ML20057D843

Submits Initial Staff Comments on GE ABWR Tier 1 930831 Submittal Re ABWR Certified Design Material.Submittal Adequately Reflects Agreements Reached for Tier 1 Material

ML20057D843
Person / Time
Site:	05200001
Issue date:	09/27/1993
From:	Boyce T Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9310060058
Download: ML20057D843 (24)

Text

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Dockrt No.52-001 September 27, 1993 i

Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Naclear Energy 175 Curtner Avenue San Jose, California 95125

SUBJECT:

INITIALS COMMENTS ON GE NUCLEAR ENERGY (GE) ADVANCED BOILING WATER REACTOR (ABWR) CERTIFIED DESIGN MATERIAL Enclosed are initial staff comments on the GE ABWR Tier 1 submittal dated August 31, 1993.

In general, the submittal adequately reflected the agreements reached for the Tier 1 material during a meeting on July 27-29, 1993.

In a letter of September 17, 1993, GE provided the Tier 2 portions of these agreements as part of Amendment 32 to the standard safety analysis report.

Based on the staff's review of this amendment, you may receive additional comments related to the Tier 1 material.

Sincerely, (Original signed by)

Thomas H. Boyce, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

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WTravers RJones, 8E23 AThadani, 8E2 MMiller 8E23 TBoyce JMoore, 15B18 TGody, Jr., 17G21 BHardin, RES LShao, RES AVietti-Cook WRussell, 12G18 J0'Brien, RES CMcCracken, 801 JLyons, 801 RPerch, 8E2 GBagchi, 7H15 DTerao, 7H15 MChiramal, 8H7 DEckenrode,. 10D24 REmch, 1004 GGrant, 17G21 GMizuno, 15B18 MFinkelstein, 15B18 TPolich, 9Al DThatcher, 7E4 TCollins, 8E23 MRubin, 10E7 CBerlinger, 7E2 RLatta, 9Al AThadani, BE2 BBoger, 10H1 S FCongel, 10E2 l

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4 Mr. Patrick W. Marriott Docket No.52-001 General Electric Company cc:

Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.

12300 Twinbrook Parkway Suite 300 Suite 315 Washington, D.C.

20006 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C.

20460 i

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

20585 Marcus A. Rowden, Esq.

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

Suite 800 Washington, D.C.

20004 Jay M. Gutierrez, Esq.

Newman & Holtzinger, P.C.

1615 L Street, N.W.

Suite 1000 Washington, D.C.

20036 Mr. Steve Goldberg l

Budget Examiner 725 17th Street, N.W.

Room 8002 Washington, D.C.

20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 i

Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874

PROJECTS COMMENTS 1.

Interface requirements should include methodology to meet interfaces in Section 4.

2.

QA problem - Figure 2.15.10a (Rev.1)

Lines on top of building in figure do not line up with the building.

3.

QA problem - Figure 2.15.6, Fire Water Protection Water Supply System.

Check valves shown on figure are backwards.

If system was built according to the figure, the system would not perform its intended function.

4.

Section 1.1, Definitions.

Delete the word "other" in the definition of Type Test.

5.

The GE request to use Japanese metric units in the design certification material is under staff review, and will be responded to separately.

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4 PROJECTS COMMENTS ON GE ABWR SSAR SECTION 14.3 General: This section needs to be complete and detailed enough to document the basis for the staff's finding in the FSER that the ITAAC are "necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the 1

provisions of the Atomic Energy Act, and the Commission's rules and regulations." (Section 52.97(b)(1)).

1.

Discusses criteria for level of detail in Tier 1 material, but not the methodology of development.

- Needs to discuss conduct of development of material (e.g.,

multidisciplinary review, industry review, etc.).

- Needs to discuss how the technical requirements for each system were identified (different sections of SSAR supporting each system).

Needs i

to discuss systematic selection process for Tier 1 information.

- Needs to discuss how safety assumptions and insights were developed and incorporated.

2.

Should discuss unique aspects of specific discipline areas such as I&C, electrical, structural, etc.

Submittal focuses on generalities instead.

3.

Criteria are mixed for "in" and for "not in" Tier 1.

Could be organized into criteria for "in" ir; one section and criteria for "not in" in another section, or similar criteria could be combined in one paragraph.

4.

Does not provide rationale for most criteria, only selected criteria.

Can rationale be specified for all criteria?

5.

Should reference Section 52.97 rather than 52.47.

6.

Should include discussion of the development of the DAC, site parameters and interfaces as stated in the first paragraph.

7.

Additional comments are provided as markups to Section 14.3.

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.ta wswr 14,3 Methodology 4or etermining the Gontents of the Certified Design Material 14.3.1 Criterla The certified design material consists of 1) design descriptions of systems and their associated inspections, tests, analyses, and acceptance criteria (ITAAC),2) descriptions of design and construction activities that are applicable to more than one p[g,. GJS system, together with their associated ITAAC,3) interface requirements for those sr' f ' ~ #~

systems that are not within the scope of the ABWR Standard Design but which are necessary to support tue ABWR, and 4) site parameters which identify bounding IM conditions for an acceptable site for the ABWR Standar,d Design.

