ML20128B428

Forwards Requests for Info for Simplified BWR Design Application

ML20128B428
Person / Time
Site:	05200004
Issue date:	01/28/1993
From:	Jacqueline Thompson Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9302030051
Download: ML20128B428 (45)

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f _ Docket No.52-004! January l2.8L 1993

• U Mr.' Patrick W.cHarriott, Manager .

' Licensing & Consulting. Services.

GE Nuclear, Energy.'

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175 Curtner Avenue San Jose, California 95125- -

Dear Mr. Marriott:

'SUCJECT: ' TRANSMITTAL OF REQUESTS FOR ADDITIONAL INFORMATION (RAls) FOR THE SIMPLIFIED BOILING. WATER REACTOR (SBWR) DESIGN-In our. letter!to you dated December 10,-1992, the staff completedcits __

acceptance review of the SBWR application and found'..that the application'was in:omplete and in_ some areas deficient. Some of these deficient- areas were t also listed in our December 10, 4692,-letter. These deficient areas will need to be addressed before the staff can complete its -review.. Tlie staff.is aware of GE's plans to address some of the:e deficient areas with a supplement to their initial SBWR applicatio_n by February 28, 1993.' GE must submit its-complete application and all subsequent amendments-under oath ~or_ affirmation, -

in accordance with the~ requirements of 10 CFR 52.45(d).

In order to facilitate the SBWR review process, the staf_f has completed an-initial review of the SBWR standard safety analysis report' (SSAR) and has prepared a nuraber of round zero RAls, where there was sufficient information ,

presented. These RAls are provided in the enclosure. ~ As previously mentioned.-

in our earlier letter, the enclosed RAls are not intended to complete the RAl=

stage of the review. Our intent is _to provide GE questions; regarding initial staff concerns in order to facilitate the review process.

This requirement affects nine or fewer respondents and;- therefore, is.nat

- subject to Office Management and Budget review under P.L.96-511. -If 'you have-any questions or comments concerning'this matter, you can contact me at-  ;

(301) 504-1113 cr Melinda Malloy at (301)' 504-1178.

. Sincerely, (Original signed by).

John W. Thompson, Acting Project Manager Standardization Project Directorate.

Associate Director for-Advanced-Reactors and License Renewal .

Office of Nuclear Reactor. Regulation *

Enclosure:

SBWR Acceptance Review Round Zero RAls

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See next.page 010M 2.

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Docket file PDST R/f JThompson MMalloy PDR PShea R11T.RlfLV110N wIo_encloiur.e:

DCrutchfield Wiravers CPoslusny JRichardson, 7D26 AThadani, BE2 JNWilson RBorchardt THasselberg TKenyon TWambach MFranovich GBagchi, 7H15 JMoore, 15B18 GMizuno, 15B18 JSniezek, EDO ACRS (11)

GZech, 10A19 GHolahan, BE2 JRichardson, 7026 BBoger, 10HS RJones, BE23 RBarrett, 8D1 RPerch, 8H7 CBerlinger. 7E4 JWermiel, 10D24 TMurley/FMiraglia, 12G18 GGrant, EDO RGramm, 9Al CMcCracken, 8D1

Mr. Patrick W. Harriott Dockot No.52-004 General Electric Company cc: Mr. Laurence S. Gifford GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 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 Mr. Daniel F. Giessing U. S. Department of Energy-HE-42 Washington, D.C. 20585 Mr. Jeffrey C. Baechler GE Nuclear Energy 175 Curtner Avenue, MC-782 San Jose, California 95123 Mr. Frank A. Ross Program Manager, ALWR Office of LWR Safety & Technology U.S. Department of Energy NE-42 19901 Germantown Road Germantown, Maryland 20874

SIMPLIFIED BOILING WATER REACTOR (SBWR) ACCEPTANCE REVIEW

. ROUND ZERO REQUEST FOR ADDITIONAL INFORMATION (RAl) QUESTIONS i RAI NUMBER REQE11 0T58.0 The staff has completed its initial acceptance review of the technical specification (TS) selection criteria and content for the FBWR application, as presented in Section 16.0 of the Standard Safety Analysis Report (SSAR). Based on the staff's review, 1 further clarification is needed for the following OTSB RAls.

OTSB.1 The SSAR states that in accordance witF the criteria of the Commission's " Policy Statement on Technical Specification Improvements," limiting conditions for operation (LCOs) are provided. In order to verify this, a complete set of LCOs for the l SBWR's passive systems is neeried (as well as providing schedules  ;

for those systems designated as to be determined (TBD)). -

1 OTSB.2 GE needs to identify the specific differences in the proposed TS requirements from those contained in Rev. O of NUREGs 1433 and 1434.

OTSB.3 GE needs to provide specific justification for the changes to the completion times and surveillance intervals in accordance with the basis for the staff's evaluation of related topical reports.

ECGB.O SSAR Sections 2 and 3 RAls are listed in the following ECGB.1-20 questions.  !

ECGB.1 Provide the justification as to why no informt. tion was provided for the design features described in SSAR Sections 2.5.5,

" Stability of Slopes," and 2.5.6,

• Embankment and Dams."

ECGB.2 In SSAR Table 3.2-1, item B21.6, the feedwater line classification break from Quality Group (QG)-B to QG D is located at the seismic interface restraint. In Figure 21.5.1-1, this break is located at the shut-off valve. The staff's position is that this classifi-cation break should be at the seismic restraint. Revise Fig-ure 21.5.1-1 to agree with Table 3.2-1.

ECGB.3 SSAR Section 6.7 states that the SBWR alternate to a main steam isolation valve leakage control system is contained in Appen-dix 19H. Since this Appendix will not be submitted until Febru-ary 28. 1993, the staff cannot complete its review of this issue.

Hewevttr, Table 3.2-1 appears to contain acceptable commitments to the staff positions relative to the structural integrity of piping systems and components applicable to this issue with the exception of the following:

a. If the proposed alternate leakage path contains both the main steam drain lines and the turbine by-pass lines, Items B21.13 and N37 in Table 3.2-1 should contain a commitment that these

• lines will be dynamically analyzed for the safe shutdown -

earthquake (SSE) up to the condenser. This same commitment should be added to Section 10.4.4 for the turbine by-pass lines. ,

Enclosure

'l i

b. Table 3.2-1, Item N61 and Subsection 10.4.1 should both  :

contain a commitment that the condenser anchorage is dynami-  ;

cally analyzed for the SSE. )

ECGB.4 InSSARTable3.2-1,ItemC61,RemoteShutdownSystem(RSS),is  :

classified as not safety-related and non-seismic. It is stated in '

this table that the RSS controls some components that are in the "

control rod drive (CRD), reactor water cleanup (RWCU)/ shutdown cooling (SDC), reactor component cooling water system (RCCWS), and heating, ventilation, and air conditioning (HVAC). Subsec- ,

tion 7.4.2 in the SSAR states that the RSS does not include control interfaces with safety-related equipment. In the .

advanced boiling water reactor (ABWR), the RSS is Safety Class 3, quality assurance (QA) B, and seismic Category 1. Since some of the components controlled by the RSS in the SBWR may be safety-related, provide the basis for the non-safety and non-seismic .

classifications for the RSS. '

ECGB.5 In SSAR Table 3.2-1, Items E50.2 and E50.3, the piping and valves (including su) ports) in the gravity driven cooling system (GDCS),

from the chec( valves upstreaa of the squib valves to the suppres-sion and GDCS pools and from the GDCS pools to the lower-drywell are QG C. According to Section 6.3 in the SSAR, the GDCS is considered to be one of the SBWR emergency core cooling systems.

Therefore, in accordance with Regulatory Position C.I.a of Regula-tory Guide (RG) 1.26, this portion of the GDCS should be classi-fied as QG B. Either revise Items E50.2 and E50.3 in Table 3.2.1 and applicable portions of Figure 21.5.3-2 in the SSAR-to agree with the staff position, or provide the basis for the QG C classi-fication.

ECGB.6 In SSAR Table 3.2-1, item G21.4 of the fuel and auxiliary pools cooling system, all of the piping and valves between inboard containment isolation valves and their termination points inside containment are classified as non-safety, QG D, no quality assur-ance requirement, and seismic Category 11. Some of these classi-fications are not totally consistent with applicable portions of the ABWR. However, the discussions in Subsections. 6.2.1.1,

" Pressure Suppression Containment," and 6.2.2, " Passive Contain ,

ment Cooling System," (PCCS) imply that the PCCS performs the safety-related functions of some of those systems listed in item G21.4. Provide a more detailed discussion of the bases for the classifications in item G21.4.

ECGB.7 In SSAR Table 3.2-1, Item G21.6, piping and valves between the low-pressure coolant injection (LPCI) gate valve (F332 on Fig-ure 21.9.1-1, Sh. 2) and the interface with the RWCU/ shutdown cooling system (SCS) is shown as non-safety-related and no QG i classification. On Figure 21.9.1-1, this portion of piping is shown as safety-related, QG B (8"-FD-B) and it connects to an 8"-

FD-B line in the RWCU/SDC system (Ref. Figure 21.5.4-2. Sh. 2).

Revise item G21.6 to agree with the classifications in Fig-ure 21.9.1-1.

ECCB.8 In SSAR Table 3.2-1, items K and U74, Radioactive Waste Management Systems and Radwaste Building Structure, commitments are made that 1 a quality assurance program meeting the guidance of RG 1.143 is J applied to all of the non-safety items in these systems and i structures. In addition, commitments to Section 5. " Seismic Design for Radwaste Management Systems and Structures Housing Radwaste Management Systems," in RG 1.143 should be made in this table for both items K and U74. Since the SBWR does not include the OBE as a design requirement, provide the seismic design criteria that will be implemented to conform to Section 5 of RG 1.143.

ECGB.9 In SSAR Table 3.2-1, Jtem V73, the stack is classified as non-safety and non-seismic. In Figure 21.1.2-2, Sh. 2, the stack appears to be a part of the reactor building outer shel1. In _

Section 3.8.4.1, the reactor building outer shell is identified as-seismic Category 1. Either revise Item U73 to classify the stack as safety-related and seismic Category I, or provide the basis for -

the non-safety and non-seismic classifications.

ECGB.10 In SSAR Subsection 3.6.2.1.1, in the paragraph on page 3.6-13 entitled " Hon-ASME Class Piping," add the following commitment to be consistent with Standard Review Plan (SRP) 3.6.2: " Separation and interaction requirements between seismically analyzed and non-

. seismically analyzed piping are met as described in Subsec-tion 3.7.3.8."

ECGB ll In SSAR Subsection 3.6.2.4, it is stated that the SBWR does not require guard pipes. Subsection 3.6.2.1.1 provides criteria for

" sleeve assemblies" in the containment penetration areas, and Subsections 5.4.6.3 and 6.2.4.3.2, mention the use of guard pipes #

in the containment penetrations for the steam supply and conden-sate lines of the isolation condenser system, in Section 5.4.6.3, it is stated that the design intent for these guard pipes is either to show that the stresses and fatigue usage factors do not exceed special limits in SRP 3.6.2, or to show by proof testing that the guard pipes and transition fittings do not experience crack initiation or crack growth.

! a. Revise Subsections 3.6.2.1.1 and 3.6.2.4 to clarify that the criteria for sleeves in Subsection 3.6.2.1.1 are applicable to guard pipes in all containment penetration areas,

b. It is the staff's position that the experimental analysis proof testing briefly discussed in Subsection 5.4.6.3 cannot.

be used in lieu of the criteria in SRP 3.6.?.

ECGB.12 In SSAR Section 3.7.3, it is stated that for seismic subsystem

, analysis of ASME components, ASME Section 111, Appendix N, " Dyna-l mic Analysis Methods," is applicable. Appendix N is a non-mandatory appendix that is still evolving and does not currently agree with some staff positions. Therefore, it has not been

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i endorsed by the staff, and the staff has no immediate plans to review it. Some of the rules which are either in Appendix N, or are being proposed for future addenda to that standard, and which do not agree with staff positions, address issues such as damping .

values, use of the load coefficient method, use of the independent-support motion response s >ectrum method of analysis, and the 50-percent nonexceedance pro) ability level in N-l?23.2, N-1724, and .

