ML20149L897

Forwards Request for Addl Info Re GE Application for Certification of Advanced BWR Design by 880430.Request Addresses Areas of SRP Chapters 4,5,6 & 15,reviewed by Mechanical,Matls & Chemical Engineering Branches

ML20149L897
Person / Time
Site:	05000605
Issue date:	02/22/1988
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Artigas R GENERAL ELECTRIC CO.
References
NUDOCS 8802250119
Download: ML20149L897 (15)

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February 22, 1988 Docket No. STN 50-605 Ricardo Artigas, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Onerations Mail Code 682 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.

Our request is contained in the enclosure.

Our request for information addresses the areas of SRP Chapters 4, 5, 6 and 15 reviewed by the Mechanical Engineering, Materials Engineering, and Chemical Engineering Branches as well as those pertions reviewed by the Division of Radiation Protection and Emergency Preparedness. Questions related to the reviews by the Reactor Systems Branch and the Plant Systems Branch will be provided in the near future.

r In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by April 30, 1988. Also, in future correspondence with the NRC, please refer to the ABWR Docket Number STN 50-605.

If you have any questions on these matters, please call me at (301) 492-1104.

t Sincerely, l

original signed b Dino C. ScaTetti, yProject Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regclation

Enclosure:

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FEB 2 21933 Docket No. STN 50-605 Ricardo Artigas, Manager Licensing & Consulting Services General Electric Company Nuclear Energy Business Operations Mail Code 682 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN In our review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information.

Our request is contained in the enclosure. Our request for information addresses the areas of SRP Chapters 4, 5, 6 and 15 reviewed by the Mechanical Engineering, Materials Engineering, and Chemical Engineering Branches as well as those portions reviewed by the Division of Radiation Protection and Emergency Preparedness. Questions related to the reviews by the Reactor Systems Branch and the Plant Systems Branch will be provided in the near future.

In order for us to maintain the ABWR review schedule, we request that you provide your responses to this request by April 30, 1988. Also, in future correspondence with the NRC, please refer to the ABWR Docket Number STN 50-605.

If you have any questions on these matters, please call me at (301) 492-1104.

Sincerely, amqSA-Dino C. Scaletti, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects - 111, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated

GNC LO S u r..G.,

REQUEST FOR ADDITIONAL INFORMATION ADVANCED BWR SAFETY ANALYSIS REPORT t

DOCKET NO.:

50-605 MECHANICAL ENGINEERING BRANCH l

Chapter 5 - Reactor Coolant System and Connected Systems 210.1 In Section 5.2.1.2, the statement is made that Section 50.55a of 10 CFR 50 requires NRC staff approval of ASME Code Cases only for Class 1 components.

Revise this statement to be consistent with thecurrent(1987)editionof10CFR50.55awhichrequiresstaff j

approval of Code Cases for ASME Class 1, 2 and 3 components.

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210.2 Revise Table 5.2-1 or provide additional tables in Section 5.2.1.2 i

i which identifies all ASME Code Cases that will be used in the i

construction and in-plant operations of all ASME Class 1, 2 and 3 components in the ABWR, All Code Cases in these tables should be I

identified by Code Case number, revisic.n and title. These tables should include those applicable Code Cases that are listed either l

as acceptable or conditionally acceptable in Regulatory Guides,1.84, j

i 1.85 and 1.147.

For those Code Cases listed as conditionally l

l acceptable, verify that the construction of all applicable components

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will be in compliance with the additional Regulatory Guide conditions.

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RE_ QUEST FOR ADDITIONAL INFORMATION CHAPTERS 4.5_AND 6 GE_ABWR STANDARD, SAFETY ANALYSIS REPORT MATERIALS ENGINEERING BRANC_H DIVI _SION OF ENGINEERING AND SYSTEMS TECHNOLOGY Q250.1 Subsection 5.2.4.1 should state that the system boundary includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant system, or connected to the reactor coolant systems, up to and including A)

The outermost contatriment isolation valve in system piping that penetrates the primary reactor containment.

B)

The second of two valves normally closed during normal reactor operation in system piping that does not penetrate primary reactor containment.

C)

The reactor coolant system and relief valves.