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istandardswhich pertain to the safety of the plant. The following criteria were j

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,g" considered in determining which information warranted incit clon in the certified design descriptions:

(1) The information in the certified design descriptions is tr. ken from the l

Standard Safety Analysis Report (SSAR) for the ABWR.

i (2) l Not allinformation in the SSAR warrants inclus. ion in the certified desig3 l

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i descriptions.iinstead, the certified design descs ptions contain only that J Eformation in the SSAR that is most significant to safety. C determining what is most significant to safety, several factors were considered, including the following:

(a) Whether the feature or function in question is necessary to satisfy thc/SRC"s recubtinns in Part 5 JhdrS W ld wN[ETN; f

) ' Whether the Teature or tuncuon m quesuon pertains to a safety-related structure, system, or component.

(c) Whether the feature or function in question is specified in the NRC's Standard Review Plan as being necessary to nl} Sno0Lb UL1LVb C

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(d) Whether the feature or function in question represents an I

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preventing or mitigating severe accidents.

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Whether the feature or function in question has had a significant impact r4 the safety or operation of existing nuclear power plants.

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• 4 21A6100Rev 2 ABWR Standard Safety Analysis Report (g) Whether the feature or function in question is typically the subject of a provision in the technical specifications.

The absence or existence of any of one of these factors was not [ 3% disp %e in determining which informadon is significant to safety. -instead, these and other factors were taken into account in making this determination. (S) The safety-related features and functions of structures, systems and components are discussed in the certified design descriptions. p on-s safe %cs and funcdons of safetMy-structures, systems, and , components are no iscussed in the certified design descriptions. pMy [ 7) Structures, systems, and components that are not classified as safety-(4 related are discussed in the certified design descriptions only to the ("f56, rw#rEfextent that theylperform significan(safet/ functions or have features to prevent a significant adverse impact upon the safety-related functions of ther structures, systems, or componentM (5) In general, the certified design descriptions for structures, systems, and components are limited to a discussion of design features and functions. The design basis of structures, systems, and components, and 7'g' explanations of their importance to safety, are provided in the SSAR apd,.g 'M' m m u y f >Q.s h.'"'Cf Mn, a. s t\\ are not included in the certified desi n degr1_ti ns. i % (6) The certified design descripdons f{cus on' hardware. The certified design descriptions do not contain requirements related to operating conditions or to operadons, maintenance, or other programs because these matters are controlled by other means such as the technical specifications. (7) TM-eernfiedtesign.4e4aipdent-ditems4he-dc4ge e gtgg,,, r synems.2ndampenems. In general, the certified design descripdons do not discuss the process for designing and constructing a plant that _. j references the ABWR Standard DAbecause the safety funcdon of a ,f

gafi f structure, system, or component is dependent upon its final as-built condition and not the process used to achieve that condidon.__ _

l ~GcFp~tnvolve weldi6ginil deTsigns that ar_c dependent upon as-built 7,, e or as-procured information, such as pip ng, instruments and conuoW) '>#.ld and radiation protection J a:*~ W2GW, '47D ~ ~ ' (~q wp, (S) In general, the certified design descriptions address fixtures, not N 5 A portable equipment or consumables which are frequently changed. One exception to this general rule pertains to nuclear fuel, fuel channels, and control rods. These components are discussed in the certified design Methodologyfw Driernming the Contents of the Cernfied Design Material DrcJt 2

i .c c . J.-J a, 21A6100 Rav 2 ABWR standard sarety Anarysio nopart r descriptions due to their importance tgbnd the desire to control fth2r design thioighouithe'lileame of a plant that references the ABWR ~- cerufied design. ~- (9)_ _In general, the certified design descriptions do not discuss component +. ' C/,,) manufacturers because thrre :: cf m a variety of component ty 3 types (e.g., valve types), component internals, or component q internals, or manufactt rers.tha: can ps.fsm the rJe:y funs..,ig - - yn tpcuica-fg [4 g l'.e p p W Arp ) J - a 7 (a.aw. 1 (10) The certified design descriptions cannot and do not contain any proprietary informadon. (11)lIn order to allow the applicant or licensee of a plant that references the )~ f ABWR Standard Design to take advantage ofimprovements in # '~~ ' technology,Ftlic certified ~ design descriptions in general do not prescribe {Mesig$Yn catures that are the subject of rapidly evol ,j j hc s. 6 hep. qd 8A viNbr%.21.0m$. AN:bk- &m a Mr M ~~ h >- b ~9 ele The ITAAC for the certified desyn consist of those inspections, tests, and analyses that / ( V't, Y w r ::=f.+ ' are necessary and sufficient to demonstrate that a plant that references the ABWR i Standard Design is built and will operate in accordance with the mused design u#NMc - g i' I. g'/. dei;dvus. The following criteria were considered in determining which { 3 ( informadon warranted inclusion in the certified design ITAAC:

, g (1) The scope and content of the ITAAC correspond to the scope and

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n-(2) One inspection, test, and ahalyses may verify one or more provisions in PM@ *. the cerdfied design description. In pardcular, an ITAAC which calls for ~ a system funct.onal test or an inspection of basic configuration may verify a number of provisions in the certified design descripdon. Therefore, i there is not necessarily a one to-one correspondence between the ITAAC and the certified design descriptions. I x65N The ITAAC must be cmpmd prior to fuelloading. ThereforQ,f/> j (S) e ITAAC do not include any inspections, tests, or analyses that are dependent upon conditions that only exist after fuel load. (4) ITAAC verify fixtures in the plant. Therefore, there are no ITAAC for nuclear fuel, fuel channels, and control rods, because they are consumables which are frequently changed by a licensee. Additionally,it is not possible to have ITAAC for nuclear fuel, fuel channels, and control / rods. because these components are not installed in the plant until after authorizadon is given for fuel load. i MrAvdologyfor Driermining the Contenu qf the Cmiped Desugn Material - Dr$ 3 4

' AU:. IO L: hAi!. 5E IFU EM E!.! ~ ! 't/ l P E. II 21 A0100 RQv 2 ABWR Standard Safety Analysis Report in general, the ITAAC verify the safety featurehr functionhf structures, (5) systems, and components. With hmited exceptions (e.g., welding), the ITAAC do not address typical construction processes for the reasons discussed previously. 14.3.2 Appilcation of the Critoria Using the criteria and provisions discussed above, tesiews were performed of each system (including structures) within the scope of the ABWR Standard Design to determine the contents of the certified design descriptions and ITAAC. Because the safety-significance of the systems varies, a graded approach was taken in preparing the certified design descriptions and ITAAC. This approach is summarized below: Safety related systems - Discussion of the major safety-related features and performance characteristics of the system. Y Systems needed for beyond design-basis events _-(Brie)@scussion of the safety $ features and performance characteristics. ..p.63 7 Systems potentially impacting safety-Discussion of the aspects of these systems that have safety significance. i q Systems which affect overall p%nt design - Discussion of any aspect of the , design that affects the overall design of other safety-significant 9 stems. Other non-safety-related systems - No discussion, except identification of the j system title. For safety related systems, the certified design descriptions generally include the following informadon, as applicable: the system's name and scope; the system's purpose; system's safety-related modes of operation; the system's classification (i.e.. safety-related Seismic Category, and ASME Code Class); the system's location; the basic configuration of the system's safety-significant components (usually shown by means of a figure); the type of electrical power provided for the system; the electrical independence and physical separation of divisions within the system; the system's important instruments, controls, and alarms, to the extent located in the Main l Control Room or Remote Shutdown System; identification ofwhich of the important Class lE electrical equipment within the system is qualified for a harsh environment; motor-operated vah*es within the system that have an active safety-related function; and any other features or functions that are sigraficant to safety. The certified design descriptions for non-safety-related systems also include the information listed above, but only to the extent that the information is relevant to the i Methodologyfor Determining the Contenu of the Certified Design Ma:erial Drqft 4 I i

2 I- . ;rf :a) 4 ',, i, i.., 21A6100 Rev. 2 ABWR stansora sorsey Anarynin nuort system and is significant to safety. Since much of this information is not relevant to non safety-related systems, the cerdfied design descriptions for non-safety-related systems are generally substantially less extensive than the descriptions for safety related f J rg// sy stems. m e/ L- _p J /4 3,J In addidon to reviewing each system, reviews were also conducted of sections in the SSAR that cross system boundaries to determine whether any informadon in these sections should also be included in the certified design descriptions ofindividual systems (if not already included). In particular, such reviews were performed of the t 3 flooding analysis in Chapter 3 of the SSAR, the analysis of overpressure protection in . idhi Chapter 5 of the SSAR, containment analyses in Chapter 6 of the SSAR, the core $6"Ib,y cooling analyses in Chfpter 6 and 15 of the SSAR, the analysis of Fire Protection in { Cahpter 9 of the SSAR, the safety analysis of transients and the anticipated transients n 'j without scram in Chapter 15 of the SSAR, the radiological analyses in Chapters 15 and 19 of the SSAR, and the importan' insights from the probabilistic risk a.ssessmen in v Chapter 19 of the SSAR. ,4 y Tables 14.5-1,-2, and -3 provide examples of the types ofinformation that were i idendfied from these reviews as warranting discussion in the certified design descripdons. This information has been provided in the certified design descriptions i for the ABWR MeshoMogyfor Determining the Contents of the Cerufled Design Mmenal Draft 5

j REACTOR SYSTEMS COMMENTS ON GE ABWR TIER 1 REV. 1 1. In our July 8, 1993 FSER on Chapter 5.4.6, RCIC, the RCIC system was approved based on its ability to function without AC power for at least 8 hours. Amendment 32 changed both Tier 1 and Tier 2 material to indicate that RCIC will now be designed to function without AC power for only 2 hours. This is unacceptable to the staff, and will remain an open item until GE modifies Tier 1 and Tier 2 materials to indicate ability to function for 8 hours. 4 1 1 1 b I