N-1725 of Appendix N. Revise the SSAR to delete all references to Appendix N and replace them with applicable RGs, SRPs, or staff approved ASME Code Cases. j ECGB.13 In SSAR Subsection 3.7.3.1, Electric Power Research Institute (EPRI) NP-6628 , " Procedure for Seismic Evaluation and Design of Small(NClG-14)ing,"

Bore Pip is referenced as an alte.oate procedure to be used in lieu of seismic-analysis for_ piping 2 inches and smaller in diameter. This procedure incorporates, in part, the use of a seismic experience-based approach for the ,

design or qualification of small bore safety-related piping. The staff has not accepted this procedure. Currently, the staff only accepts a suitable dynamic analysis or a suitable qualification test except when the use of an equivalent-static analysis has been demonstrated to be adequate for the design of such piping systems.

Revise Subsections 3.7.3.1 and 3.7.6 to delete the reference to EPRI NP-6628 (NCIG-14).

ECGB.14 During a July 17, 1991, meeting with GE, the staff was informed that the SBWR is mainly based on state-of-the-art light water reactor (LWR) technology of operating boiling water reactors  :

(BWRs) and ABWRs. Thus, only a. limited amount of development testing for certification of the GDCS, PCCS and depressurization valve (DPV) is needed. Among those planned tests, only laboratory tests of component or scaled model were performed. Some planned  ;

full-scale tests for-decay heat removal, mechanical- performance of-heat exchanger, isolation condenser, and their performance under integral system conditions will not be conducted until 1995. For those tests already performed, large uncertainties in heat trans-fer were reported. It appears that all thermal loads needed for design of mechanical components and piping systems under passive operating and accident conditions are not yet been accurately defined. At this stage of the SBWR design, how does GE ensure-that the design loadings for mechanical.compo ents are adequate 7--

ECGB.15 During the July 17, 1991, meeting, the' staff-was also informed-that for predicting response of the SBWR internals to loss-of- g coolant accident. (LOCA) using the GDCS, a 1/508 sector-scaled 3BWR

• test was performed, such that the data base can be used to qualify the thermal-hydraulic computer codes (TRAC and TRACG) for- accident ,

analysis. However, for normal-operation, since _the flow rate in the SBWR core area is dependent- upon natural--circulation,-the flow- -

velocities may vary in a broad range under different operating conditions. In cases of low flow rate, thermal mixing.inside the reactor vessel may not be tho' rough,.and the flow may._be stratified l

l m _ __ _ . _ _ _ . _ _ _._ _ _ _ _ _ _._ _ _. . .

into several regions with different thermal conditions. In such cases, reactor internal components may experience uneven thermal loads and result in high thermal stresses and high cumulative fatigue effects. Since the thermal loads are difficult to be accurately-predicted analytica11y due to complexity of flow-pass geometries and complicated boundary conditions, an instrumented full-scale prototype testing of reactor internals under various operating transients appears necessary for. confirming the thermal loads for the reactor internal component design. Discuss if such a test is planned, or if none is planned, why it is not necessary.

ECGB.16 In SSAR Subsection 3.9.3.4 and Table 3.2-1, portions of the isolation condenser system (ICS) outside containment. are identi-fied as being classified as ASME Class 2 (QG B). The staff agrees that this classification is consistent with the exceptions allowed in 10 CFR 50.55a(c)(2). However, because portions of the ICS are

• subjected to reactor operating temperature and pressure continu-ously during operation, and because this system is used to trans-fer decay and residual heat from the reactor after it has been shutdown and isolated, it appears that some of the Class 2 steam-supply and condensate piping and the condensers could be subjected to a significant number of thermal transients. Provide a descrip-tion of the methodology to be used to account for the effects of fatigue in this Class 2 portion of the ICS. In addition, in '

Figure 21.5.4-1, identify the exact-location of the safety classi-fication change from Class 1 to Class 2 in both the steam supply ,

and condensate return piping.

ECGB.17 In SSAR Subsection 3.9.3.7.1, it is stated that to minimize the use of snubbers, special engineered pipe supports such as energy absorbers and limit stops may be used,

a. With respect to energy absorbers, it should be noted in this subsection that (1) ASME Code Case N-420 can only be used as conditioned by RG 1.84, and (2) ASME Code Case N-420 cannot be used in the same analysis that uses the damping values in ASME Code Case N-411. Revise Subsection 3.9.3.7.1 and any other applicable subsection in the SSAR to add these conditions.

b. The use of limit stops is currently being reviewed by the staff on a plant-specific basis. One plant has been condi-tionally approved to use this alternative in a part of o n pilot piping system. Pending the results of the staff's evaluation of this program, the use of limit stops is not acceptable. Revise Subsection 3.9.3.7.1 and any other appli-cable subsection to either delete the paragraph on limit stops ~

or commit to using this alternative only after it has been approved by the staff.

ECGB.18 The information in SSAR Section 3.9.6 infers that exemptions from the Code testing requirements may be requested.

1 All of the plants which have been licensed by NRC have been permitted to request relief from the ASME Section XI inservice t testing (IST) rules for pumps and valves. These pumps and valves are generally installed in systems in which it is impractical to meet the Section XI rules because of limitations in the system design which preclude testing without significant design changes.

In other cases, the staff approved alternatives to the Section XI requirements because imposition of the Section XI rules would have resulted in hardships to the licensee without a compensating increase in the level of safety. The underlying reason for the regulation allowing these reliefs from the code was that the detailed system designs for all of these plants were essentially completed prior to the time that the staff promulgated 10 CFR S0.55a(g) that incorporated by reference the ASME Code Section XI rules. A 31 ant such as SBWR, for which the final design is not complete, las sufficient lead time available to include provisions for this type of testing in the detailed design of applicable .

piping systems. Therefore, exemptions from the applicable code testing requirements will not be granted for SBWR. However, with regard to subsequent or future code revisions to the applicable ASME Code for the SBWR plant, requests for relief from certain updated code requirements may still be submitted for staff review in accordance with 10 CFR 50.55a(f). Revise SSAR Subsection 3.9.6 to provide a more explicit commitment that SBWR will be designed to accommodate testing per the code requirements for IST of valves (and pumps, if applicable). ,

ECGB.19 In SSAR Section 3.9.6 GE stated that safety-related pumps and valves will be included in the IST program for the SBWR. The unique SBWR design places significant reliance on passive safety systems, but also depend on non-safety systems (which are tradi-tional safety systems in current LWRs) to prevent challenges to passive systems. Iherefore, it is very important that testability of both safety-related valves and important non-safety pumps and valves be provided early in the design phase. The applicant is requested to provide detailed information to ensure that all safety-related valves can be in situ tested to demonstrate their design capabilities and to monitor their condition.

The staff has not completed it's review of the extent to which important non-safety components may have to meet safety-grade .

criteria. However, there are uncertainties concerning the lack of a proven operational performance history for the valves in the passive systems. These uncertainties may increase the need to rely on the important non-safety systems _and components in provid-ing the defense-in-depth to prevent and mitigate accidents and

, core damage. The staff is still evaluating this issue for the i passive plant designs. The specific staf.f positions on the i inservice testing requirements for the important_non-safety components will be determined when the staff completes its review

of the issue of regulatory treatment of non-safety systems. The applicant will then be requested to revise Section 3.9.6 te agree ,

with the staff's position.

ECGB.20 In SSAR Section 3.9.6.2 of the SSAR, GE stated that the motor. ,

operated valves (MOVs) equipment specifications require the incorporation of the results of either in situ or prototype testing with full fluw and pressure and/or full differential pressure to verify the proper sizing and correct switch settings of the valves. In Section 3.9.7.3 of the SSAR, GE also stated that the concerns and issues identified in Generic Letter '

(GL) 89-10 for MOVs will be addressed by the a>plicant referencing the SBWR design before plant startup. The motlod of assessing the '

loads, the method of sizing the actuator, and the setting of-torque and limit switches will be specifically addressed. How- ,

ever, the staff has determined that all the concerns .nd issues identified in GL 89-10 and its supplements that re b te to tests. -

analyses, and acceptance criteria to determine the adequacy of valve design and to ensure the ability of MOVs to meet functional performance requirements under all design basis conditions, including recovery from inadvertent valve mispositioning, must be .

addressed to demonstrate the design basis capability of MOVs. The staff has also determined that this issue should be addressed under a generic inspections, tests, analyses, and acceptance criteria (ITAAC) rather than a combined license (COL) action item.

GE should develop an acceptable generic-!TAAC for demonstrating MOV capability, as discussed above.

ECGB.21 in the SSAR, GE has committed that the MOV equipment specifica-tions will require the incorporation of the results of either in situ or prototype testing with full flow and differential pressure to verify the proper sizing and switch settings of the valves. GE also committed that all SBWR safety-related piping systems will incorporate provisions for testing to demonstrate the operability of check valves under design basis conditions.

Based on operating experience, the staff has determined that a similar connitment is needed for the s)ecifications for other power-operated valves to incorporate tle results of either in plant or prototype testing to verify design basis capability.

Based on past experience with estimating thrust and torque requirements and other parameters for valve operation, the staff believes that this assurance cannot be provided by analytical approaches alone and will require that proper sizing and adjust-ment of other power-operated valves be verified by a generic ITAAC. GE should develop an acceptable generic ITAAC for demon-strating the capability of other power-operated valves.

ECGB.22 Several piping systems connected to the reactor coolant pressure boundary have design pressurcs below the rated reactor coolant system (RCS) pressure. Also some systems that are rated at full reactor pressure on the discharge sido of pumps have pump suction

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t pressure below RCS pressure. To protect these systems or portions of systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high-pressure RCS and the low-pressure system. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low-pressure systems.

In SSAR Section 3.9.6, GE stated that the periodic leak testing of ,

the RCS pressure isolation valves (PlV) in Table 3.9.8 will be i performed in accordance with Chapter 16 surveiliance requirement (SR) 3.6.1.5.10. The referenced SR appears to be incorrectly identified and the correct section should be SR 3.4.3.1.-

SR 3.4.3.1 states that the RCS PlV leak testing frequency will be in accordance with inservice testing program or once per refueling interval.

However, it should be noted that the above-referenced inservice testing 3rogram (SSAR Table 3.9.8) will not be submitted by GE until Fesruary 28, 1993. Therefore, the staff's review of this i issue cannot be completed at this time. .GE is requested to provide a list of RCS PlVs. Moreover, the staff has determined <

that the leak testing frequency as stated in SR 3.4.3.1 is not fully acceptable for SBWR, GE is requested to address other leak testing frequencies that are contained in several of the standard TS and currently implemented by many operating plants. Those frequencies include leak testing prior to entering Mode 2 whenever ,

the unit has been in Mode 5 for 7 days or more, if leak testing has not been performed in the previous 9 months, and leak testing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve.

ECGB.23 In SSAR Section 3.9.6, GE stated that IST of safety-related pumps and valves will be performed in accordance with the requirements of ASME OM Code 1990, Subsections ISTB, ISTC, and A)pendix I, it should be noted that Subsections ISTB and ISTC of tie ASME OM Code 1990 are essentially the same as OM Standards Part 6, "Inser-vice Testing of Pumps," and Part 10, " Inservice lesting of Valves," respectively. However, OM Standards Part'6 and Part 10 are referenced in Section XI of the 1988 Addenda and 1989 Edition':

The 1988 Addenda and the 1989 Edition of Section XI have been incorporated by reference into 10 CFR 50.55a and are acceptable for tie passive LWR IST provided the analysis of leakage rates and corrective action requirements of Paragraphs 4.2.2.3(e) and 4.2.2.3(f) of Part 10 are applied to containment isolation valve

~

testing. Therefore, Section 3.9.6 should be revised to refer the 1988 Addenda and 1989 Edition.

ECGB.24 Regarding SSAR Section 3.7, " Seismic Design," the seismic Category I systems and components are designed to remain functional for earthquake loadings. Provide the basis for why this section of the SSAR does not address the structural integrity l of the systems and components.

