ML20011E204

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Forwards Response to 881216 Request for Addl Info Re CESSAR-DC,Chapters 3,4,5 & 6 Re Turbine Missiles,Control Element Drive Structural Matls,Cleaning & Contamination Protection Procedures & Reactor Internals Matls
ML20011E204
Person / Time
Site: 05000470
Issue date: 01/25/1990
From: Scherer A
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
PROJECT-675A LD-90-008, NUDOCS 9002090070
Download: ML20011E204 (35)
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t' ggggggggggg January 25, 1990

~ LD-90-008 1

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Project No.' 675 v ' .,

S. Nuclear Regulatory Commission

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Attn: :' Document- Control- Desk

.Washington, D. C. 20555

Subject:

Response to Nuclear Regulatory Commission Request for-

" . Additional Information- Concerning Chapters. 3, 4, 5 and 6, Materials. Engineering Branch

Reference:

Letter, G. S. Vissing (NRC), to A. E. Scherer (C-E),

dated December 16, 1988

Dear Sirs:

The ' Reference-requested that Combustion Engineering provide 7 additional information concerning CESSAR-DC, Chapters l3 r- 4, 5, . and

6. Enclosure I to this letter provides our responses and' Enclosure II-
provides .the proposed corresponding revisions to CESSAR-DC -

,1 Should you' have any questions, please feel free to contact me or Mr. ' S. E. Ritterbusch of my staff at (203) 285-5206.-

Very truly yours, COMBUSTION ENGINEERING, INC.

Director Nuclear Licensing AES:jeb

Enclosures:

As Stated 0

cc: F. Ross (DOE-Germantown)

R. Singh (NRC) -

ON, I l \

Power Systems 1000 Prospect Hill Road (203) 688-1911

~ Combustion Engineering, Inc. Post Ot$ce Box 500 Telex: 99297 2Og o 900125 "

675A PDC

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Enclosure I: to-  ;

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  • '- Lp.go.008 Page 1 of 14 8 .

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-l RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION CONCERNING CHAPTERS 3, 4, 5 AND 6-MATERIALS ENGINEERING

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Enclosure I  :

Page 2 of 14 j

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Ouestion 252.'1 (3.5.1.3) l= .

L 3.5.1.3 Turbine Missiles a L

L - The applicant stated that the probability of turbine missile generatico-L and adverse impact on seismic category I systems is assured to be L acceptably low. This is vague and unacceptable. In earlier licensing reviews, the staff requires that the failure probability due to.

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low-trajectory turbine missile disabling an essential system should be ,

less than 1.0E-7 per year. In the last several years the staff has changed the intent of licensing requirements. The current emphasis is  !

I placed on the failure probability of turbine missile generation which L

should be .less than IE-5 per year for an unfavorably oriented turbine and a IE-4 per year for a favorably oriented turbine. The staff's position on .

this issue is documented in the following references; the first reference l l 1s for General Electric turbines and the second is for Westinghouse <

L turbines: ,

1) Safety Evaluation Report related to the Operation of Hope Creek Generating' Station, Supplement No. 6,-NUREG-1048, July 1986.

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2) ~ Letter, from C. E. Rossi to NRC to J. A. Martin of Westinghouse, '

Orlando, Florida, dated February 2,1987.

A The applicant needs to submit additional information to show that the failure probability of turbine missile generation is within IE-5 to IE-4 per year, depending on the turbine orientation.

Resoonse 252.1 (3.5.1.3)

The System 80+. Standard Design complies with the intent of Regulatory Guide 1.115 " Protection Against Low-Trajectory Turbine Missiles". This regulatory guide indicates that an incidence rate of 1.0E-4 per turbine-year is appropriate for material failures at speeds up to 130% of turbine operating ' speed, assuming historical failure data on conventional units.

l Turbine disc integrity and turbine orientation help assure that the probability of a turbine missile impact on Seismic Category I-systems will l be low. CESSAR-DC requires that acceptable disc integrity be assured -

through design and inspection, as described in Sections 10.2.2 and 10.2.4.

n In addition, the speed controls described in Section 10.2.2 ensure that the maximum turbine overspeed for the System 80+ design will be less than ,

l 112% of operating speed.

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i Enclosure !