SEVERE ACCIDENT BRANCH COMMENTS ON GE ABWR TIER I REV. I 1 Section Description 1 2.1.1 Submit DD change "The vessel skirt does not have openings connecting the upper and lower drywell regions." (GE fax dated 9/15/93) 2 2.14.6 Submit DD change "The COPS pneumatic actuated valves shown on Figure 2.14.6 have active safety-related functions to bothopenandclose,andperformthpsefunctionsagainsta differential pressure of 7.34 kg/cm (absolute) +/- 5% and under fluid flow and temperature conditions." (GE fax dated 9/15/93) 3 2.14.6 Submit ITAAC change - ITAAC 8 to indicate " perform thesp functions against a differential pressure of 7.34 kg/cm I (absolute) +/- 5% and under". (GE fax dated 9/15/93) i 4 2.14.6 Submit Figure change - Add footnote to Figure 2.14.6 to indicate "3. The ACS include provisions for supplying nitrogen to the COPS piping between the inner and outer rupture disks." (GE fax dated 9/15/93) 5 2.14.1 Propose DD change "Each vacuum breaker has a position indication switch that provides position indication and an alarm in the main control room (MCR) with adequate sensitivity to detect the suppression pool bypass capability of the containment." (0 pen item 248 and J-7) i 6 2.14.1 Propose new ITAAC "The vacuum breaker position indication switch can detect a valve opening of ???..." (0 pen item 24B 4 and J-7) 7 2.14.1 Containment Sumps - We are still evaluating this issue and will provide comments ASAP. i

ABWR Tier 1 Certified Desian Material Comments Auaust 31. 1993 Submittal Structural Enaineerina i 2.14.1 Primary Containment System Desian Description In the second paragraph on Page 2.14.1-3, the first sentence should read, "The containment internal structures designated Seismic Category I are designed and constructed to accommodate the dynamic and static load conditions........" 2.15.10 Reactor Buildina ITAAC i ITAAC #8 and #9 provided in the April 23, 1993 submittal appear to have been inadvertently deleted or else they were moved to page 2.15.10-19 i which was not included in the final copy. They should be reinstated in Table 2.15.10 on Page 2.15.10-18 (or 2.15.10-19). i 2.15.12 Control Buildina l. Desian Description a. In the fifth paragraph on Page 2.15.12-2, the second sentence should I read, "It is designed and constructed to accommodate the dynamic and static load conditions..........." i b. In the same paragraph, " fires" should be deleted from Load l Condition #2, Internal Events. 2. Fiaures 2.15.12a throuah 2.15.12a i a. The panel wall (El. 3500mm) shown in Figure 2.15.12a adjacent to Column Line CD should also be shown in Figure 2.15.12e. i b. The panel wall (El. 7900mm) shown in Figure 2.15.12b between Column Lines C5 and C6 should also be shown in Figure 2.15.12f. c. The panel wall (El.12300mm) shown in Figure 2.15.129 adjacent to l t Column Line C1 should also be shown in Figure 2.15.12b. l 2.15.13 Radwaste Buildina Desian Description j In the fifth paragraph on Page 2.15.13-1, the first sentence should read, "The external walls of the RW/B below grade and the basemat are desianed and constructed to accommodate the dynamic and static loading conditions........."

ELECTRICAL COMMENTS ON GE ABWR TIER 1 REV. I 1. In ITAAC table 2.12.1, there appears to be a typo in the Acceptance Criteria column of entry 5. 2. ITAAC figure 2.12.12 shows a connection to "CVCF Inverter" but figure 2.12.14 shows a connection to the "CVCF Power Supply." It appears these should read the same. 3. Design description 2.12.12 was revised regarding the manual connections between divisions via the soare battery chargers which were removed from the description and the figure. The commitment should be simplified to avoid confusion. The commitment should state no automatic connections between divisions and interlocks are provided to prevent manual paralleling / connection between divisions. The ITAAC should also be revised to match any changes. 4. The design description for the EPD (2.12.1) has a commitment that there w are no automatic connections between divisions. To be consistent, such a design commitment regarding no automatic connections between divisions should be included for design descriptions 2.12.14 and 2.12.15. 5. In the design description for lighting (2.12.17), the commitment regarding non-Class lE AC standby lighting serving passageways was deleted. Was this intentional or a typo? If it was intentional, provide the basis.