ECGB.25 SSAR Section 3.7 states that the exhaust stack is classified as non-safety-related. Provide the basis for how postulated failures of this structure would not affect the function or integrity of any safety-related component or structure. j ECGB.26 Regarding figures 3.7.1 and 3.7.2 in SSAR Seci. ion 3.7.1.1.2,

" Design Time History," show the design response spectra for damping ratios of 2, 5, 7, and 10 percent. However, Figures 3.7.6 through 3.7.17 show the spectra enveloping for damping ratios of 2, 3, 4, and 7 percent. What is the basis for not showing that Figures 3.7.1 and 3.7.2 should reflect 3 and 4 percent damping response spectra values and 5- and 10-percent damping ratios for Figares 3.7.6 through 3.7.177 The staff believes it is not acceptable for GE tu use the 5- and 10-percent damping ratios in the analysis and design of structures, systems, and components, if the design time history cannot satisfy the enveloping criteria for these two damping ratios. Also, please show (or provide the basis i for not including) the power spectrum density function er.veloping condition for the vertical time history. ,

ECGB.27 Regarding SSAR Section 3.7.1.2, " Percentage of Critical Damping i Values," what is the basis for not showing or listing the damping values for the electrical components such as cable trays, conduit, .

heating, ventilation, and air conditioning, etc?

ECGB.28 Regarding SSAR Section 3.7.2.1.1, " Time History Method," what is the definition for the term " highest frequency (or shortest period) of significant," and why is this not defined in the SSAR7 ECGB.29 What is the basis for the following statement contained in SSAR Section 3.7.2.1.17 "For the frequency domain solution, the frequency interval is selected to accurately define the transfer functions at structural frequencies within the range of significant."

ECGB.30 Regarding SSAR Section 3.7.2.3, " Procedures Used for Analytical Modeling," what is the basis for not including in the seismic analysis the lump mass to the node points and the consideration of the dynamic effects such as water sloshing, etc.?

ECGB.31 Regarding SSAR Section 3.7.2.5, " Development of Floor Response Spectra (FRS)," what is the basis for justifying the acceptability of the direct generation method of the FRS7 ECGB.32 Regarding SSAR Section 3.7.2.7, " Combination of Modal Res)onses,"

please clarify why the combination methods discussed in t11s

, section are also applicable for the modal time history analysis.

ECGB.33 Regarding SSAR Section 3.7.2.14, " Dynamic Stability of Seismic Category I Structures," please provide the basis for not discussing the problem of dynamic instability of seismic Cate-gory I structures due to sliding.

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ECGB.34 Regarding SSAR Section 3.7.3.12. " Seismic Category 1 Buried Piping, Conduits, Conduits and Tunnels," please explain the  :

difference between the seismic Category I features and the seismic Category C-I features.

ECGB.35 Regarding SSAR Section 3.B.6, " Sail-Structure Interaction," please provide the basis for not documenting the validation and quality assurance status of the modified SASSI code which is to be used for the analysis of the SBWR structures.

ECGB.36 Regarding SSAR Section 3.7.4, " Seismic Instrumentation," Subsec-tion 3.7.4.1 states that the number of time-history accelerographs (THAs) contained in the plant will be consistent with the number of 1HAs contained in draft NRC RG DG-1016 (a proposed revision to-RG 1.12). Draft RG DG-1016 suggests that a plant be equip)ed with 8 THAs. SSAR Subsections 3.7.4.2 and 3.7.4.3 state that tie plant will be equipped with 4 triaxial THAs. Please provide the basis for this apparent inconsistency.

ECGB.37 Regarding SSAR Section 3.8, " Design of Seismic Category 1 Struc-tures," please provide the basis for not including detailed ,

structural drawing in the SSAR.

ECGB.38 Regarding SSAR Section 3.8, American National Standards Insti-tute/American Institute of Steel Construction (ANSI /AISC) N690 Specifications and ASCE 4-86 have not yet been accepted by the-staff. Please provide the basis for the commitment to use these standards.

ECGB.39 Regarding SSAR Section 3.8.1.7, " Design and Analysis Procedures,"-  ;

please provide the definition of a " critical location" and the criteria for the selection of " critical locations."

ECGB.40 Regarding SSAR Section 3.8.1.7, " Design and Analysis Procedures,"

please provide the basis for not including detailed procedures for the reinforced concrete containment vessel (RCCV) analysis and design in the SSAR.

ECGB.41 Regarding SSAR Section 3.8.3.1, " Description of the Internal Structures," the inner periphery radius of the diaphragm floor of 7.65m documented in the section is different from the radius of 7.8m, as shown in SSAR Figure 21.1.2-2, Sheet 2. Please explain-this apparent inconsistency.

ECGB.42 _Regarding SSAR Section 3.8.4.3, " Loads," please provide the basis for not addressing the flooding and 'ornado missile loadings in the SSAR.

ECGB.43 Regarding SSAR Section 3.3.1, equation (3.3-1), GE states, "Impor-tance factor I depends on the type of exposure and appro)riate values of- I are listed in SSAR Table 3.3-1." However,111s definition of I is not consistent with that in Reference 3.3-1 L

i which states that importance Factor I is used to adjust the design  !

wind speed to that with annual probabilities of being exceeded i other than the value 0.02 (i.e., 50-year recurrence). This factor converts the wind speed of a 50-year recurrence to either a-25-year or 100-year rt.currence and this does not depend on the type of exposure. Explain this discrepancy in applying this i factor.

ECGB.44 Regarding SSAR Table 3.3-2, this table lists the velocity pressure distribution and gust factors at various heights without providing

. the mean roof heights. Therefore, provide or explain the. follow-ing.

• Provide the mean roof he h ht above grade for the reactor building

. Explain why the windward wall pressure is 0.86Ghqh instead of 0.8Ghqz .

• Provide the calculations for the. values listed in this table and explain how this table is used

. ECGB.45 Regarding SSAR Section 3.3.3, " COL license information," the site-specific design basis tornado part is missing. The site-specific derign basis wind is provided in SSAR Section 2.3.1 (not 2.2.1). ,

EELB.O GE has covered all pertinent aspects of the electrical power system design in the SSAR submittal except for the following <

subject areas which are discussed in RAI Numbers EELB.1-3 listed below.

EELB.1 If not addressed in the forthcoming SSAR Section l'.8, " Interfaces .

for Standard Design," and Section 1.9, "Conformance with Standard Review Plan," then please provide an explanation and how the SBWR incorporates into the design the reliability assurance program (rap) that addresses the TS, inservice inspection / inservice testing programs, the maintenance programs, plant procedures, ar.d the security program (see Commission paper SECY-89-013,." Design Requirements Related to the Evolutionary Advanced Light Water Reactors (ALWR)," for further details).

EELB.2 If not addressed in the forthcomin3 SSAR Section 1.8, " Interfaces '

for Standard Design," and Section'l.9, "Conformance with Standard Review Plan," then please provide an explanation and how the SBWR incorporates into the design the policy issues discussed in SECY-90-016. " Evolutionary Light Water Reactor (LWR) Certification issues and Their Relationship to Current Regulatory !3quirements,"

I and the draft Commission paper dated February 27, 1992.

EELB.3 If not addressed in the forthcoming 5SAR Section 1.8, " Interfaces for Standard Design," and Section 1.9, "Conformance with Standard l Review Plan," then please provide an explanation and how the-'SBWR-11 -

L

---, <---'+-- *v'

t incorporates into the design operational experience (see staff requirementsmemoranda(SRM)datedFebruary15andMarch5,1992).

SRXB.1 SSAR Section 1.2.2.6 Remote Shutdown System. Traditionally  ;

controls for safety relief valves (SRVs) are given in the remote  ;

shutdown panel. But for the SBWR, no contrels are provided for SRVs or automatic DPVs in the remote shutdown panel. Explain why .

no controls are required for SRVs or DPVs. Also, provide the basis for selection of controls and instrumentation to be included t on the remote shutdown panel, SRXB.2 SSAR Section 1.2.2.6 Remote Shutdown System. Electrical power l distribution system is included in the remote shutdown panel. But '

it is not clear whether diesel generators controln are provided in the remote shutdown panel.

OXB.3 SSAR Section 1.2.2.7, Reactor Protection System. Unlike in ,

current boiling water reactors (BWRs), the reactor.is-also scrammed on reactor pressure vessel (RPV) level 8. Is tnere any safety significance for this trip or is this only for the main.

turbine protection?

SRXB.4 SSAR Section 1.2.2.4.1. Reference to reactor water cleanu)/

shutdown cooling system is incorrect because it refer to tie same i

~

section in the SSAR (Section 1.2.2.4.1).

SRXB.S SSAR Section 1.2.2.4.3. Only RPV level 1 and.not drywell high pressure is given as the initiation signal for the GDCS, In current BWRs, RPV low level and drywell high pressure signals are used, providing initiation diversity. Why is only the level signal used for starting the GDCS? Explain in detail why diver-sity is not necessary for the start-up of emergency core cooling systems (ECCS) in the SBWR.

SRXB.6 SSAR Section 1.7. Does this section include drawing standards, piping and instrumentation diagrams (P&lD), P&lD standard symbols, ,

l graphical symbols for use in instrument electrical diagrams

(IEDs), etc., similar to the submittal given in ABWR SSAR Sec-

. tion 1.7?

SRXB.7 SSAR Section 3.1.4.4. General Design Criteria (GDC) 33 requires a ,

system to supply reactor makeup for protection against small breaks in the reactor coolant pressure boundary.- GE takes credit for automatic depressurization (ADS) and integrated control (ICS) >

systems even though they are not water injection systems or make-l up systems to meet GDC 33. The CRD system which can be used for L water injection is not a safety-grade system. GE should explain I

in detail why a safety grade high-pressure core injection system is not necessary to meet GDC 33. .

k

SRXB.8 SSAR Section 4.1, Summary Description, references a non-existing Subsection 1.3.1.1 for a summary of the important design and performance characteristics, some of which are given in SSAR Table 1.3-1 and Tables 4.4-1 and 4.4-2. Please providt a complete summary table as required by the SRP Section 4.1.

SRXB.9 SSAR Subsections 4.1.4.2, fuel Design Analysis, 4.1.4.3, Reactor System Dynamics, 4.1.4.4, fluclear Analysis anti 4.1.4.6, Thermal-Hydraulic Calculations, state that nuclear and thermal-hydraulic analysis techniques and computer codes are " based on" or " adapted" using liRC-approved criteria. Please discuss and provide addi-tional references and/or approved code names to satisfy SRP 4.1 and 4.3.3 requirements.

SRXB.10 SSAR Section 4.2, fuel System Design, states the fuel to be used in the SBWR is "any fuel derign that is based on an f1RC-approved design or meets the criteria documented in Appendix 4B." Refer-ence is made to Amendment 15 to f4EDE-240ll-P-A, which applies to 8x8 and 8x8R operating reactor lattice geometries. Explain why compliance with the referenced acceptance criteria for operating reactor fuel is considered sufficient for the shorter SBWR fuel.

This approach was rejected in favor of a reference fuel and core design for the ABWR. Provide a reference design or explain why the SBWR fuel and core design approach should be different from that approved for ABWR.

SRXB.ll SSAR Section 4.2 states that the control rod design to be used in the SBWR is "any design that is based on an f4RC-approved design or meets the criteria documented in Appendix 4C," " Compliance with these criteria constitutes NRC acceptance and approval of the designs without specific 11RC review" is incorrectly stated.

Please provide further justification for the use of currently approved designs for SBWR applications.

SRXB.12 SSAR Section 4.3 provides an " example" core loading map for a typical equilibrium-cycle core (with a currently approved fuel design (BP8x8R)) which is used for the system dynamic response analyses given in SSAR Section 6.3, Emergency Core Cooling Sys-tems, and Chapter 15, Accident Analyses. Please provide results for a typical initial core and discuss the requirements for transition cycle analysis.

SRXB.13 SSAR Section 4.3, by reference to Appendix 4A, Typical Control Rod Patterns, provides an " example" set of control rod patterns for an equilibrium core at rated power / flow and equilibrium xenon condi-tions along with the associated axial and radial power and expo-sure distributions at 15 cycle exposure steps. Please provide equivalent results for an initial core and discuss the approach used for transition cycle analyses to ensure the design or limit-ing power distributions remain within the design power peaking factor components given in the comparative design SSAR Table 1.3-1 of Chapter 1.

)

_a

t SRXB.14 The relationship of the inferred power distributio'is to the t monitoring instrumentation is not addressed as required by-SRP 4.3.2.2. Please provide a discussion of the procedure used to develop the power distribution and provide justification for use '

of only four fixed axial incore detectors instead of having a -

detailed axial base shape from movable traversing incore probe (TIP)-like detectors. .

