Noble Gas 2.3.1 Containment Effluent Radioactivity - Noble Gas 2.11.20 (Sampling Only) Condensate Storage Tank Level 2.7.1 Cooling Water Temperature to ESF System Components 2.11.3 Cooling Water Flow to ESF System Components 2.11.3 Emergency Ventilation Damper Position--- 2.15.5 Service Area Radiation Exposure Rate ---- 2.3.2 O Purge Flows - Noble Gases and Vent Flow Rate 2.3.1 Identified Release points - Particulates 2.3.1 and Halogens Airborn Radio Halogens and Particulars--- 2.3.1 Pant and Environs Radiation / Radioactivity (Portable Instruments)---- COL Req. Meterological Data (Wind Speed, Wind Direction, and Atmospheric Stability)-- COL Req. On Site Analysis Capability (Primary Coolant, Sump and Containment Air Grab Sampling) COL Req. 19B.2-37.1 73: Detached Thermal Sleeve Thermal Steves Are Welded to the Vessel Nonle Safe-Ends FW Injection Une Nonle none HPCF Injection Line Nonle none RHR Injec; ion Line Nonle none O 16

i r Table 12 t Generic Safety issues (Cont.) s i Verifying SSAR Entry Parameter Value ITAAC 19B.2-38 75: Generic implications of ATWS Events at Salem Nuclear Plant Separate Scram Groups 4 2.2.7 Solid State Load Drivers Per Scram Group 8 3.4 Contactors for Manual Scram Per Scram Group 2 3.4 198.2-39 78: Monitoring of Fatigue Transient Limits for Reactor Coolant System Environmental Effects are included in the Design Bases for Carbon Steel RCPB Components Piping none Reactor Vessel Nozzle Safe Ends none CDF includes Environmental Effects on Fatigue Resistance of Materials none 19B.2-39.1 79: Unanalyzed Reactor Vessel Thermal Stress During Natural Convection Cooldown i Procedural Controls on Plant Operation to Address Pressure /Temeprature Limits for _I the Reactor Coolant System COL Req. i i 17

i Table 12 Generic Safety issues (Cont.) Verifying SSAR Entrv Parameter Value ITAAC 19B.2-40 83: Control Room Habitability Control Room HVAC Filtration System Referto 19B.2-26 Control Room Designed to Withstand Effects of Natural Phenomena 2.15.12 Minimize Combustible Material none Individual Respirators Available COL Req. Site-Specific Consideration of Potential Chemical Releases none (Interface Req. 2.15.5) Fire Alarm System Provided 2.15.6 Fire Hr. s and Portable Fire Extinquishers 2.15.6 Available 198,2-41 86: Long Range Plan for Dealing with Stress Corrosion Cracking in BWR Piping Use only Stainless Steel Type 316 with a Maximum Carbon Content of.02% none All Materials Supplied in the Solution Heat Treated Condition none impurity Levels in Reactor Water Are Maintained at a Minimum none Provisions Will Be Made for inservice COL Req. inspections 1 O 10 l )

Table 12 O aemeric s reiv i

  1. e- (co i >

Verifying SSAR Entry Parameter Value ITAAC 19B.2-42 87: Failure of HPCI Steam Line Without i Isolation In-Service inspection Program for MOV Used for Isolation of CUW and RCIC COL Req. Opening and/or Closing of installed MOVs Used for isolation of CUW and RCIC Will be Conducted Under Peroperational Differentail Pressure, Fluid Flow and i Temperature Conditions 2.4.4 2.6.1 Remote Manual Air Operated Valve on Bottom Head Drainline none Flow Restrictor in CUW Main Suction Line 2.6.1 Bottom Head Drainline Tees into CUW O' 2.6.1 Suction Line at an Elevation Above TAF 19B.2-43 89: Stiff Pipe Clamps Pipe Clamps Only Installed on: Straight Runs of Pipe none Bends with a Radius of at least 5 Pipe Diameters none 198.2-44 103: Design For Probable Maximum Precipitation ] Design Maximum Rainfall Rate (cm/hr) 49.3 5.0 i (Site Parameters) Design Maximum Short Term Rate (cm/5 min) 15.7 5.0 i (Site Parameters) - Evaluation to Ensure Specific Site Data Within ABWR Envelope Parameters COL Req. l O 19 i

i Table 12 Generic Safety issues (Cont.) l Verifying SSAR Entry Parameter Value ITAAC 19B.2-45 105: Interfacing System LOCA at BWRs Design Pressure of Some Low Pressure Components Upgraded to 28.8 atg (Design Description Only) RHR System 2.4.1 2.4.2 HPCF System RCIC System 2.4.4 2.2.2 CRD System SLC System 2.2.4 (later) CUW System 2.6.1 (later) FPC System none Nuclear Boiler System none Reactor Recirculation System none MUWC System none MUWP System none Radwaste System (LCW Receiving Tank and HCW Receiving Tank) none 19B.2-46 106: Piping And the Use of Highly Combustible Gases in Vital Areas Compressed Gas Systems of HWC and Main Generator Hydrogen System COL Reg. Designed in Compliance with SRP 9.5.1 19B.2-47 (Not Used) 198.2-48 118: Tendon Anchorage Failure Primary Containment Structure is of a Reinforced Concrete Design 2.14.1 19B.2-49 120: On-Line Testability of Protection Systems Manual and Automatic Testability of RPS, LDIS and ECCS Initiation Logic During Reactor Operation 3.4 O 20