Page 3 of 14 The System 80+ turbine generator has been oriented consistent with the intent of Regulatory Guide 1.115. That is, no safety related equipment will be located within the low trajectory missile strike zone. As a result of turbine design requirements, speed controls, and orientation, Combustun Engineering believes that the guidance of Regulatory Guida 1.115 (Revision 1) is met.

CESSAR DC, Section 3.5.1.3, will be revised to reflect this response, including a new figure showing the low trajectory missile strike zone.

Question 252.1 (4.5.1.2) 4.5.1 Control Element Drive Structural Materials 4.5.1.2 Control of the Use of 90 ksi Yield Strength Material SRP 4.5.1 reenmmends not to use stainless steel that has yield strength higher than W ksi because it is susceptible to stress corrosion cracking. The applicant specified ASTM A276 Type 4400 martensitic stainless steel to be used in the control rod drive structure. The applicant stated that the steel will be used in the r.on stress application and that this material is presently being used in several operating nuclear plants successfully. However, the staff recommends that either A276 Type 440C stainless steel not be used in the control rod drive system, or provide justification for its use.

,tc oonse 252.1 (4.5.1.2)

The usage of ASTM A276 Type 4400 material in the CEDM is based upon its hardness and is limited to an application where the stresses are compressive, precluding the potential for stress corrosion cracking. In addition, field experience of more than 16 years has proven satisfactory material performance. Moreover, it should be noted that the A276 Type 4400 material is used as a ball valve internal to the CEDM pressure housing and is not defined as part of the pressure boundary. The usage of this material was also described in the response to Question 251.11.

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i Ouestion 252.1 (4.5.1.5)  ;

1 4.5.1.5 Cleaning and Contamination Protection Procedures  ;

1 The cleaning procedures should satisfy ANSI N45.2.1 1973,

  • Cleaning of l Fluid Systems and Associated Components During Construction Phase of i Nuclear Power Plant.' j Resoonte 252.1 (4.5.1.5) i Regulatory Guide 1.37 stipulates that the on site cleaning of materials and components should follow the guidance of ANSI N45.2.1-1973. Cleaning requirements are currently included in ANSI /ASME NQA 2 1983. Compliance  !

with the intent of Regulatory Guide 1.37 is described in Section 4.5.1.5.

This statement will be supplemented to clarify compliance with ANSI /ASME NQA-2-1983. j t

t Duestion 252.1 (4.5.2.1) t 4.5.2 Reactor Internals Materials -

4.5.2.1 Material Specifications The materials used in fabrication of reactor internal structuret should '

meet the requirements of GDC 1 and 10 CFR 50.55a. The permitted material >

specifications should satisfy those given in the ASME Code, Section Ill, NG 2000 and those shown in ASME Code Cases approved for use by Regulatory Guide 1.85, " Code Case Acceptability, ASME Section Ill Materials".

Response 252.1 (4.5.2.1)

Conformance with all the General Design Criteria is discussed in CESSAR DC, Section 3.1. Also, accordance with the codes required in 10 CFR 50.55a is stated in the response to GDC 1 (Section 3.1.1).

A supplement has been made to Section 4.5.2.1 (Response 251.11) to clarify compliance with the requirements of Article NG 2000 of the ASME

! Code and the guidance of Regulatory Guide 1.85.

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Page 5 of 14  !

Ouestion 252.1 f5.2.3) l 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2.3.1 Material Specifications '

The' specifications for permitted materials should satisfy the ASME Code, '

Section III, Appendix I;Section II, parts A, 8.-and C; and Regulatory  !

Guide 1.85. I Resnante 252.1 (5.2.3.1) l Combustion Engineering agrees with the position stated in the question. l This issue was discussed in the response to Question 251.11 (5.2.3.1).

Question 252.1 (5.2.3.2) '

5.2.3.2 Compatibility with Reactor Coolant "

Unstabilized austenitic stainless steel of the AISI Type 3XX series used ,

for components of the reactor coolant pressure boundary should conform to i Regulatory Guide 1.44, ' Control of the Use of Sensitized Stainless '

Steel.' If cladding is not used in ferritic low alloy steels as recommended in the ASME Code,Section III, NB 3120. " Corrosion'.

  • Resoonse 252.1 (5.2.3.2)

Combustion Engineering agrees with the above statement. Please see the response to Question 251.11 (5.2.3.2.2) for additional information.