SRXB.15 A complete set of reactivity coefficients are not presented as c required by SRP 4.3.2.3. Please provide additional information to supplement the Table 1.3-1 design end-of-cycle Doppler and void coefficient values.

SRXB.16 Control requirements are not provided as rauired by SRP 4.3.2.4 .

except for an example of the all-rods-in anc the strongest rod ,

withdrawn K-effective values at the stated minimum cold shutdown margin condition at the limiting cycle ex)osure for the reference .

equilibrium cycle case. Please provide tie additional information required by the SRP and include initial core results. l SRXB.17 Appendix 4D.- Stability Evaluation states that "the most limiting stability condition in the SBWR normal operating region is at the rated power / flow condition" and that 'the SBWR is designed so that power oscillations are not possible throughout the whole operating region (including plant startup)." Stability performance during .

plant startup conditions is of concern because of the possibility-of a wide range of power / flow / pressure / water level and.subcooling i conditions as well as skewing of axial and radial power distribu-tions due to control rod withdrawal during heatup. The current reactor heatup and pressurization procedure for the Dodewaard natural circulation BWR plant startup is similar to that outlined for an SBWR; however, the actual plant designs differ signift-cantly. Please provide further analysis to evaluate bounding ranges of plant conditions and procedures to justify the assertion that no unstable mode is expected to be encountered during SBWR startups.

SRXB.18 available to GE.

The SSAR indicates thatstates that proprietary the geysering mode canHitachi test data'f be avoided i core inlet subcooling is kept near zero (0 to 9 'F) during plant startu) u)"

, to 2 percent of rated power. Other proprietary' test data, w11c1 has been referenced in open literature by Tokyo tlniversity, suggests contradictory effects from low subcooling. Please provide an evaluation of all available test data, and discuss any.

additional analyses and/or tests planned to resolve this issue.

SRXB.19 Identify the essential portions of the CRD system which are safety related. Describe how the safety-related portions of the ,

system are isolated from the non-essential portion of the system. -

SRXB.20 CRD pumps are used for high pressure make-up of the reactor.

Confirm that the pumps power supply is from the diesel generator bus.

l 4

SRXB.21 Describe the relative core location of control rods sharing a scram accumulator. tan a failure of the scram accumuletor fail to insert adjacent rods? If so, discuss the consequences of that failure.

SRXB.22 The CRD system in conjunction with the rod control _and information i system (RCIS) provides for selected control rod run in (SCRRI) to mitigate the loss of feedwater heating event. Describe in detail how the SCRRI system works.

SRXB.23 Submit detailed drawings of the hydraulic control unit OICV)'and describe in detail the design of the HCOs.

SRXB.24 We understand that the control rod has no velocity limiter.

Discuss in detail the reason for velocity limiter elimination.

SRXB.25 In o>erating BWRs, the ball check valves ensures rod insertion in tio event the accumulator is not charged or the inlet scram valve fails to open if the reactor pressure is above 600 asig.

For ABWR this feature is not provided. Confirm whether t11s feature exists for the SBWR.

SRXB.26 The aerformance of essentially all types of safety / relief valves has seen less than expected for a safety component. Because of reportable events involving malfunctions of these valves on operating reactors, the staff is of the o) inion that-significantly better safety / relief valves performance saould be required of new plants. Provide a detailed description of im)rovements between SBWR SRVs and presently operating 31 ants in tie areas, listed below. In addition, explain why tie noted differences will provide the needed improvenents,

i. Setpoint_ drift and "weepina" are aeneric nrablemi. How will the SBWR SRVs resolve the generic problems.

ii. L11ve and valve operator tyfjtand/_or desian. Include in the- i discussion of improvements in the air actuator especially materials used for components such as diaphragms- and seals.

Discuss the safety margins and confidence levels associated with the air accumulator design. Discuss the capability of the operator to detect low pressure in the accumulator, 111. Specificationi. What new provisions have been employed to ensure that valve and valve actuator specifications include design requirements for operation under expected environmen--

tal conditions (especially temperature, humidity, and vibra-tion)?

iv, letting. Prior to installation, SRVs'should be proof tested under environmental conditions and for time period represen-tative of the most severe operating conditions to which they may be subjected.

i l

v. Quality Assurance. What new programs have been instituted to assure that valves are manufactured to specifications and '

will operate to specifications?

For exam)1e, what tests are performed by the applicant to assure t1at the blowdown capacity is correct?

vi. Valve ooerability. Provide a summary of the surveillance  ;

program to be used to monitor the performance of the SRVs.

vii. VAly_e inipection and overhaul. Operating experience has shown that SRVs failure may be caused by exceeding the manufacturer's recommended service life for the internals of the SRV or air actuator. At what frequency do you intend to require visual inspection and overhaul of the SRVs? For.

both safety relief and ADS, what provisions exist to ensure that valve inspection and overhaul are in accordance with the manufacturer's recommendations and that the desic'1 service life would not be exceeded for any component of the SRV7 SRXB.27 Can the safety / relief valves be closed by operators when these valves are actuated as part of the ADS function? If so, how long after ADS actuation can this be accomplished?

SRXB.28 The isolation condenser system (ICS) is a safety-related system  ;

and the SBWR ICS design-is similar to the ICS -in operating plants '

like Dresden 2 and 3, Millstone 1 Nine Mlle Point I and Oyster Creek. But the operational experience with the ICS in those plants has been of concern to the staff. The staff's experience with operational events relating to the ICS has indicated numerous design deficiencies and several operational problems. Has GE performed a systematic study of the operational experience related to ICS plants? What design changes and improvements have been made to the SBWR ICS design to correct potential design deficien-cies in operating ICS plants?

SRXB.29 The " safety-grade" isolation condenser (IC) calculations assume that the IC pool is saturated during system operation. While this minimizes the temperature difference between the arimary coolant and the IC 3001, it may not minimize the overall 1 eat transfer, ,

due to the ligh efficiency of heat transfer during boiling on the outside of the tubes. Show that the assumptions made result in the minimum heat transfer. If the IC pool is colder and-does not=

boil, can heat removal still be adequately maintained?

SRXB.30 It has been stated that safety analyses do not take credit" for the isolation condenser. While ignoring the presence of the IC as a heat sink may be conservative, it must still be recognized that the component is there and is in communication with the primary.

system. For instance, the presence of the DPVs on the IC-stub lines imply that water can be drawn back through the ICS from.the

cold side of the 3rimary system when the DPVs are actuated, in addition, since tie IC pools are in communication with the PCCS heat exchanger (HX) pools, any pool heatu) caused by IC operation will affect the operation of the PCCS. Slow that there are no system interactions involving the IC that can degrade the plant  :

response during a LOCA.

SRXB.31 The staff is aware of plans to perform tests of a full-scale 10 module in the " PANTHERS" test facility at SIET, Italy. These tests have been determined by the staff to be required as part of r

. the design certification testing. No reference to any of these tests is contained in the SBWR SSAR.  :

The prosisions of 10 CFR 52.47(b)(2) require that the specific-testing supporting the certification of the design must be-descri Jed as part of the application. Furthermore, SECY-91-273,

" Review of Vendors' Test Programs to Support the Design Certifica-tion of Passive Light Water Reactors," requires that the passive plant vendors submit their test program plans, test matrices, and, upon test completion, the qualified raw data to the NRC for review as part of the design certification process. Please provide detailed information on the IC tests at " PANTHERS," as indicated,

. and discuss how the data will be used to support analysis of IC performance in the SBWR. i SRXB.32 One function of the reactor water cleanup system is to prevent thermal stratification in the reactor vessel lower head, if the system stops functioning, a stratified layer of cold water may begin to build up in the vessel lower head. This could have the effect of lowering the overall driving head for natural circula-tion in the primary system.

a. How will stratification affect normal natural circulation flow in the reactor vessel?  :

b. What impact would the stratification have on the operation of the' safety systems, including the ECCS and isolation conders-ers, in the event of a transient or an accident?

SRXB.33 In the " Technical Introduction" volume provided to the staff at the September 3 presentation, the figure of the RWCU system in the

" Safety and Auxiliary Systems" section shows a head spray connec-i tion. However, the slide comparing the RWCU systems in the ABWR and the SBWR in the same section indicates that the head spray does not exist in the SBWR.- Which of these is correct?

SRXB.34 In a presentation to the staff on September 3, 1992, it was stated that the squib valve on the ECCS line between the suppression pool and the reactor vessel was timed to open 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after an ,

accident. In Section 6.3.2.1 of the SSAR, on page 6.3-4, it is-stated that these valves are actuated 30 minutes after an acci-dent. Which of these statements is correct?

SRXB.h The operation of timers is crucial to the actuation of various parts of the ECCS in the SBWR. The automatic depressurization system and the gravity driven cooling system, including both GDCS pool and suppression pool injection, depend on elapsed time signals to accomplish their functions. How many timers are provided for eac1 of these systems, and how are these tin.ars contro1166 ano powered? Verify that nn . ingle failure can disable all essential tiring capability.

SRXB.30 What is the water level inside the drywell during the course of a LC CA? Is there acy sequence involving a Dreak slus a single auNe fatiure that can result in an unremeraale or unisolable icn cf FIAry coolant outside of the containment? This includes both interpstem LOCAs and leakage of water d' ectly from the drywell after an in-containment LOCA.

SRXB.37 In SSAR Table 6.2-2 Generic Assumptions / Initial Conditions for LOCA Analyses, page,6.2-55, the short term analysis assumptions include use of the Moody critical flow model to calculate depress-urization of the reactor vessel. It is known that the Moody model over-predicts considerably the rate of inventory loss from the vessel, and thus the rate of primary system depressurization. In conventional plants, this is generally assumed to be conservative However, in the SBWR, the lack of a high-pressure emergency core .

cooling (Etr.) capability makes reduction of the prinary system pressure essential, and an acceleration of that rate through increased inventory and pressure reduction, permitting earlier actuation of the ADS and low-pressure GDCS, may not be conserva-tive in calculating plant accident response. Demonstrate that the assumptions made in these calculations do indeed produce the most conservative analytical results.

SRXB.38 Passing refcrence to PCCS test programs is made in the last paragraph of SSAR Section 6.2.2.3, page 6.2-25. No specific docu-mentation on the test programs is included in the list of refer-ences; it is inferred that the " GIRAFFE" tests at Toshiba and the planned " PANTHERS" tests at SIET are the test 3rograms mentioned.

The provisions of 10 CFR 52.47(b)(2) require 11at the specific testing supporting the certification of the design must be described as part of the application. Furthermore, SECY-91-273 requires that the passive plant vendors submit their test program plans, test matrices, and, upon test completion, the qualified raw data to the NRC for rev;ew as part of the design certification process. Please provide detailed information on the " GIRAFFE" and

" PANTHERS" programs as indicated, and discuss how the data have been or will be used to support the assertions made regarding PCCS and containment performance. This should include any additional tests that are planned in the recently modified " GIRAFFE" loop.

SRXB.39 The staff f s aware of GE's plans to conduct integral long-term cooling experiments in the " PANDA" facility at the Paul Scherrer Institute (PSI). These tests have been determined by the staff to

Addi-berequiredaspartofthedesigncertificationtestin$s.

tional tests in the "LINX" and AIDA" facilities are a o planned -

at PSI. No ref4rence to any of these tests is contained in the SBWR SSAR. The provisions of 10 CFR 52.47(b)(2) require that the  ;

specific testing supporting the certificatton of the design must be described as part of the application. Furthermore,

• SECY-91-273 requires that the passive plant vendors submit their test program plans, test matrices, and, upon test com)1etion, the ,

qualified raw data to'the NRC for review as part of tie design certification process, please provide detailed information on the-

" PANDA," "LINX," and "AIDA" programs as indicated, and discuss how  :

the data will be used to support analysis of the PCCS in the SBh'R. t SRXB.40 The staff is aware of testing of the CDCS in the GIST facility at GE in San Jose, California. The final report on these tests has beer, previously made available to the staff. However, these tests -

are not referenced in Chapter 6 of the SSAR, nor is any indication- ,

given as to how the results have been used to support analyses of the SBWR accident response. The staff is also aware that the GDCS design represented in the GIST tests is not the same as that in- '

the current SBWR design. The provisions of-10 CFR 52.47(b)(2)~

require that the specific testing supporting th6 certification o' the design must be described as-part of the application. Furthrr-more, SECY-91-273 requires that the passive plant vendors submit their test program plans, test matrices, and, upon test comple-tion, the qualified raw data to the NRC for review as part of the design certification process. Please provide any detailed infor--

mation on te " GIST" tests that is not included in the test program final report. In addition, discuss how the data will be used to support accident analyses for the SBWR; the discussion '

should include the issue of the change in GDCS design since completion of the GlST program. ,

SRXB.41 A very brief discussion of the.scuib valve test program-is made in-Section 6.3.3.2, page 6.3-14, anc the final test report is refer--

enced on page 6.3-22. However, the final test report has not been made available to the staff, nor has GE indicated how the data-from the test program will be used to support performance and reliability claiias for the DPV valves. The provisions of 10 CFR 52.47(b)(2) require that the specific testing supporting the-certification of the design must be described as ) art of the- '

application.- Furthermore, SECY-91-273 requires tlat the passive plant vendors submit their test program plans, test matrices, and, upon test completion, the qualified raw data to the NRC for review as part of the design certification process. Please provide'the-final test resort and the additional information as indicated, and discuss how tie data have been used to support the assertions made regarding DPV performance.