Q250.2 Subsection 5.2.4.2 should satisfy the requirements in ASME Code IWA-1500.

Q250.3 Subsection 6.6.8 should discuss the augmented inservice inspection for those portions of high energy piping enclosed in guard pipes.

Q251.1 Subsection 5.3.1.1 should state that the materials will comply with the provisions of the ASME Code,Section III, Appendix I, and meet the specification requirements of 10 CFR 50, Appendix G.

Q251.2 Subsection 5.3.1.2 should state specific subsection NB of ASME Code to which the manufacturing and fabrication specifications were alluded, i

. Q251.3 Subsections 5.3.1.4.4 and 5.3.1.4.5 should be rewritten; the cross-reference is unacceptable.

Subsections 5.3.1.4.7, 5.3.1.5.2, 5.3.1.6.3, and 5.3.2.1.5; Revision 2 of Regulatory Guide 1.99 should be added in these subsections.

Q251.4 Subsection 5.3.1.6.1; the third capsule of the vessel surveillance program is designated as a standby; however, according to ASTM 185-82, the capsule should be withdrawn at the end of life. Provide justification for this deviation.

Q251.5 Subsection 5.3.1.6.3 states that according to estimates of worst-case irradiation effects, the adjusted reference temperature at end-of-life is less than 100'F, and the end-of-life upper-shelf energy exceeds 50 ft-lb. Provide the calculation and analysis associated with the estimate.

Q251.6 Subsection 5.3.2.1shouldclarifywheredj$ Reference 2 located. Has the NRC staff reviewed and approved Reference 27 If not, the staff needs to review Reference 2 in order to complete the review of this subsection.

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Question 251.7 Subsection 5.3.2.1.1, 5.3.2.1.2, 5.3.2.1.3, and 5.3.2.1.5 need to be rewritten.

The level of detail must be comparable to that of Standard Review Plan 5.3.2.

and Branch Technical Positi(n MTEB 5-2.

Question 251.8 Subsection 5.3.3 cited three GE documents:

1)

GE quality assurance prugram, 2)

"Approved" inspection procedures, and 3)

NEDO-10029.

Has the NRC staff reviewed and approved above documents? The staff cannot satisfactorily review this subsection without reviewing the above three documents.

. Q251.9 Subsection 5.3.3.1.1.1 discusses the 60-year life of the ABWR reactor vessel.

The NRC requirements and calculations on the fracture toughness and material properties are based on a 40-year life. Provide justification for the applicability of NRC's requirements on the 60-year life reactor vessel.

Q251.10 Subsection 5.3.3.2 should include the following information: neutron fluence, shift in reference temperature RT and upper shelf energy. The staff needs these information to compare to t k of predicted values using Regulatory Guide 1.99.

Q251.11 Subsection 5.3.3.6 should indicate that operating condition should satisfy the pressure-temperature limits prescribed in subsection 5.3.2.

Q252.1 Subsection 4.5.1.1 (1) should state that "The properties of the materials selected for the control rod drive mechanism must be equivalent to those given in Appendix I to Section III of the ASME Code or parts A and B of Section II of the ASME Code or are included in Regulatory Guide 1.85, except that cold-worked auster.itic stainless steels should have a 0.2% offset yield strength no greater

', than 90,000 psi."

Q252.2 Subsection 4.5.1.1 (2) should state that "All material for use in this system must be selected for their compatibility with the reactor coolant as described in Articles NB-2160 and NB-3120 of the ASME Code."

Q252.3 Subsection 4.5.2.2; The first sentence should read "Core support structures are fabricated in accordance with requirements of ASME Code,Section III.

Subsection NG-4000, and the examination and acceptance criteria shown in NG-5000."

4 Q252.4 Subsection 4.5.2.3; The following statelent should be added to the last sentence of the first paragraph, "The examination will satisfy the i

requirements of NG-5300."

Q252.5 Subsection 4.5.2.4 should state that "Furnace sensitized material should not be allowed."

Q252.6 Subsection 4.5.2.5 should state that "All materials used for reactor internals will be selected for their compatibility with the reactor coolant as shown in ASME Code Seccion III, NG-2160 and NG-3120. The fabrication and cleaning controls will preclude contamination of nickel base alloys by chloride ions, fluoride ions, or lead."