L Ouestion 252.1 (5.2.3.3)

! 5.2.3.3 Fabrication and Processing of Austenitic Stainless Steel 1

1 Regulatory Guide 1.36, " nonmetallic Thermal Insulation for Austenitic Stainless Steel" should be used in the selection of non metallic thermal insulation for stainless steel.

Resoonse 252.1 (5.2.3.3)

Combustion Engineering agrees with the above statement. Please see the response to Question 251.11 (5.2.3.4) for additional information.

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I Question 252.1 (5.2.4) i I

5.2.4 In-Service Inspection and Testing of Reactor Coolant Pressure Boundary . j The applicant should discuss the pre service inspection program in the- )

section. As for the ISI program, the applicant should discuss the i following areas on in-service inspection of the RCP boundary:

o ' System Boundary Subject to Ins section o Examination Categories and Met 1ods o Inspection Intervals o Evaluation of Examination Results  ;

o System Leakage and Hydrostatic Pressure Tests  ;

o Code Exemptions 1 o Relief Requests  ;

Resnonse 252.1 (5.2.4) l o " System Boundary Subject to Inspection" is defined in Section 5.2 in accordance with ANSI /ANS 51.1-1983, which is in agreement with the ,

acceptance criteria in SRP 5.2.4.11.1. l o Examination Categories and Methods: Section 5.2.4 will be amended to state compliance with the ASME Code.

o Inspection Intervals: A supplement will be made to Section 5.2.4 to state compliance with the ASME Code, o " Evaluation of Examination Results" and compliance with the ASME '

code is covered in the amended section 5.2.4.2.

o System '.eakage and Hydrostatic Pressure Test: Section 5.2.4 will be  ;

amended to state compliance with the ASME Code. The reactor coolant L pressure boundary leakage detection is covered in Section 5.2.5.

o C E does not expect to request any Code Exemptions and Relief Requests as part of.the System 80+ design certification.

The pre-service inspection of the reactor coolant system is discussed in

  • Sections 5.3.1.3, 5.4.1.4, 5.4.2.5, 5.4.3.4, 5.4.10.4, 5.4.12.4 and 5.4.13.4.

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Question 252.1 (5.3.1 M 5.3.1 Reactor Vessel Materials i 5.3.1.6.5 Irradiation Locations  !

The design and locatior of the surveillance capsules shall permit insertion of replacement capsules. If the capsule holders are attached to the vessel wall or cladding, inspection shall be done according to the requirements for permanent structural attachments as given in the ASME Code, Sections III and XI.  ;

i Resoonte 252.1 (5.3.1.6.5) 1 Section 5.3.1.6.5 will be revised to describe the capability of inserting I l

replacement capsules and to show conformance with the ASME Code Sections III and XI.

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Ouestion 252.1 (5.3.1.6.6) 5.3.1.6.6 Withdrawal Schedule 'i The minimum recommended number of surveillance capsules and their  !

withdrawal schedule depend on the predicted increase in reference '

temperature as indicated in ASTM E-185. Although the CESSAR indicates -

that the predicted increase in reference temperature will be only 50'F, the staff believes that, as a minimum, the surveillance program should be based on the schedule for a higher reference temperature increase between

( 100'F and 200'F because of the uncertainty in material properties and the ,

! 60-year proposed life of the reactor vessel. The applicant needs to '

I modify its capsule withdrawal schedule and extrapolate the ASTM i

! recommended schedule to the end of life based on 60 reactor years.

l Resoonse 252.1 (5.3.1.6.6) i l.

The System 80+ reactor vessel is designed to be very insensitive to neutron irradiation based on the prediction methods of Regulatory Guide g 1.99, Revision 2. The design basis included more than the stipulated margin added to the predicted shift to account for some of the uncertainty in the longer design life of the reactor vessel. '

Furthermore, the System 80+ surveillance program includes six surveillance capsules installed at the beginning of vessel life; this is one more than the five capsules stipulated in ASTM E-185-82 for a l

predicted end-of life (EOL) shift in excess of 200'F. If actual shift

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I measurements obtained from evaluation of the three* capsules scheduled f

for withdrawal indicate the EOL shift will exceed the 50'F shift j prediction, the three standby capsules will be available to meet additional monitoring needs. )

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The first and second surveillance capsules will be scheduled for j withdrawal and evaluation early enough in the design life to provide l timely information for making any needed changes to the subsequent  ;

withdrawal schedule.  !