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SRXB.42 The staff is aware of separate effects heat transfer tests at the Massachusetts Institute of Technology and the University of California at Berkeley, for the purpose of investigating condensa-tion in the presence of non-condensible gases. While some infor-mation has been provided to the staff on the results of these tests, final reports have not been provided. These tests are not referenced in the SSAR, nor is any indication given as to how the results have been or will be used to support analyses of PCCS performance. The pruvisions of 10 CFR 52.47(b)(2) require that '

the s)ecific testing supporting the certification of the design must se described as part of the application. Furthermore, SECV-91-273 requires that the passive plant vendors submit their ,

test program plans, test matrices, and .upon test com)1etion, the qualified raw data to the NRC for review as part of tie design certification process. Please )rovide detailed information on the MIT and UCB tests, and discuss low the data will be used to support analyses of PCCS performance for the SBWR.

SRXB.43 In SSAR Table 15.0-2: for event 15.1.3, pressure regulator failure-open, the maximum neutron flux approaches 243.9-percent nuclear boiling ratio (NBR). What is the change in minimum critical power ratio (MCPR) for this event? Doesn't this tran-sient have an effect on MCPR7 Why does a scram not occur at a flux level of 194-percent NBR (Hi flux scram should occur at- -

about 125-percent NBR).

SRXB.44 For event 15.2.1, pressure regulator failure closed, why is -

critical power ratio (CPR), N/A in Table 15.0-27 There is a neutron flux increase to the Hi: flux scram setpoint which may indicate a reduction in CPR (Flux reaches 233-porcent NBR),

SRXB.45 For event 15.4.9, control rod drop accident, what would be the consequences if the separatinn-detection alarm failed-for a stuck control rod, and it could drop to its maximum distance? Will the distance of the rod drop be limited so as to preclude unacceptable consequences?

SRXB.46 for event 15.5.1, inadvertent start-up of an isolation condenser, what is the single failure assumed for this event, other than the, initiator? The SRP requires that an incident of moderate fre-

• quency in combination with any single active failure or operator error be considered.

SRXB.47 for the transients associated with a decrease in reactor coolant temperature, an increase in reactor pressure, and an increase in reactor coolant inventory, a single failure of a mitigative system should be assumed. These events should be analyzed in combination with any_ single component failure or single operator error.

SPSB.O A review of SSAR Chapter 19, Probabilistic Risk Assessment (PRA) indicates that several fundamental parts of the PRA required for SBWR certification was not included in the application for SBWR

- - , . . ,, - . , , , , n , , , . , . ,,s. ,.

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. t design certification. These sections, which constitute a najor portion of the required certification PRA, need to be addressed by GE and are stated in the following SPSB RAls.

SPSB.1 Submit containment performance analysis (level 2 PRA).

SPSB.2 Submit consequence analysis (level 3 PRA).

SPSB.3 Submit risk analysis for external events (seismic, tornados, floods, fires,etc.).

SPSB.4 Submit risk analysis for shutdown and low-power operation.-

SPSB.5 Submit sensitivity analysis (investigate the sensitivity of the SBWR design risk estimate to variations in human error probabili-ties, key assumptions, such as those related to success criteria and level 2 phenomenology, and unavailabilities of risk-important- s features which have significant uncertainties associated with them).

SPSB.6 Provide PRA-based input to design acceptance criteria (DAC)/ITAAC, ,

RAP, and TS (result of the insights gained from the importance, ,

uncertainty, and sensitivity analyses). '

SPSB.7 Submit full uncertainty analysis for the level 1 PRA (core damage frequency) including the identification and characterization of the dominant contributors to uncertainties associated with sequence and total core damage frequency estimates. The staff also expects an uncertainty analysis for levels 2 and 3 as part of the SBWR PRA submittal (an analysis along _the lines of the study performed for the ABWR PRA would suffice).

SPSB.8 Provi6 insights drawn from the PRA about design strengths and relative weaknesses, design robustness when challenged by severe accidents, and how PRA was used in the design process for preven-tive and mitigative systems.

SPSB.9 Provide all fault trees used in the PRA. Also, provide cutsets for each fault tree or for each of the event tree headings for which fault tree logic was developed.

SPSB.10 Discuss, step-by-step, the process that was used in the importance analysis to rank the various design features and human actions according to their importance in maintaining the risk levels assessed in the PRA.

SPSB.ll Provide lists of dominant cutsets for each of the quantified accident sequences. Events appearing in the cutsets should have the same basic designators as in the fault trees. (If " modular-ized" events are reported in the cutsets, please provide the link between these events and the basic events of the fault trees.)

r

Enough cutsets should be reported so that a reviewer can spot-check accuracy and consistency with fault trees. Since these cutsets will be collectively used to check the logic models of the PRA, they should also be provided on magnetic media for use by the staff. Define and justify the cutset truncation level. For each cutset, report its percent contribution to the sequence core damage frequency (CDF). Also, provide a brief discussion of the major contributors to each sequence CDF.

SPSB.12 Provide unavailabilities for all basic events modeled in the fault trees. Use the same event designators as in cutsets and fault trees (if "modularized" events apaear in any submitted cutsets, please provide the link between taese events and the basic events of the fault trees). Also, provide a similar list for component unavailabilities due to testing and maintenance.

SPSB.13 Provide documentation of dependencies between systems or functions which are displayed in omitted branch points in the event trees.

List any assumptions made.

SPSB.14 Describe the accident sequence quantification process and the capabilities of the computer ccde used for that purpose.

HICB.O The following instrumentation and control (I&C) system key topics nave not been adequately addressed, in order to perform a com-plete review, GE needs to address the following HICB RAls listed below.

HICB.1 In general, the SSAR has addressed the SRP but has not addressed the substantial amount of additional criteria related to the use of digital control equipment in the I&C systems that has been addressed in the ABWR design review.

HICB.2 There is no comparison of the SBWR design to the EPRI ALWR (Passive Plant) Man-Machine Interface System (MMIS) Requirements Document.

HICB.3 There is nt submittal for the Tier 1 design description or ITAACs.

The review of the SSAR must be concurrent with the review of .the ITAACs.

HICB.4 There is no specific description of the SBWR I&C systems hardware design. The SSAR states that the environmental qualification information sill be submitted on February 28, 1993. The SSAR does not address eb.ctromagnetic compatibility, fiber optic qualifica-tion, or other issues specific to the digital equipment that is described in the SSAR submittal.

HICB 5 There is no documentation of the SBWR software design verification and validation, configuration management control, or other aspects of software design management. The SSAR does not describe the software standards and design methods to be used.

l

HICB.6 There is no documentation of conformance with the THI action items. The SSAR states that this information will be submitted on February 26, 1993. 1 HICB.7 There is no documentation of failure modes and effects analysis (fMEA) for the I&C systems. The SSAR states that a FMEA will be submitted on February 28, 1993. However, the SSAR does not describe if the 180 systems will be specifically addressed.

HlCB 8 The unresolved safety issues / generic safety issues (USI/GSI) applicability has not been addressed. The SSAR states that this will be provided in the February 28, 1993, submittal.

HICB.9 The TS have not been provided. Several issues such as bypass.

capability and surveillance intervals and methods described in the TS must be evaluated prior to a staff final safety evaluation report.

HICB.10 The applicant has not provided a' defense-in-depth study to address potential common-mode failures of I&C system equipment.

HHFB.O Several sections in the SSAR da not contain enough information to perform a complete human fcc..n'., review. These SSAR sections are 1.8, 13.2.1-2, 13.5.2, and SSAR CFapter 18. These deficiencies are discussed in RAI numbers HHFB.1-4.16.

HHFB.1 SSAR Section 1.8, " Interfaces for Standard Design" This uction was not included in the application. The applicant states that this section will be submitted on February 28, 1992.

HHFB.2 SSAR Section 13.2.1-2, " Training" The details of the. site-specific training program are not within the scope of the SBWR standard design certification. However, the application should provide a description of the process to ensure that technically relevant training information is provided to the COL applicant.

HHFB.3 SSAR Section 13.S.2, " Plant Procedures" The details of the site-specific procedure development program are not within the scope of the SBWR standard design certification.

However, the application should provide a description of the process to ensure that technically relevant procedure development information is provided to the COL applicant.

HHFB.4.0 SSAR Chapter 18.0 Human Factors Engineeri'.g HHF8.4.1 ITAAC were not included in the application. The applicant states that SSAR Chapter 18 will be submitted on February 28, 1992.

l l

. j HHFB.4.2 DAC for this chapter were not. included in the application. l HHFB.4.3 The applicant has not described how the GS!s and USIs involving human-system concerns are resolved in the SBWR design.

HHFB.4.4 The applicant has not described MMIS design goals in " operator-centered" terms.

HHFB.4.5 The application does not identify the specific sources of opera- ,

tional experience used to develop the control room standard design features, nor how the lessons learned from such experience were incorporated into the SBWR MMIS design and implementation process described in Appendix 18E.

HHFB.4.6 The application does not identify the methodology _ used for selec--

tion of the design goals, the bases and the criteria used for selection of an individual standard design feature, or why the -

feature was selected for use.

HHFB.4.7 SSAR Section 18.5 states that the remote shutdown system (RSS)'

design is described in SSAR Subsections 7.4.1.4 and 7.4.2.4, and that the controls and instrumentation required for system o> era-tion are discussed in SSAR Subsection 7.4.1.4.4. None of 11ese '

subsections are included in the application, and the information is not included in the application.

HHFB.4.8 The bases, criteria, and inventory of controls, displays, and alarms for design-of the RSS control panel are not contained in the application.

HHFB.4.9 The application does not identify any standard design features for the RSS control system or control panel.

HHFB.4.10 The application does not state whether the process described in-Appendix 18E will be applied to the design of'any portion of the RSS, including the RSS control panel.

HHFG.4.ll The plant systems and controls to which the SBWR MMIS design and implementation process will be applied are not explicitly identi-fled in the application, t

HHFB.4.12 The process described in Appendix 18E reiterates the process depicted in Drawing 21.18E-1. However, the application does not:

• describe the qualifications and experience of the team that developed the process described in Appendix 18E; l * ' identify what standards and/or guidance were used to develop the process described in Apnendix 18E; .

l L

L j

r i

e e provide the MMIS design definition used as the basis for the '

MMIS design and implecentation process mentioned in SSAR See-tion 18E.3.6;

• state the purpose for each process element; e identify who is responsible for pv.formance of each process element; and

• describe how the individual process elements are aerformed, i.e., methodology to be used and the criteria to se applied.

HHFB.4.13 The application does not contain a description of the human factors engineering verification and validation program to be used throughout the SBWR MMIS design and implementation process. .

HHFB.4.14 According to the process as described in Appendix 18E, completion of process Elements 1 through 6 precedes sut,miltal of the SSAR.

No information regarding the conduct, results, or documentation of these efforts are contained in the application.

HHFB.4.15 The application does not discuss the methodology used for and re:ults of the following tasks:

• operating experience review;

• system functional requirements; e allocation of functions; e task analysis;

• human factors verification and validation program.