u PLANT SYSTEMS COMMENTS ON RWCU DESIGN DECRIPTION t 1. Add statement discussing system isolation on 'a '. tion of SLCs and the Leak Detection System (LDS). 2. Add statement that safety-related components (isolation valves) are physically separate. i ITAAC 1. Add tests to verify that system isolates on initiation of SLCs and actuation of the LDS. 2. Add inspection to verify physical separation of safety-related components (isolation valves). SSAR I Power supplies not discussed in SSAR text. This discussion should be added. 1 i I

SPLB COMMENTS ON MODIFICATIONS TO THE ABWR ITAAC The following comments are provided on GE's revised DD & ITAAC. This does not include the review of Amendment 32 to verify that the requested SSAR changes have been made. 1. 2.9.1 - RADWASTE SYSTEM a. GE committed to adding "These valves close upon receiving a LOCA signal" to the Design Description (DD). As of Amendment 32, this t has not been added. b. The staff is still unclear as to why the drain system for the "non-ECCS" areas of secondary containment do not need backflow protection. 2. 2.10.7 - MAIN TURBINE Add "IVs" under actions for protective action #3, on page 2.10.7-2. 3. 2.11.3 - REACTOR BUILDING COOLING WATER GE added a level detector on the RCW standpipe shown on j rigures 2.11.3 a-c. The sensor on Fig. 2.11.3c is not correct. 4. 2.14.4 - STANDBY GAS TREATMENT SYSTEM Negative pressure should be relative to the " surrounding spaces", not "outside atmosphere" in the DD and ITAAC. Outside atmosphere is a sub-set of surrounding spaces since all sides of the structures housing SGTS are not exposed to atmosphere. 5. 2.15.5 - HEATING. VEN.11LATING AND AIR CONDITIONING SYSTEMS a. The MCAE should be maintained at least 3.2 mm water relative to the surrounding s-s, not the outside atmosphere. This is the regulatory po. '. ion in SRP 6.4. b. The TSC should be maintained at a positive pressure relative to the surrounding spaces, not the outside atmosphere. 5. 2.15.10 - REACTOR BUILDING ITAAC entries # 8 and 9 are missing. Note that Rev. 1 of the ITAAC was supposed to add pages 2.15.10-19/20 that may contain these ITAAC entries, but the pages were not in the package.

I&C COMMENTS ON GE ABWR TIER 1 REV. 1 1. See markups attached. l If ~ * - - - ,m.- ,r,.

E Table 2.2.1 Rod Control and Information System b Y D3 inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspection, Tests. Analyses Acceptance Criteria 1. The equipment comprising the RCIS is 1. Inspections of the as-built system will be 1. The as built RCIS conforms with the defined in Section 2.2.1. conducted. description in Section 2.2.1. 2. The RCIS consists of redundant 2. Tests will be performed by simulating 2. There is no loss of RCIS output upon loss microprocessor based controllers (except failure of each operating RCIS controller. of any one controller. for controllers associated with individual FMCRDs). 3.% fC2f/ W E A rod wortu minimizer which uses control 3. Tests will be conducted on the RCIS using 3. A control rod block signal occurs when an ^ rod position signals to enforce simulated control rod position signals, out-of-sequence control rod movement is preestablished sequences for control rod and simulated neutron flux signals. simulated and when reactor power is movement when the reactor power below the low power setpoint. (neutron flux) is below the low power setpoint by issuing a control rod block y signal when an out of sequence control M rod movement is attempted. h, 4. The RCIS provides an automatic thermal 4. Tests will be conducted on the RCIS using 4. A control rod block signal occurs upon f power monitor which uses control rod simulated control rod position signals, simulation of a control rod movement o position signals, neutron flux signals, and neutron flux signals, and fuel operating which would cause fuel thermal limits to fuel operating thermal limits to enforce thermal limits, be approached. fuel thermal limits when the reactor power is above the low power setpoint and the plant is in automatic operation. m 8. 5. The RCIS provides a selected control rod 5. Tests will be conducted on the RCIS using 5. A control rod insertion signal occurs for 9 run in function which uses a signal from simulated control rod run-in signal from those positions assigned to this function e j the RFC System to insert selected control RFC System. upon receipt of a simulated signal from j. s rods into the core. the RFC System. = n 6. The RCIS provides an automatic control 6. Tests will be conducted on the RCIS using 6. A control rod run-in signal occurs upon 5, rod run-in which uses a scram-follow a simulated scram-follow signal from the receipt of a simulated scram-follow h f signal from the RPS to insert all control RPS. signal. g. [ rods into the core. m h. E l 3 3 E 1 O O O