r e - Table 12 ( Generic Safety issues (Cont.) F Verifying SSAR Entry Parameter Value ITAAC 19B.2-50 121: Hydrogen Control for Large, Dry PWR Containment (Not Applicable to BWRs and Pressure Suppression Containment) Containment inerted During Normal Operation 2.14.6 19B.2-51 124: Auxiliary Feedwater System Reliability (Not Applicable to BWRs) 19B.2-52 128: Electrical Power Reliability Four Class 1E dc Divisions 2.12.2 Two-out-of-Four Logic to Activate O Safety Systems 3.4 Two-out-of Three Logic to Activate Safety Systems if One Division is Out of Service 3.4 No Bus Tie Breakers Between Divisions none Non-Class 1E Loads (Except FMCRD Motors) are Powered from Non-Class 1E Sources 2.12.1 FMCRD Motors Use Class 1E Power Which has Zone-selective Interlocked Circuit Breakers. none de Divisions Backed by Class 1E D/Gs 2.12.1 2.12.12 Non-Class 1E PIP Loads Backed by i Off-site Combustion Turbine Generator 2.12.1 O 21

r Table 12 O Generic Safety issues (Cont.) Verifying SSAR Entrv Parametnr Value ITAAC 19B.2-52 128: Electrical Power Reliability (Cont.) Two Separate and independent Connections from the Off-site Sources to Each Class 1E Bus and to Each PIP Bus. none (Interface Req. 2.12.1) 19B.2-53 142: Leakage Through Electrical isolators in instrument Circuits Fiber Optic Isolation Devices Used for Electrical isolation of Logic Level and 3.4 Analog Signals 19B.2-54 143: Availability of Chilled Water Systems and Room Cooling /^ Safety-Related HECW System Provides Chilled Water to Main Control Room Air Conditioning, DG zone Coolers and Control Building Essential Electrical i Equipment 2.11.6 i Essential Equipment HVAC System Provides Controlled Temperature Environment for Safety-Related 2.15.5 Equipment Under Accident Conditions ABWR Can Withstand a Station Blackout for 8 Hours none 19B.2-55 145: Actions to Reduce Common Cause Failures l Maintenance Program in Accordance with 10CFR50.65 COL Req. 19B.2-56 151: Reliability of Anticipated Transient Without l Scram Recirculation Pump Trip in BWRs Gate Tum-off Thyristors (GTOs) Are Switched Off to Trip RIP ASD Inverters none Oi 22

r j Table 12 p) Generic Safety issues (Cont.) y Verifying SSAR Entrv Parameter Value ITAAC 19B.2-57 153: Loss of Essential Service Water in Light-Water Reactors RSW Divsions Total Number 3 2.11.9 Physically and Electrically Separate 2.11.9 Number Required for Safe Shutdown 2 none Pumps per Division 2 none (Not in Certified Des.) RCW Heat Exchangers per Divsion 3 2.11.3 RSW Design includes Adequate NPSH at Low UHS Water COL Req. Levels Low Point Drains and High Point Vents --- COL Req. Prevention of Organic Fouling COL Req. Component Material for Site Water Conditions COL Req. Protection from Flooding, Spraying, Steam impingement, Pipe Whip, Jet Forces, Missiles, Fire, Failure of Non-Seismic Category l Equipment andIce COL Req. 19B.2.58 155.1: More Realistic Source Term Assumptions Dose Commiment to 10CFR100 Limits none O 23

7 3 i Table 12 -( Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2.59 A-17: Systems Interaction in Nuclear Power Plants r Redundant Safety-Related Equipment and Systems Divisionally Separated Covered by Multiple Sys. Entries Redundant Electrical Power Systems Covered by Divisionally Separated Multiple Sys. Entries Divisions Designed Against 2.15.10 Intra-Divisional Flooding 2.15.12 Safety Grade and Non-Safety Grade Systems Spatially Separated none O 19B.2.60 A-29: Nuclear power Plant Design for the Reduction of Vulnerability to Industrial Sabotage Redundant Safety-Related Equipment and Systems Divisionally Separated Covered by Multiple Sys. Entries Redundant Electrical Power Systems Divisionally Separated Covered by Multiple Sys. Entries Controlled Access to Safety-Related Areas -- 2.16.3 (Design Des. Only) Indication in Control Room for Postulated Safeguard Information Sabotage Activity 10CFR73 Part 21 O 24

g Table 12 (- Generic Safety issues (Cont.) Verifying SSAR Entry Parameter Value ITAAC 19B.2.61 B-05: Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containments Dynamic Loads for Two-Way Reinforced Concrete Stabs Determined by the Methodoigy identified in ACI 349-85 none 19B.2.62 029: Bolting Degradation or Failure in Nuclear . Power Plants Only Material Resistant Against Corrosion Wastage and Intergranular Stress Corrosion Cracking Used none RCPB Component Fabricated, Tested and Installed in Accordance with ASME Code, 2.1.1 Sections til and XI (Design O 4 Des. Only) COL Req Perform Periodic Inservice inspections Double O-Rings for Vessel Main Closure none Leak Detention Devise Between O-Rings none t O 25

Table 12 Generic Safety issues (Cont.) Verifying SSAR Entrv Parameter Value ITAAC 198.2.63 82: Beyond Design Basis Accidents in Spent Fuel Pools Spent Fuel Pool Seismic Category 1 2.15.10 Stainless Steel Uned none l 2.6.2 Low Water Level Alarm (Level Indication only) 2.6.2 Over-Flow Weirs to Skimmer t Check Valve in Discharge Line 2.6.2 Automatic Inventory Makeup none High Flow Alarm on Liner Drain none 19B.2-64 113: Dynamic Qualification Testing of Large Bore Hydraulic Snubbers Perform Dynamic Tests with Expected As-Installed Conditions none i { i O 1 26 i .}}

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