HHFB.4.16 Appendix 18E discusses Figure 18E-1 and lable 18E.2-1; however, -

neither the figure nor the table are contained in the application.

PEPB.O In general, the staff has concluded that GE's application for FDA and SBWR design certification regarding emergency preparedness requirements contained sufficient information to establish that emergency preparedness requirements have been factored into the design bases of the SBWR, with the exception of SSAR Section 1.8.

The following PEPB RAl's discusses the staff's concerns regarding SSAR Section 13.3.

PEPB.1 Table 13.3-1 contains a summary list of SBWR design. considerations pertaining to emergency planning. The table lists a Technical Support Center (TSC), Emergency Operations Facility (EOF), Opera-tions Support Center (OSC), Emergency Operations Center (E0C),

Fixed or Mobile Laboratory Facilities, Post-Accident Sampling (PASS) Capability, ar.d Onsite Decontamination Facility (ODF). The staff agrees that the E0F and EOC are not within the scope of the W

I SBWR design. However, it is the staff's position that the OSC and ODFs are required for SBWR design certification. The staff also notes that the reference as well as the emergency preparedness requirements given for the E0C in Table 13.3-1 are apparently incorrect. The E0C is usually a state or local government offsite  !

facility. Please provide the basis for inclusion of this facility in Table 13.3-1.

PEPB.2 More detailed information concerning the following facilities is needed:

• OSC - provide information on the OSC for the SBWR in suffi-cient detail to determine that the facility will meet the .,

requirements of Supplement I to NUREG-0737 and the guidance of NUREG-0696.

. 0DFs - provide sufficient information to determine.that the ODFs for the SBWR will be adequate in accordance with 10 CFR Part 50, Appendix E, Section IV.E.3.

• TSC - provide information on the TSC for the SBWR in suffi-cient detail to determine that this facility will meet the requirements of Supplement I to NUREG-0737 and the guidance of NUREG-0696.

+

• Mobile or fixed laboratory facilities - provide information on laboratory facilities for the SBWR clarifying the rale of mobile or fixed laboratory facilities in the SBWR design and the provisions made to acquire data from these facilities.

• PASS - provide sufficient information to determine that the PASS for the SBWR will meet the requirements of NUREG-0737.

PRPB.O SSAR Chapter 12 has been partially reviewed by the staff. The general finding of the staff's review is that SSAR Chapter 12 is incomplete. Specifically, the description of in-plant airborne and contained radioactive sources are inadequate. Contained sources in the radwaste building have not been submitted. A note in SSAR Section 12.2.5 indicates that this will be submitted in -

February 1993. The plant locations and source geometries are not given for those contained sources that are described in Chap-ter 12. No in-plant airborne radioactive sources are described for the SBWR design. In addition, our review found significant omissions / deficiencies in the radiation zone diagrams provided.

Some diagrams are missing. The staff counted 8 plant layout figures (Figures 21.1.2-2, Sheets 1 through 21.1.2.4) that do not have corresponding radiation zone figures. Missing-features on the zeae diagrams that are provided included: boundaries for the contamination / radiation control areas and their access traffic patterns; identification of very high radiation areas, as defined in 10 CFR art 20; location of health physics (HP) facilities, 7

i

including the onsite counting labs and their design-basis radia-tion levels;-location of all post-accident vital areas and-their:

access / egress routes during accident conditions.

p' EMCB.O The staff has completed its review of the SBWR SSAR in the func - -

tional areas of materials and chemical engineering. The following SSAR sections were-reviewed: 3.5.1.3, 4.5.1, 4.5.2, 5.2.3, 5.2.4, 5.3.1, 5.3,2, 5.3.3, 5.4.8, 6.1.1, 6.1.2, 6.2.7. 6.5.5, 9.2.3, 9.3.2, 9.3.9, 9.3.10, 10.2.3, 10.3.6, 10.4.6. The following EMCB RAls cover these sections of the SSAR.

EMCB.1 SSAR Section 4.5.1, Control Rod System Structural Materials - 1 SSAR Paragraph 4.5.1.1, " Materials Specifications," states that the following cobalt based materials will be used in the CRD system:

+ guide Roller - Stellite No. 3;

• guide Roller Pin - Haynes Alloy No. 25;

• guide Shaft - Stellite No. 6; and

+ guide Shaft Bushing - Stellite No. 12.

The use of cobalt should be avoided except in cases where no other alternative exist. The applicant should provide. justification that no other alternatives exist for the SBWR control rod drive (CRD) system. In addition, Figures 4.6-1 and 4.6-2 of the SSAR should be revised to show the individual assemblies described in Paragraph 4.5.1.1 of the SSAR. 3 SSAR Paragraph-4.5.1.1, " Materials Specifications," states that no-cold-worked austenitic stainless steels except those with-controlled haydness or strain are employed in the CRD system' These controls are acceptable. .However, the applicant should also.

commit to meet the staff position that the yield strength of cold -

worked austenitic stainless steel will not exceed 90,000 psi.

EMCB.2 SSAR Section 5.2.3 Reactor Pressure Boundary Materials SSAR Paragraph 5.2.4.4.1 states that the SBWR design complies with RG 1.44 and with the guidelines of NUREG-0313. The applicant should commit that the SBWR design complies with NUREG-0313, Revision 2.

The applicant should also commit that cold-worked austenitic-stainless steel will ccnform with the staff position that the yield strength of the stael does not exceed 90,000 psi.

p l .

SSAR Paragraph 5.2.3.1, " Materials Specifications," must state that the materials _for the reactor coolant pressure boundary will-be in conformance with the American Society of Mechanical Engi-neers (ASME) Code,Section III.

SSAR Paragraph 5.2.3.2.-2 "BWR Chemistry of the Reactor Coolant,"

states that hydrogen water chemistry will be used for the SBWR..

The applicant must commit to meet the guidelines of RG 1.56,

" Maintenance of Water Purity in Boiling Water Reactors;" F t!'s Np-4947-SR, " Hydrogen Water Chemistry Guidelines;" and EPRI's NP-3589-SR-LD, "BWR Water Chemistry Guidelines."

SSAR Paragraph 5.2.3.3.2, " Control of. Welding," states that low alloy steel components are either held for an extended time at preheat to ensure removal of hydrogen or preheat is maintained until post-weld heat treatment (PWHT). This approach does not meet regulatory position C2 of RG 1.50 which require that preheat is maintained until PWHT. _The staff had previously approved alternative approaches to complying with this requirement. 1hese.

approaches involved the use of intermittent heating at 400-500 'F.

for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by slow cooling to ambient temperatures or requiring that the component be radiographically examined after final PWHT. The applicant must commit to meet one of these approaches.

SSAR Table 5.2-4, " Reactor Coolant Pressure Boundary Materials,"

is a list that shows the mounting bolts for the CRD system will be made of SA 194, Grade B7 material. This is an apparent typograph-ical error and should be checked to see if "SA 193, Grade B7" is the correct statement.

EMCB.3 SSAR Section 5.2.4 Preservice and Inservice Inspection and Testing of Reactor Coolant Boundary Section 5.2.4 of the SSAR states that the design to perform preservice inspection is based on the requirements of the ASME-Code,Section XI, 1989 Edition. The- development of the preservice and inservice inspection program plans will be the responsibility of the COL applicant and will be based on the ASME Code, Sec-tion XI, Edition and Addenda specified in accordance with 10 CFR Part 50, Section 50.55a. For design certification, GE is respon-sible for designing the reactor pressure vessel for accessibility to perform preservice and inservice inspection. Responsibility for designing other components for preservice and inservice inspection is the responsibility of the. COL applicant. The COL applicant will be responsible for specifying the Edition of the ASME Code,Section XI, to be used, based on the procurement date of the component per 10 CFR Part 50, Section 50.55a. The ASME Code requirements discussed in this section are provided for information and are based on the 1989 Edition of ASME Section XI."

t The.1989 Edition of ASME Section XI is referenced in 10 CFR 50.55a(b). Therefore, this_ national standard is acceptable for .

use for the preservice inspection'(PSI) pursuant to the require-ments of 10 CFR 50.55a(g),

The concept of designing the components to perform the~preservice inspection based on the 1989 Edition of ASME Section XI is a  ;

reasonable approach. However, the staff concludes that the COL -

applicant must resolve any differences between the-reference code (the 1989 Edition) and the code edition ~ required by 10 CFR 50.55(g).

SSAR Paragraph 5.2.4.2, " Accessibility," states that "all items ,

within the Class I boundary are designed, to the extent practical, to provide access for the examinations required by ASME Sec-tion XI, IWB-2500."

Since the preservice inspection requirements are established and-known at the time each component'is ordered, 10 CFR 50.55a(g) does ,

not have provisions for " relief requests" for- impractical examina-tion requirements. ASME Section XI has' provisions to use certain shop and field examinations in lieu of the onsite preservice-examination. Therefore, the utility-applicant must incorporate .'

plans for NDE during construction in order to meet all access requirements of the regulations.

SSAR Paragraph 5.2.4.2 describes access for examination of the-reactor pressure vessel (RPV).

Examination Category B-A of table IWB-2500-1 requires that an inservice inspection be performed on essentially 100 percent of the weld length of all RPV shell welds as indicated in SSAR Para-graph 5.2.4.3.2. The design-of the RPV, biological shield wall-and vessel insulation incorporates access for examinations from-the outside diameter surface. Automated examinations-from the inside diameter surface-may be required to completely examine the shell welds and to evaluate the origin of reflectors detected during the inservice inspection. Describe access to the RPV welds for ultrasonic examination from the inside diameter surface.

SSAR Paragraph 5.2.4.3.2 " Examination Methods" indicates that the examination techniques will be based on the 1989 Edition of ASME'Section XI as supplemented:by RG 1.150 for the RPV.

The ASME Section XI indicates that the preservice examination should be conducted with equipment and techniques equivalent to those that are expected to be used for subsequent inservice examinations. Improvements in the ultrasonic testing of reactor coolant pressure boundary (RCPB) companents will occur-in the near future. The ASME has published in ASME Section XI,' Appendix VII,

" Qualification of Nondestructive Exaaination Personnel for Ultra-sonic Examination," and Appendix VIAI, " Performance Demonstration for Ultrasonic Examination Systems.' The NRC has referenced in 10 CFR 50.55a(b) the ASME Section XI edition that includes the published Ap)endix VII. In addition,-the NRC staff has estab-lished a tecinical contact to coordinate the implementation of Appendix VIII. Therefore, the SSAR should include provisions that ultrasonic testing during the preservice inspection be performed in accordance with Appendices VII and VIII~ pursuant to 10 CFR 50.55a(g)(3).

EMCB.4 SSAR Section 5.3 Reactor Vessel SSAR Paragraph 5.3.1.2, "Special Procedures Used for Manufacturing and Fabrication," specifies maximum limits on copper, phosphorous and sulfur for base end weld materials in the beltline region.

The applicant must also include a maximum limit of 0.05 vanadium for weld materials in the beltline region.

For staff position regarding compliance with the_ recommendations of RG 1.50, see Section 5.2.3, Question 5.

SSAR Paragraph 5.3.1.6.1, " Compliance with Reactor Vessel Materi-als Surveillance Program Requirements," states that three capsules are provided to meet the 10 CFR Part 50 Appendix H requirements.

The staff finds this commitment not acceptable since the SBWR is designed for a 60-year life. The applicant must commit to provide at least four capsules and require a minimum capsule lead factor of 1. ,

SSAR Paragraph 5.3.1.8, " Regulatory Guide 1.65," states that the RPV studs, nut, and washer materials will be ultrasonically examined after final heat treatment and prior to treading. The applicant must also commit to surface examine those items using magnetic particle or liquid penetrant examination after final heat treatment and prior to treading.

SSAR Paragraph 5.3.3.2.1, " Summary Description," states that the interior of the RPV is' clad with stainless steel wel.1 overlay and the bottom head is clad with Ni-Cr-Fe alloy. The applicant must specify the cladding process used and identify the weld materials by specification and type.

SSAR Paragraph 5.3.3.2.2, " Reactor Vessel Design Data," states that CRD' forged stub tubes for the CRD housing are made of ASME SB-564 materials. The applicant must specify which grade of materials will be used. The applicant should also include the material specifications for the RPV drain nozzles and partial penetration instrumentation water level nozzles.