s Table 2.7.5 Essential Multiplexing System b P IXl Inspections, Tests, Analyses and Acceptance Criteria Q Design Commitment inspections, Tests, Analyses Acceptance Criteria [ D 1. The equipment comprising the 1. Inspection of the as-built EMS and NEMS 1. The as-built EMS and NEMS conforme K Multiplexing System is defined in Section will be conducted. with the description in Section 2.7.5. 2.7.5. 2. EMS uses a deterministic 2. Tests of the EMS communications 2. EMS uses a deterministic communications protocol. protocol will be conducted in a test communications protocol. facility. 3. Data communications from EMS to non-3. Tests on the EMS data communications 3. EMS communications only permits data safety-related systems or devices uses an will be conducted in a test facility. transfer from the EMS to the non-safety-isolating transmission medium and related systems or devices. Control or buffering devices. Data cannot be timing signals are not exchanged transmitted from the non-safety related between EMS and non-safety-related side to EMS. systems or devices. M> 4. The EMS features automatic self-test and 4. Tests will be conducted on each as-built 4. There is no loss of EMS data { automatically reconfigures after EMS division by individually simulating communication as a result of the fault. s detecting failure of one channel (either a the following, while s'imultaneously Fault occurrence is displayed in the main 2 cable break or device failure) within a transmitting and monitoring test data control room. o division. The system returns to normal streams; operation after reconfiguration with n a. Single cable break. interruption of data communication. b. Loss of one RMU. c. Loss of one CMU. 5. Loss of data communicationsin a d'nsion 5. Tests will be performed in one division of 5. Data communication is lost without of EMS does not cause transient u EMS at a time. While simulated input generation of transient or erroneous 5' erroneous data to occur at system signals are being transmitted cable signals. outputs. segments in redundant paths will be g disconnected and EMS outputs R 5 monitored. I E: ? it: 2t E E s a. 9 9 9

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25AS447 Rev. O ABWR oesign cenisestian usterist J 3.4 Instrumentation and Control Introduction Subsection A provides a description of the configuration of safety-related, digital instrumentation and control (I&C) equipment encompassed by Safety System Logic and Control (SSLC). Subsection B contains a description of the hardware and software development process used in the design, testing, and installation ofI&C equipment. This includes descriptions of the processes used to establish programs that assess and mitigate the effects of electromagnetic interference, establish setpoints for instrument channels, and ensure the qualification of the installed equipment. Subsection C discusses the diverse features implemented in I&C system design to provide backup support for postulated worst-case common-mode failures of SSLC. IA c The devices addressed in this section are electronic components of the ABWR's safety-- rela +ed systems. These components are configured as real-time microcontrollers that use microprocessors and other programmable logic devices to perform data acquisition, data communications, and system logic processing. These components also contain automatic, on-line self-diagnostic features to monitor these tasks and off-line test capability to aid in maintenance and surveillance. The operating programs for these controllers are integrated into the hardware as firmware [ software permanently stored in programmable read-only memory (PROM)]. A controller's operating system can permit field adjustment of selected parameters under prol,er change control. Adjustable parameters are stored in electrically-alternate read-only memory (EAROM) or equivalent. A. Safety System Logic and Control Design Description Safety-related monitoring and trip logic for the plant protection systems resides in SSLC equipment. SSLC integrates the automatic decision-making and trip logic functions and manual operator initiation functions associated with the safety actions of the safety-related systems. SSLC generates the protecdve function signals that activate reactor trip ) and provide safety-related mitigation of reactor accidents.The relationship between SSLC and systems for plant protection is shown in Figure 3.4a. i SSLC equipment comprises microprocessor-based, software-controlled signal j processors that perform signal conditioning, setpoint comparison, trip logic, ssstem initiation and reset self-test, calibration, and bypass functions. The signal processors associated with a particular safety-related system are an integral part of that system. Functions in common, such as self-test, calibration, bypass control, power supplies and certain switches and indicators, belong to SSLC. However, SSLC is not, by itself, a svstem; SSLC is the aggregate of signal processors for several safety-related systems. SSLC hardware and software are classified as Class 1E, safety-related. Instrumentation and Control 3.4-1

~ 4 J l Table 3.4 Instrumentation and Control b a 12 j inspections, Tests, Analyses and Acceptance Criteria Q U g Design Commitment inspections, Tests, Analyses Acceptance Criteria 3 } Safety System Logic and Control 9 1. The equipment comprising SSLC is 1. Inspections of the as-built SSLC will be 1. The as-built SSLC conforms with the j defined in Section 3.4(Al. The equipment conducted. description in Section 3.4(Al. Diverse comprising diverse backup support backup support equipment for SSLC functions for SSLC is defined in Section conforms with the description in Section 3.4 (C). 3.4 (C). 2. Safety-related monitoring and trip logic 2. Tests will be performed on as-installed 2. A test report exists which concludes that for the plant protection systems resides in SSLC using simulated input signals. the SSLC design basis performance SSLC equipment. SSLC integrates the System outputs will be monitored to requirements are met. automatic decision-making and trip logic determine operability of safety-related functions and manual operator initiation functions. functions associated with the safety M actions of the safety-related systems. SSLC generates the protective function 0 signals that activate reactor trip and ? provide safety-related mitigation of reactor accidents. 3. The DTM, TLU, and OLUs for flPS and 3. 3. MSIV in each of the four instrumentation Tests w.ll be performed on SSLC by a. The test signal exists only in the Class a. i divisions are powered from their pr viding a test signal to the I&C 1E division under test in SSLC. respective divisional Class 1E AC sources. The DJMs and SLUs for ESF 1 and ESF 2 d n at tim in Diihstons I, ll, and ill are powered from thed respective divisional Class 1E DC b. Inspection of the as-installed Class 1E b. In SSLC, physical separation or s nces, in SSLC, independence is divisions in SSLC will be performed. electrical isolation exists between p ovided between Class 1E divisions and Class 1E divisions. Physical b 15 con Class 1E divisions and non-Class separation or electricalisolation exists i 1E e juipment, between these Class 1E divisions and R non-Class 1E equipment. }.