Q252.7 Subsection 5.2.3.2.2. is mostly an academic discussion of BWR water chemistry effect on intergranular stress corrosion cracking (IGSCC) in sensitized stainless steels. The subsection should discuss the actual ABWR water chemistry effects on IGSCC. The subsection is vague about specific remedies I

or preventive measures to avoid IGSCC in ABWR, For example, the subsection failed to discuss how much hydrogen is needed for injection into the feedwater I

system or how would the "tight conductivity control" be implemented.

Also, provide references for the "Laboratory studies...." and "available

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i evidence..." that were mentioned in this subsection.

e Q252.8 Subsection 5.2.3.2.3 should state that the requirements of GDC 4 relative to compatibility of components with environmental conditions are met by compliance with the applicable provisions of the ASME Code and by compliance with the recommendation of Regulatory Guide 1.44 Soecify the "very low limits" of the contaminants in the reactor coolant,

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. Q252.9 Subsection 5.2.3.3.1 should clarify where and how was 45 ft-lb. Charpy V value obtained.

The ferritic material used for pipino, pumps, and valves should comply with Appendix G. Section G-3100, of ASME Code Section III.

This subsection should indicate that "calibration of instruments and equipment shall meet the requirements of the Code, section III, paragraph NB-2360."

Q252.10 Subsection 5.2.3.4.1.1 should be rewritten to include more detailed discussion on avoidance of significant sensitization and on how the ABWR design complies with the NRC regulatory requirements.

Q252.11 Subsection 5.2.3.4.2.3 states that the ABWR design meets the intent of this Regulatory Guide (1.71) by utilizing the alternate approach given in subsection 1.8.

We cannot review this subsection because we have not received subsection 1.8.

In addition, this subsection should be rewritten because it lacks detailed discussion about welder qualification.

Q252.12 Subsection 6.1.1.1 should discuss ferritic steel welding in detail.

It should also discuss the control of ferrite content in stainless steel weld metal similar to that of Regulatory Guide 1.31.

i 0252.13 Subsections 6.1.1.1.3.1, 6.1.1.1.3.2, and 6.1.1.1.3.5 should be rewritten because the cross-reference is unacceptable.

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REQUEST FOR ADDITIONAL INFORMATION ADVANCED BOILING WATER REACTOR (ABWR)

ECEB 281.1 In Section 5.1 (page 5.1-2) the function of the reactor cleanup system filter demineralizer should include the removal of radioactive corrosion and fission products in addition to particulate and dissolved impurities.

ECEB 281.2 InSection5.2.3.2.2(page5.2-7)irradiationassistedstresscorrosion cracking (IASCC) of reactor internal components and its mitigation are not discussed. Present laboratory data and plant experience has shown that IASCC can be initiated even at low conductivity (d 0.3.S/cm) after long exposure to A

radiation.

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ECEB 281.3 I

In Section 5.2.3.2.2 (page 5.2-7 & 8) the ABWR Standard Plant design does not clearly incorporate Hydrogen Water Chemistry to mitigate IGSCC. Since the plant design life is 60 years, Hydrogen Water Chemistry may be of greater, importance in reducing reactor coolant Electrochemical Corrosion Potentie0 to i

prevent IGSCC as well as IASCC, If Hydrogen Water Chemistry is the referenced ABWR Standard Design the follow.ng documents should be cited:

d EPRI NP-5283-SR-A, Guidelines for Permanent BWR Hydrogen Water Chemistry Installations - 1987 Revision EPRI NP-4947-SR-LD, BWR Hydrogen Water Chemistry Guidelines:

1987 i

Revision (tobepublished)

ECEB 281.4 In Section 5.2.3.2.2 (page 5.2-9) the utilization of the General Electric Zinc Injection Passiviatation (GEZIP) process for radiation buildup control for the i

ABWR is not discussed. GEZIP was identified as a required design feature in i

ABWR presentation to NRC staff.