SSAR Paragraph 5.3.4, " COL license Information," should be revised to reflect that the COL applicant is to provide to the NRC staff for review actual PT limits curves for the specific RPV.

l EMCB.5 SSAR Section 5.4.8 Reactor Water Cleanup / Shutdown Cooling System There should be a provision for automatically maintaining flow through filter /demineralizer units in the event system flow decreases to a point where the bed may drop from septum.

The SSAR should address spent resin transfer from the domineralizers including a description of the monitoring system.

The SSAR should describe the provisions for venting the RWCU components during drain and fill operations.

GE should more explicitly specify materials of construction of the RWCU System. From the statement made in the SSAR, it is net clear if stainless steel-is used in the whole RWCU/SDC system or- ,

in its water cleanup portion only.

SSAR, page 5.4-29, 3rd paragraph. It should be 56 *C/hr (100 'F/hr) cooldown rate, instead of 56 'C (100 'F).

EMCB.6 SSAR Section 6.1 Engineered Safety Feature Materials Table 6.1-1, " Engineered Safety Features Component Materials,"

states that containment vessel liner plate may be SA-285 Grade A up to 64mm. This is not acceptable because it does not meet the requirements of Paragraph NE-2221(c),Section III of the ASME Code which require that only. fully killed and vacuum degassed steels are used for the containment construction.

EMCB.7 SSAR Section 6.1.2 Organic Materials The SSAR imposes radiation exposure limit for organic materials of IE10 rads. GE should state that this limit applies for the whole life of the plant.

COL License Information should require that protective coatings in -

post-accident environments consider the generation of hydrogen from Zn containing primers and topcoats.

The protective coatings should meet the requirements of ANSI 101.2, " Protective Coatings-(Paints) for Light Water Nuclear Reactor Containment Facilities."

EMCB.8 SSAR Section 6.2.7 COL License Information SRP 6.2.7, " Fracture Prevention of Containment Boundary," requires that ferritic. materials that are part of containment pressure

-boundary meet the fracture toughness invoked for Class 2' materials effective with Summer 1977 Addenda. The applicant must make this commitment.

EHCB.9 SSAR Section 9.3.2 Process and Post Accident Sampling System The process and post accident sampling System (PASS) should meet.

the requirements of Section II.B.3 of NUREG-0737.

One of the requirements of NUREG-0737 is that PASS should have capability to analyze liquid "mples with the upper nuclide concentration of 10 C1/g. If the upper limit for measuring PASS -

samples is only 1 Ci/g, it may take inordinately.long time for the samples to decay to this level of activity. In some cases, this decay time may cause unacceptable delay in obtaining the results.

SRP 9.3.2, " Process and Post-Accident Sampling System," requires

~

that, in addition to the sampling described in the SSAR, process sampling system should have capability to take the following samples: sump inside containment, main condenser evacuation system offgas and inlet and outlet of gaseous radwaste storage tank.

EMCB.10 SSAR Section 9.3.9 Hydrogen Water Chemistry The design of the hydrogen water chemistry system in SBWR should meet the requirements of EPRI Report NP-4500-SR-LD, " Guide-lines for Permanent BWR Hydrogen Water Chemistry Installations."

EMCB.11 SSAR Section 9.3.10 0xygen Injection System SSAR Paragraph 9.3.10.1, " Design Basis," states that during power operation, deaeration in the main condenser may reduce the conden-sate oxygen concentration below 20 ppb, thus requiring that oxygen be added. The amount required is up to approximately 5 cubic feet per hour. The last sentence should read: The amount required is up to approximately 5 standard cubic feet per hour.

EMCB.12 SSAR Section 10.2.3 Turbine Integrity l

SSAR Paragraph 10.2.3.1, " Materials Selection," states that Charpy tests will be performed in accordance with ASTM A-170. This is a typographical error and should be corrected to read ASTM A-370. ,

SSAR Paragraph 10.2.3.4, " Turbine" states that the turbine rotor design will be solid forged monoblock rather than shrunk-on disks.

The applicant must specify that the center of the shaft will be bored to remove metal impurities and permit inspection.

RPEB.O The following RPEB RAls cover SSAR Section 14.2, Initial Test Program.

RPEB.1 SSAR Section 14.2.1.1' Construction Test Objectives SSAR Section 14.2.1.1 provides the purpose and scope of the construction and installation test program. This section states that the test abstracts will not be provided. It is the staff's I

position that GE should state how the construction and installa-tion tests will be developed and who will be responsible for-performing those tests.

RPEB.2 SSAR Section 14.2.2 Test Procedures.

SSAR Section 14.2.2 discusses, in part, review, evaluation, and approval of initial plant test results. It is stated that the final approval of test results is obtained from the Startup Coordinating Group and the appropriate level of plant management as defined in the Startup Administrative Manual. It is the staff's position that this section of the SSAR should be modified to clarify that the review and approval of preoperational test results are normally required prior to fuel loading. If portions of any preoperational tests are intended to be conducted, or their results approved, after fuel loading, the staff has determined that the applicant referencing the GESBWR-DC should be required to: (1) list each test; (2) state which portions of each test

,will be delayed until after fuel loading; (3) provide technical justification for delaying these portions; and (4) state the power levels where each test will be completed.

RPEB.3 SSAR Section 14.2.3 Test Program's Conformance with Regulatory Guides It is the staff's position that SSAR Section 14.2.3 should be modified to include the following items:

• RG 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," in accor-dance with SRP Section 14.2;

• The applicable portion (shutdown cooling) of RG 1.139, "Guid-ance for Residual Heat Removal," in accordance with SRP Section 14.2.

RPEB.4 SSAR Section 14.2.4 Utilization of Reactor Operating and' Testing Experience in the Development of Test Program-While the staff agrees with the statement in SSAR Section 14.2 that many parts of the SBWR plant desigr. have the benefits of experience acquired with-the successful and safe startup of more than 30 previous BWR design plants, the SBWR design is o'ne of the first standardized nuclear power plant designs which uses simpli-fled, inherent, and passive means to accomplish its safety func-tions. Therefore, it is the staff's position that SSAR Sec-tion 14.2.4 should address a review of the vendor test program as required by 10 CFR 52.47(b)(2)(i)(A) and incorporate results of the vendor test program into the initial test program, as appro-priate.

RPEB.5 SSAR Section 14.2.7 Test _ Program Schedule and Sequence It is the staff's position that SSAR Section 14.2.7 should address the requirement of the applicant referencing the GESBWR-DC to:

(1) list each test, that will not be performed prior to exceeding 25-percent power, for all plant structures, systems, and compo-nents that are relied upon to prevent, to limit ~, or to mitigate-the consequences of postulated accidents; (2) provide technical justification for delaying the tests; and (3) state the power levels where each test will be completed. ,

RPEB.6 SSAR Section 14.2.8 Individual Test Description It is the staff's position that test abstracts in this section of the SSAR should address the following:

• The level of detail in the test abstracts is insufficient to determine conformance with RG 1.68, Position C.2.

• Several test abstracts include imprecise acceptance criteria (i.e., acceptable, allowable, design, estimated, expected, selected, specified, appropriate, and within limits). Indi-vidual test abstracts should clearly specify the bases for determining acceptable-system and component performance.

Acceptable criteria include specific references to regulatory-guides, technical specifications, assumptions used in the safety analysis, other GESBWR-DC sections, and applicable codes and standards.-

RPEB.7 General Comments on SSAR Section 14.2-It is the staff's position that individual tests -listed in Sec-tion 14.2 of the SSAR for structures, systems, components, and features that are not essential to the demonstration of confor-mance with design requirements important to safety, but which meet any of the following criteria, should be identified.

• Those that will be used for shutdown and cooldown of the reactor under normal plant conditions and for maintaining the reactor in a safe condition for an extended shutdown period.

• Those that will be used for shutdown and cooldown of the reactor under transient (infrequent or moderately frequent events) conditions and postulated accident conditions and for maintaining the reactor in a safe condition for an extended shutdown period following such conditions.

• Those that will be used for establishing conformance with safety limits or limiting conditions for operation that will be included in the facility technical specifications.

)

• iThose that are classified as engineered safety features or will be used to support or ensure the_ operations of engineered safety. features within design limits.

• Those that are assumed to function or for which credit is taken in the accident analysis for the facility, as described in the SSAR.

• Those that will be used to process, store, control, or limit -

the release of radioactive materials.

RPEB.8 General Comments on' SSAR Section 14.2.8 It is the staff's position that the Table of Contents in Sec-tions 14.2.8.1 and 14.2.8.2 should be extended to 1_ist all preop-erational tests and startup tests covered in SectionL14.2 of-the SSAR.

RPEB.9 Chapter 17.3, Reliability Assurance Program During Design Phase, GE initially developed a design reliability assurance program _

(D-RAP).for the ABWR. Most of-the specific comments listed in the following RPEB RAls are based on the-differences the staff noted between the text of the ABWR and SBWR D-RAP submittals. GE should identify if these two programs are to be maintained independently or if common methodology and management of the programs will be used as, for example, in the Quality Assurance Program.

SSAR Section 17.3.1 states that a plant owner / operator will have an cperational reliability assurance program (0-RAP). However, an-ownsr/o_oerator will also be required to have a D-RAP for those risk-significant systems, structures, and components-(SSCs) that are not covered by the GE-NE D-RAP and those risk-significant SSCs?

that are designed or procured by the owner / operator or their agent. GE should clarify that SSAR Section 17.3.1 describes the GE-NE D-RAP and GE should state that an-owner / operator wi'l be required to provide both a D-RAP and-an 0-RAP.

RPEB.10 SSAR Section 17.3.4 states a major factor in plant reliability assurance is risk-focused maintenance. However, the description appears to be limited to safety-related maintenance and not risk-focused maintenance. GE should clarify what is meant by risk-focused maintenance in SSAR Section 17.3.4.

RPEB.11 SSAR Section 17.3.5 refers to Figure 17.3-1, " Typical GE-NE Organizational Chart for an SBWR Project." The staff noted this organizational chart differs from the chart provided in the ABWR' SSAR and pertains only to the GE-NE portion of the D-RAP. The section also describes the correct D-RAP organization.in the future tense. GE should: (1) state that a combined operating license applicant will need to supply a D-RAP organization -

description at the time of application for those risk-significant SSCs that are designed or procured by the applicant; (2) clarify:

the differences between the'ABWR and SBWR D-RAP organizations; and _;

(3) use the present tense to describe the GE-NE 0-RAP organization that is currently in place.

RPEB.12 SSAR Section 17.3.7 states the reliability of risk-significant 'l SSCs, which are identified by the PRA, will be evaluated at the detailed design stage by appropriate design reviews and relia-bility analyses. GE should clarify the meaning of " detailed design stage" and indicated if it is before or after FDA. While the use of PRA to-determine risk-significant SSCs is preferred,.

there are systems or events (e.g., fires) where use of importance measures are limited by the level of detail in the PRA models.

Therefore, GE should expand its definition of ways of identifying risk-significant SCCs to include the use of deterministic or other methods.

RPEB.13 SSAR Section 17.3.9 describes 0-RAP activities and the process-used to identify which activities are needed for a particular SSC.

Periodic testing is one of the 0-RAP activities described-in this '

setion. However, the description of periodic testing is not -

clear as it relates to inspection of SSCs such as tanks and pipes.

(For exaple, is inspection the only periodic testing that will be performed on SSCs such as tanks and pipes, or.will-inspections be limited to SSCs such as tanks and pipes?) GE should clarify what periodic testing will be applied to SSCs such as-tanks and pipes ,

in SSAR Section 17.3.9.

RPEB.14 SSAR Section 17.3.10 outlines portions of a referencing'appli-cant's 0-RAP. The 0-RAP will have various programmatic interfaces that are listed in this section including procurement of replace-ment equipment. However, the initial equipment procurement done by the combined operating license. applicant is not addressed. GE should include both initial and replacement equipment procurement.

in the list of programmatic interfaces in SSAR Section 17.3.10.

RPEB.15 SSAR Section 17.3.11.4 describes the identification of risk-significant SSCs. However, Table 17.3-1, "ICS Components With-Largest Contribution to Core Damage Frequency," is not referred in this section or any other section in Chapter 17.3. Also, SSAR ,

Table 17.3-1 must include risk importance measures (Risk Achieve--

ment and Fussell-Veseley) associated with the components listed in the table. GE should reference Table 17.3-1 in the text of the -

SSAR and should reference or discuss the associated importance-measures of the component's contribution to core damage frequency.