== N 7 E S ii- ~ b l % b TNir w LOE w

a RADIATION PROTECTION COMMENTS ON GE ABWR TIER 1 REV. I 1. See markups attached.

• e 25AS447 Rev. 0 ABWR oesin ertmeation Materist i

c ExcInn Area Baha'y : j /'37# p-3 p%y.Au a cAk b & ^ iset/ W $ 300'"g (M A a m a m w & c &- mp3 g ,fteAAwf able 5.0 ABWR Site Parameters l N N,g.g' d %>Azust bum Ground Water Level: Extreme Wind: Basic hind Speed: 61.0 cm below grade 177 km/hrm/197 km/hr* Maximum Dood (or Tsunami) level: Tornado 30.5 cm below grade

• Max 2 mum tornado wind speed:

4S3 km/hr 2

• Maximum pressure drop:

0.141 kg/cm d Precipitation (for Roof Design):

• Missile spectra:

Spectrum I* j

• Maximum rainfall rate:

49.3 cm/hr* 2

• Maximum snow load:

0.024 kg/cm Ambient Design Temperature: Soil Properties: 17c Exceedance Values

• Minimum static bearing em
• Maximum:

37.8'C dry bulb capacity-7.32 kg/cm 25*C wet bulb (coincident)

• Minimum shear wave velocity:

305 m/sec* 26.6*C wet bu!b (non-coincident)

• Liquefaction potential:

None at plant site f a Mmimum- -23.3*C resulting from site Or Exceedance Yajues (Historical L2mit) specific SSE ground c

• Max: mum:

46.1*C dry bulb motion 267C wet bulb (coincident) 27.2'C wet bulb (non coincident) Seistnologn

• Mmimum:

-40'C

• SSE response spectra: See Figures 5.0a and 5.0b*

i Meteorological Dispersion (Chi /Q):

• Maximum 2 hour 95Fc E\\B 1.37 x 10-3 sec/m3
• Maximum 2-hour 95'1 LPZ 4.11 x 10" sec/m3 l
• Maximum annual average 4

3 (6760) hour) LPZ 1.17 x 10 sec/m i (1)50-year recurrence interval; value to be stihred for design of non-safety-related structures only. (2)100-year recurrence interval; value to be utilized for design for safety-related structures only. 2 (3) Maximum value for 1 hour over 2.6 km probable maximum precipitation (PMP) with ratio of 5 minutes to 1 hour PMP of 0.32. Maximum short term rate: 15.7cm/5 min. (4) Spectrum i missiles consist of a massive high kinetic energy missile which deforms on impact, a rigid missile to test penetration resistance. and a small rigid missile of a size sufficient to just pass through any openmgs in protective barriers.These missiles consists of an 1800 kg automobile, a 125 kg. 20 cm diameter armor piercing artillery shell, and a 2.54 cm diameter solid steel sphere, all impacting at 35% of the maximum borizontal windspeed of the design basis tornado.The first two missiles are assumed to impact at normal incidence, the fart to impinge upon barrier openings in the most damaging directions. (5) At founcation level of the reactor and control buildi.ngs. (6)This is the minimum shear wave velocity at low stra ns after the soil property uncertainties have been apphed. (7) Free field, at plant grade etevation. 5 O2 Srte Parameters