ECEB 281.5 In Section 5.2.3.2.2 (page 5.2-9) prefilming of stainless steel appears to be a promising method to reduce the buildup rate of activated corrosion products during subsequent plant operation. SIL No. 428 recomends prooperational testing of the recirculation system conducted at temperatures 230*F be done with the dissolved oxygen level controlled to between 200 and 400 ppb.

Is control of radiation buildup through preoperational oxygen control being considered for the BWR Standard Plant? Are mechanical polishing and i

electropolishing of piping internal surfaces also being considered for

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i reducing radiation buildup?

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ECEB 281.6 l

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i In Section 5.2.3.2.2.2 (page 5.2-9) Cobalt 60 is identified as the principle contributor to shutdown radiation levels, especially the recirculation piping l

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system of BWRs. Stellite contributes about 90% of the total Cobalt 59 input i

to the reactor water (EPRI NP-2263, BWR Cobalt Source Identification, February l

1982). Since irradiation of Cobalt 59 yields Cobalt 60, reduction in the source of Cobalt 59 is needed to reduce the buildup of shutdown radiation l

1evels.

Indicate Stellite surface areas (square feet) in nuclear steam j

i supply system and balance of plant. Provide the criteria for selecting i

Stellite plant materials for their designated application.

Provide evaluation i

j of noncobalt - containing materials whose properties are adequate to replace j

Stellite in plant applications, j

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ECEB 281-7 i

Section 5.2.3.2.2.3(4) (page 5.2-10) states that control of reactor water j

i oxygen during startup/ hot standby may be accomplished by utilizing the

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j de-aeration capabilities of the condenser.

In addition, this section states j

j that independent control of control rod drive (CRD) cooling water oxygen concentrations of 4 50 ppb during power operation is desirable to protect against t

IGSCC of CRD materials. Are either one or both of the above dissolved oxygen l

i controis incorporated in the ABWR Standard Plant design?

j ECEB 281.8

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In Section 5.2.3.2.2.3(13)(page5.2-11)itstatesthatthemainsteamline i

radiation monitor indicates an excessive amount of hydrogen being injected, j

An explanation of this occurrence should be discussed.

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4 ECEB 281.9 Section 6.4.4.2 (page 6.4-6) discusses personnel respirator use in the event I

of toxic gas intrusion into the control room. However, the chlorine detection system is not discussed. Also, any control room functions that are automatically triggered by a chlorine detector alam (closing intake datepers, j

energi:ing control room HVAC system recirculation) should be identified.

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ECEB 281-10 t

j In the October 1987 ABWR presentation to the NRC staff the design features and or requirements to improve water chemistry for GE-ABWR were specified. Address each one of these design features and/or requirements listed in Table ! in the j

ABWR Standard Safety Analysis Report.

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i TABLE I t

Comraarison of Requirements in ABWR Standard Safety Analyses Report and ABUR Presentation to NRC 5taff (October 21 & 22. 1987)

ABWR Presentation ABWR Standard r

to NRC Staff Safety Analysis Report l

t 1-Selection of Low Cobalt required design Not discussed in t

Paterials to Minimize feature Section 5.2.3 i

Radiation Buildup I

2-Hydrogen Water Chemistry required design Section 5.2.3.2.2 to Suppress IGSCC feature references normal water chemistry 3-Zinc injection to required design Not discussed in Minimize Radiation feature Section 5.2.3.2.2.2 l

Buildup 4-Full Flow Deep Bed required design Not discussed in i

Condensate System featuro Section 5.2.3.2.2.3 to Reduce Feedwater Impurities 5-Improved On-line Ion chromatography.

Only Electrochemical Monitoring Electrochemical Corrosion Potential Instrumentation to Corrosion Potential, discuss'ed in Section Assure Water Quality and Crack Arrest 5.2.3.2.2.3 Verification System required design features a

6-Improved Corrosion required design feature Not discussed in Resistant Materials for Section 5.2.3.2.2.3 s

i Steam Exttaction Piping to Einimize i

feedwater impurities 7-Highly Corrosion required design feature Not discussed in Resistant Condenser Section 5.2.3.2.2.3 Tubes to Minimize Leakage Into Condensate System 8-Maintair Electrochemical required design feature Not listed in Table Corrosion Potential Table 5.2-5 i