RPEB.16 SSAR Section 17.3.11.5 refers to components in Table 17.3-2 as-having high importance and uses that result to show how Fig-ure 17.3-2 does not provide a relat_ive measures of the components contribution to core damage frequency. As stated above,.

Table 17.3-1 also lacks such a measure. GE should provide some-relative risk importance measure in these tables so that the system design response argument can be more easily followed.

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EMEB.O The staff has preliminarily reviewed the critical valves for the SBWR plant design. A crucial part of the review will be the-reliability of.these types of valves. In order for the staff to begin its review, detailed technical information regarding the design and reliability studies will be needed. The staff recog-nizes that some of-this detailed information may not be available at the design certification stage of the review. However, perti-nent assumptions that were used in determining valve design and reliability may be beneficial to the staff and should be substi-tuted in such cases where technical information is unavailable.

The staff will also need to conduct working-level meeting (s) with GE to determine if the level of technical information submitted to the staff is adequate _in order to complete the SSAR review.

EMEB.1 Detailed technical information on the design and component reli-ability are needed for the following-list of valves:

• B32 F001, F004 (isolation condenser) N2 rotary 10" and 6" gr.te MOVs,

• B32 F005 (isolation condenser) NC 6" gate MOV,

• B32 F006 (isolation condenser) NC 6" N2 piston-operated globe valve,

• E50 F002A-F, F006A-C (gravity driven cooling system) 6" squib valves,

• E50 F003A-F, F007A-C-(gravity driven cooling system) 6" tilting disc, biased open check valve,

• E50 F009A-C (gravity driven cooling system)-4" squib valve,

• C41 F003A/B (borated water injection line) 2" squib valve,

• SRVs -8 in all (automatic depressurization) dual mode _

solenoid-operated valve (SOV)/ steam-pressure operated, and a DPVs -F004A-F- (automatic- depressurization) squib valves.

SPLB.O Section 3.11 Environmental Qualification of Safety-Related Mechanical and Electrical Equipment SPLB.1 In Section 3.11.1 of the SBWR SSAR it is stated that "a list of all 10 CFR 50.49(b) electrical and safety-related mechanical equipment that is located in a harsh environment area will be_-

included in the Environmental Qualification Document (EQD) to be prepared as mentioned in Subsection 3.11.4."

The staff finds it acceptable to qualify both electrical and mechanical equipment in accordance with the requirements of 10 CFR 50.49. However, 10 CFR 50.49 is not'a requirement for environmen-tal qualification of mechanical equipment. There are no detailed-l L

requirements for environmental qualification-of mechanicaljiquip- e ment;.however,;GDC;1, " Quality Standards and~ Records,"-GDC:--4,L

" Environmental and Missile-Design Bases," and; Appendix- B _ to 10 CFR :

Part 50, QualityJAssurance Criteria for Nuclear Power Plants and.

Fuel Repr_ocessing Plants," (Section III, " Design Control,"tand '

XVII,_ " Quality Assurance-Records"),_ contain the following requir6 ments related to; equipment qualification:-

• . Components shall be designed to be compatible with:the postu-lated environmental conditions,- including those associated -

with LOCAs,

• Measures shall be established- for.the- selection and review for; ,

suitability of application of _ materials, parts, and equipment that are essential to safety-related _ functions.

• Design control measures shall be established for verifying _:the -

adequacy of-design.- ,

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. Equipment qualification records'shall be maintained and shall include the results of tests and material analyses.

For mechanical equipment, the staff review will. condentrata on" materials (e.g., seals -gaskets, lubricants,. fluids for hydraulicx systems, diaphragms, etc.)-which are.. sensitive to eMroamental s effects. .Your review and evaluation should include the following::  ;

1) Identification of safety-related mechanical equipment f ocated - -

in harsh environmental areas, including required operat.ing time. .

2) Identification of .non-metallic. subcomponents of. this : equip-ment, ,

3) Identification'of the environmental conditions this= equipment- ,

must be qualified for. .The-environments defined in the '

electrical equipment' program.are.also applicable to mechanical-L equipment,

4) Identification of non-metallic material capabilities.

5) Evaluation

of environmental. effects.

L If it- is decided that environmental qualification-of ' mechanical l equipment will.be in accordance'with 10 CFR 50.49,1as is currently-L indicated in SBWR SSAR Section 3.11.I', then the-electrical: equip-ment and the mechanical equipment must be identified as a separate, groups. ,

p SPLB.2 In Section 3.11.2.1 the radiation source-term'used in the accident

' analysis must be identified (e.g., will TID-14844 be used in-accordance with guidance of NUREG-0588 and RGs 1.3 and 1.4). ~

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SPLB.3 Confirmed that the environmental qualification records discussed

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in S5WR SSAR Section 3.11.4 will be_in accordance with require-ments of_10 CFR 50.49(j).

The following 21 RAls refer to Section 3.4.1, Flood Protection.

SPLB.4 Identify all safety-related equipment and equipment-important-to-safety (ITS) (i.e., non-safety related equipment whose failure could adversely affect the ability of safety-related equipment to perform its safety function) requiring protection from internal and external flooding.

SPLB.5 Provide a flood analysis that identifies potential sources of internal flooding on a floor-by-floor basis-in all buildings containing safety-related equipment. How will safety-related equipment and equipment ITS be protected from flooding from these sources?

SPLB.6 Identify which safe shutdown equipment will be' located above the maximum flood height and which will be qualified for flooded-conditions.

SPLB.7 Discuss the ability of safety-related equipment to perform its

-safety function while fully flooded, partially flooded, or wet

.. (e.g., from spray)?

SPLB.3 The SSAR states that exposure to water spray will be evaluated once equipment locations and piping routings are finalized. Who will perform this evaluation, GE, or the COL applicant?

SPLG.9 Is flooding associated with the break of a high-energy line considered in the. flood analysis?

SPLB.10 Is separation of equipment utilized as a means of flood protec-tion? If so, which safety-related systems utilize separation to provide flood- protection?

SPL6.ll Identify all watertight doors and hatches on the general arrange-ment drawings.

SPLB.12 Are any internal passageways too large to close with a single door? If so, how will leakage be prevented?

SPLB.13 Provide design information on water seals, waterstops, watertight doors, and other protective features.

o SPLB.11 Identify all monitors which detect flooding in areas containing safety-related equipment and equipment ITS safety related.

Are these monitors safety related?

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SPLB.15 Do any open-cycle systems enter any buildings housing safety- .

related equipment and equipment ITS? If so, how will this equip-ment be protected from the effects of a break.in that part of.the open-cycle- system within the building?

SPLB.16 How will safety-related equipment and equipment ITS be protected-from failures of structures, systems, and components which are not within the SBWR design scope?

SPLB.17 How will the remote shutdown panel (RSP) be protected from exter-nal and internal flooding?

SPLB.17 The SSAR states that onsite storage tanks will be located to allow drainage without damaging site facilities. How will this be accomplished?

SPLB.18 The SSAR states that thu drain collection system and sumps are designed and separated so that drainage from a flooding compart-ment containing equipment for a train or division does-not flow to-compartments containing equipment for another system train or division. Provide design details of the drain collection system and sumps.

SPLB.19 Identify whether flood protection depends upon the use of a dewatering system, if so, provide seismic, safety class, and quality group classifications.

SPIS.20 Identify potential sources of external flooding from components which are within the SBWR design scope.

SPLB.20 What are the probable maximum precipitation and prnbable maximum flood?

SPLB.21 Identify safety-related equipment and equipment 3IS which are subject to groundwater seepage, and discuss how this will be controlled.

SPLB.22 Provide a discussion of possible flood hazards resulting from below-grade tunnels between buildings.

SPLB.23 SSAR Section 5.2.5 Reactor Coolant Pressure Boundary Identify and describe the monitoring of all potential intersystem leakages that are not included in SSAR Section 5.2.5.2.2 subpara-graph, "Intersystem Leakage Monitoring." Your response should include all' the applicable (for the SBWR) systems and components connected to the reactor coolant system that are listed in Table 1 of SRP Section 5.2.5 and other systems that are unique to SBWR.

Revise SSAR accordingly.

SPLB.24 SSAR Section 5.2.5- Reactor Coolant Pressure Boundary RG 1.45 Position C 7 states that procedures _for converting various indication to a common leakage equivalent should be available to-the operators. Explain how SBWR will comply with this position.

SPLB.25 SSAR Section 6.7, Main Steam Isolation Valve Leakage Control System, states that the SBWR alternate to a~ main steam isolation valve leakage control system is contained in Appendix 19H. The_-

staff finds that Appendix 19H to the SSAR, entitled as USI/GSI

. Applicability, does not contain any. information, and the applicant indicates that the information will be provided by February 28, 1993. The staff cannot start review of this subject until-the aromised information is provided, and the review schedule should ae developed based on the revised schedule of the submittal.

The staff has reviewed the information on the same subject in the EPRI Requirements Document for Passive Plants. It was identified as an open issue, page 18.0-6 in the staff's evaluation documented in the draft safety evaluation report (DSER), Section 2.3.1 and Item II.E of Annex A of Appendix of Chapter 1. The applicant's submittal should address the staff concern identified in the above

. DSER.

SPL8.26 SSAR Section 9.2.1.2 states that the plant service water system (PSWS) rejects heat from non-safety-related components in the reactor and turbine buildings to the environment. Heat is rejected to the environment by mechanical-draft cooling towers (site specific). However, the water source for the PSWS was not-addressed. Section 9.2.9 states that the COL applicant shall provide the design of the service water basin or other site-specific water supply. Provide provisions or reference criteria for the design of PSWS water intake devices.

SPLB.27 SSAR Section 6.4 Control Room Area Ventilation System SSAR Figure 21.6.4-1, Emergency Breathing Air P&lD, indicates that (1) the Pressure and Integrity of Nuclear Components (SPEC) and (2) Emergency Breathing Air System P&lD-data will'be provided "later." The above information is needed for the staff's compli-ance review of the emergency breathing air system (EBAS).

! SPLB.28 Provide detailed specific conformance review for each of the HVAC l

subsystems under SBWR SSAR Section 6.4 and Subsections 9.4.1-9.4.4 and 9.4.6-9.4.8 against the guidelines of NUREG-0800, SRP, SRP l Sections 3.4.1 for flood protection, SRP Section 3.5.1.1 for l protection against internally generated missiles, SRP Section 3.5.2 for protection against extremely generated missiles and SRP Section 3.6.1 for_ protection against high- and moderate-energy pipe breaks. This review should be in detailed fashion for each l

involved system and its components versus cross-referenced gener-alized conformance in SBWR SSAR Section 3.0.

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SPLB.29 SSAR Subsection 6;4.1.l' states that the;" Sealed-Emergency Operat-

• ing Areas (SE0A)" envelope 1s sufficiently leak tight to maintain:

positive differential: pressure of 34.5 Pac (0.005: psi) with:the-EBAS in operation. Also,-SSAR-Subsection 6.4.3 states that the.

SE0A boundary walls are designed with? low leakage constructions,: '

all boundary penetrations are sealed.1 SSAR Table:15.6-9 identi-fies unfiltered in flow of- equivalent to 0.5 cubic feet per.. minute (cfm). .

The staff considers 0.5 cfm un' filtered 'inleakagel forl the entire ; i control room envelope unrealistic as' judged from the to-date  :!

experience of the existing operating plants.- Reassess:the unfil -

tered infiltration inside SE0A envelope-and provide credible - 0 infiltration inleakage which can be supp'orted by approved method-ology and which can be-tested periodically.and verified. ' Al s o ,-

provide (1) the expected revised unfiltered infiltration rate =in the SE0A envelope'and (2) value credited for theLentire SE0A envelope infiltration rate in accident dose ~ca~1culations. Explain in detail how the SE0A envelope is isolated during accident '

conditions in order that it does not' exceed the to be revised value of the unfiltered infiltration rate used in accident dose _

v calculations.- Identify the permanent measures to be implemented-including sealing the SE0A envelope and periodic verification- and testing provisions. If sealants are used, _ provide their accept-ability and_ qualification-to maintain needed isolation through the proposed design plant life. ,

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