i %hth5 h0Q 3

• h.E=

- 32.. G

d. a ai 0

f 0 s e n 2 o k em it F t le g Rt e n 2 s c o R u n s u a c a t n F d t o o aaC n b in agd h ct e e cl a er r Ct i a r t t eb eh n0 e e 6 r p i 0 r ivoda ni s n s apl l io t i f 1 u n eae a 1 e f o e oal ino iui r n e in qi d r a meo i n r e vot t u la ai h cnn ir t ot t e ut m s oqe a o ot e e r r i qcpopa b r oh ,nmpd inr ) r ,b d e e el it st ia i t r s r s )t o ct ncl i ma ep r r la l o c b r s ue - el s t nt a n r a r e n d. enb n s a ci e on aad ois e. mhhi o eo ee c ar eol iu n s t d r t nah ni l h s s ni i vmi.cl i s t e o t r r e pi apabet qr v (a a s t a i C nea r d t ( i o emi a c e ut wgvnw f l h pi gi l t l t u e ad cn ee oit im ef t oumiq n eki i c sd ne cn iol c a inh wf l t r mlo sit a cir s c t t s o e t a en l ol a d a pt ns n a aa s o scl ur nond ab da af un m yisd o eo e nn os p r s lly ah e o i o e )s o. h h t t s t t n o aid sib ct ai i 0 i h t t ah r ht y sh t t s un nr e f t t n aiea x at a xc a e ca o mf gn c o t r lcl r i m (s e iogr t t i n mt lpme e r c n i t ) ur d t d s A ni a n n o mt n nan a oa,nna t oi r r t me no n s ioa o e lai n r e o s pi ce e p o n si ) ia r o c r r i%np e rh mt or eit uc mcnpu t t a u c eqd c e nd0 a n r r c und n e uonb h c r lu e or e noa1 op oc qe eoh opc o ooit r c c F ar acr cA F aevr ct c A o F (s ( r cao n la o Cc a. b c g n a i i 1 r r o e t t y i i b s d df r s n C n n id 0 eo e do a h u 3t o e ei c c r ( s p n r, a o M n nl r r d e : e f f l c et t d c juu a aeol a h e edd r a r e r s e uer i d cr e c s l t s am/aiur n p e cpur n o it o o r a o cr ni f a bl e u i yb a r e s coos s ih s r led w e o ly ia n ,r o c ct ocs n e r c in s u c s t a e ea aCan b A n obl a lee c a t f al p r c e m pir t ad r n d A l pnc ll w an a0 p yn i sl r A n n al oiga Wd r e 0 oa a a s aib nh o a i t cr e eon 0, p ecn oh r ma uis lf t h d t s s s it o n e e a l o pnqn io c cp0pdi r a et iu s n r t r 0 a lao a s T s ea u u r l t ir n 1 y o e e o gc ) aod mr it a lo odf yc r s n e n t r a s eoai a r n et o n epa la l k cgo o c ieu i a ns e t i t A i nmf n a e oc. tn e o t t c o e aur c l e ibadod s t l a e c ed pqal e l m i ta i et l s r a v a mstedt c i i iegC laa. cp d eaua t vl i au t s p di c ic i r vd n e s et ut h in t e pu at gn io c V in e T tc an nct s r s eoo aa as or yq yf f i r t T a T e As omfo cr ,s ii er e ap n pdd d e r c b io aa n o pc b. x E r r af oa a 2 t c c 3 e 1 e p s l b n l s ss i a l T i ee w d c i c em lc a v e ulne it t ay io n c s n r ios ind o tn od na a s r e f ot r a m er it ine e np d elaa t r m_ h i ivnimot t or n bi o or w m r e pb t iat veef o l r n C l ia oe a h o n.hf t g s o is is t s di ns is nt nvon e n ige aoio t D e ac mls pr se t s a nnl n i i e. d ia W la r s t e n e nt t c vd e nt nnl e a o ot e P cm can 1. [s,y l3y3 ht l

2SAS447 thv. O ABWR oesign certiscation steriar (2) Senice Building HVAC System. Technical Support Center HVAC System The 1 echnical Support Center (TSC) HVAC stem provides a controlled emironment for personnel comfort and safety in the TS. The system consists of an air conditioning unit with two supply fans, two exhaust fans, and an emergency filtration unit with two circulating fans. The emergency filtration unit will have at least 95% removal efficiency for all forms ofiodine (elemental, organic, particulate, and hydrogen iodide) from the influent system. Toxic gas monitors may be required in the outside air intake of the TSC HVAC System; these sensors are not in the Certified Design. The TSC HVAC System is classified as non-safety-related. The TSC HVAC System is located in the Senice Building. On receipt of a signal for high radiation in the normal air intake for the TSC ventilation system, the normal air intake damper closes, the minimum outside air intake damper opens and the ventilation air for the TSC is routed through the emergency filtration unit. In the high radiation mode, a positive pressure is maintained in the TSC relative to the outside atmosphere. Interface Requirements Toxic gas monitors will be located in the outside air intakes of the TSC HVAC System, if the site is adjacent to toxic gas sources with the potential for releases of significance to plant operating personnel in the TSC. These monitors shall have the following requirements: I (1) Be located in the outside air intake of the TSC HVAC System. i (2) Be capable of detecting toxic gas concentrations at which personnel protective i acdons must be initiated. Service Building HVAC System The Senice Building HVAC System senes the Operational Support Center (OSC) and the rest of Senice Building, excluding the TSC, and it consists of an air conditioning unit, supply fan, and two exhaust fans. Inspections, Tests, Analyses and Acceptance Criteria For portions of the CRHA HVAC System within the Certified Design, Table 2.15.5a provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CRHA HVAC Systems. Herting. Venti: sting and Air Conditioning Systems 2.15.5-11 ,}}