< 0.23V to suppress i

IGSCC 9-Erosion / Corrosion design feature Not d O ssed in f

J Resistant Materials Section 5.4.9 in Steam Extraction and Drain Lines to j

1 minimize failures l

1 Table I cont.

ABWR Presentation ABWR Standard to NRC Staff Safety Analysis Report 10- Ease of Leak Detection design feature May be in Section in and Repair of the 10.4.1 which'has Main Condenser not been submitted yet i

11-2% Reactor Water Cleanup design feature Not discussed in Section 5.2.3.2.2 System to improve reactor water quality and occupational radiation exposure 12-Full flow Recirculation design feature Not discussed in to Main Condenser from Section 5.2.3.2.2.3 i

4 Cleanup Outlet to Reduce Feedwater Impurities i

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t ENCLOSURE Q1s on ABWR FSAR Chapter 15 t

470.1 Section 15.6.2 of the ABWR FSAR provides your analysis for the radiological consequences of a failure of small lines carrying primary coolant outside of containment. This analysis only considers j

the failure of an instrument line with a n" flow restricting orifice.

j Show that this failure scenario provides the most severe radioactive i

releeses of any postulated failure of a small line. Your evaluation should include lines that meet GDC 55 as well as small lines exempt from GDC 55.

I 470.2 Provide a justification for your assumption that the plant continues to operate (and therefore no iodine peaking is experienced) during a smalllinebreakoutsidecontainment(115.6.2)accidentscenario.

Also provide the basis for the assumption that the release duration is only two hours.

I 470.3 Subsection 15.6.4.5.1.1 of the FSAR gives the iodine source term (concentrationandisotopicmix)usedtoanalyzethe l

i steam-line-break-outside-of-containment accident. The n:ble gas source term, however, is not addressed. Provide the noble gas source tem used. Also the table in $15.6.4.5.1.1 seems heavily weighted to i

the shorter lived activities (i.e., (1-134). Provide the bases for the isotopic mix used in your analysis (iodine and noble gas).

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470.4 Section 15.6.5.5 states that the analysis is based on assumptions provided in Regulatory Guide 1.3 except where noted.

For all assump-tions(e.g.,releaseassumedtooccuronehourafteraccidentinitia-tion, the chemical species fractions for iodine, the temporal decrease in primary containment leakage rates, credit for condenser leakagerates,anddoseconversionfactors)whichdeviatefromNRC guidance such as regulatory guides and ICRP2, provide a detailed description of the justification for the deviation or a reference to another section of the SSAR where the deviations are discussed in detail. Provide a comparison of the dose estimates using these assumptions versus those which would result from using the NRC i

guidance.

470.5 Provide a discussion of, or reference, to the analysis of the radiological consequences of leakage from engineered safety feature components after a design basis 1.0CA.

470.6 For the spent fuel cask drop accident, what is the assumed period for decay from the stated power condition? What is the justification for that assumption?

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470.7 The tables in Chapter 15 should be checked and revised as i

appropriate.

In several cases the foutnotes contain typographical errors related to defining the scient fic netation. Table 15.7-12 also appears to contain inappropriata references to Table 15.7-16, rather than Table 15.7-13.

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j 470.8 1

b ated that Regulatory Guides 1.3 and 1.145 were used in the t

can.u... ions of X/Q values. Based on the values presented, it appears as though a Pasquill stability Class F and one meter per second wind speed were assumed, with adjustment for meander per Figure 3 of Regulatory Guide 1.145.

If this is not the case, de-scribe the assumptions and justification used in calculating the X/Q values which are used in the Chapter 15 dose assessments.

470.9 The SGTS filter efficiencies of 99% for inorganic and organic iodine are higher than the 90% and 70% valuer, respectively, assumed in Regulatory Guide 1.25 if it can be shown that the building atmosphere is exhausted through adsorbers designed to remove iodine. Provide a justification for the use of the higher values.

i 470.10 Dose related factors such as breathing rates, iodine conversion factors and finite versus infinite cloud assumptions for calculating the whole body dose are not stated explicitly, although reference is made to Regulatory Guide 1.25 and another document. State thes'e assumptions explicitly and justify use of any values which deviate from Regulatory Guide 1.25.

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