Therefore, the System 80+ surveillance program adequately addresses uncertainty associated with a 60 year design life, r Table 5.3-7 has been revised (Amendment E) to provide specific data on the withdrawal schedule.

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Three are scheduled for withdrawal during the service life of the vessel, and a fourth is designated for removal after 60 years but not necessarily to be tested. 1

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. Enclosure !

. ' Page 9 of 14 Ouestion 252.1 (5.3.2.1) 5.3.2 Pressure-Temperature Limits 5.3.2.1 Limit Curves The applicant assumed a 50'F increase in the reference temperature, RT over a 60 year eriod. The staff believes that based on RT daggi,aken t from survei lance results of operating plants, the incrISle in RT could be much higher than 50'F for a 60 year reactor life. The apkIcant needs to submit test data to support the RTNOT of 50'F.

Resoonsa 252.1 (5.3.2.1)

The predicted adjusted RT at 1/4t for the System 80+ was derived using NOT Regulatory Guide 1.99, Revision 2, and a conservative margin of 50'F for the beltline forging. The bases include a chemistry factor of 37 (0.06 Cu,1.00 N1), and end of life fluence of 6 x 10 lI n/cm2 (vessel inside surface) and an initial RT NOT of -20'F, yielding the following:

initial RTNDT: -20'F predicted shift (1/4t): 49'F maraine 50*F (versus 34*F oer Reaulatory Guide 1.99) adjusted RTNDT(1/4t): 79'F The maximum predicted shift is 99'F (49'F predicted shift + 50'F margin),

yielding a conservative estimate of RT over the 60 year design life.

NDT This information has been included as part of a complete revision to Section 5.3.2 (Amendment E).

The reactor vessel surveillance program for System 80+ (Section 5.3.1.6) is designed to provide direct measurements of the effect of long term (i.e., 60 years) exposure, under conditions nearly identical to the i reactor vessel, on the toughness properties of the vessel beltline materials. If there is, in fact, an effect of very long time exposure on reference temperature shift, the surveillance program will provide direct i evidence of such an effect, and adjustments can be made to the reactor vessel operating limit curves. In keeping with 10 CFR 50, Appendix H, the System 80+ reactor vessel is designed to minimize irradiation damage and the surveillance program is designed to provide timely information on the actual trends, i

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Question 252.1 (5.3.3.1) l 5.3. 3'.1 Design The design details must be adequate to permit all required inspections l and to provide required access for inservice inspection in conformance l with the ASME Code,Section XI.  !

.Resoonsa 252.1-(5.3.3.1) [

i Please see the response to Question 252.1(5.2.4). i Ouestion 252.1 (5.3.3.2) 5.3.3.2 Materials of Construction ,

The materials of construction of the reactor vessel should satisfy the  !

ASME Code, Sections !!! and IX. Although many materials are acceptable i for reactor vessels according to the ASME Code,Section III, the special considerations relating to fracture toughness and radiation effectively limit the basic materials that are currently acceptable for most parts of  ;

reactor vessels to SA 533 Grade B Class 1 SA 508 Class 2, and SA 508 Class 3. The applicant specified SA 508 Class 1 as a material for the i reactor vessel as shown in Table 5.2-2. This material is not recommended in the above list; therefore, the applicant should either delete SA 508, Class I from its material specifications, or provide justification  ;

according to 10CFR50, Appendix G, Section I.A. l Resoonse 252.1 (5.3.3.2) l SA-508 Class 1 is listed in Table 5.2-2 as a reactor vessel material. i However, this material is only used for inlet and outlet nozzle safe ,

ends, where radiation damage is insignificant. No SA-508 Class 1  !

material is allowed in the beltline region.

SA 508 Class 1 is listed in Paragraph G 2110 of the ASME Code along with  ;

SA 533 Grade B Class 1, SA-508 Class 2, and SA-508 Class 3 as nuclear vessel material. Its fracture toughness has been demonstrated to be no lower than the KIR curve shown in Figure G-2210-1 of the ASME Code and, therefore, satisfies 10 CFR 50, Appendix G.

Table 5.2-2 of CESSAR-DC will be revised to clearly indicate that SA 508 Class 1 material is for reactor vessel nozzle safe ends and to remove it from the list of materials for reactor vessel forgings.

Enclosure I Page 11 of 14 4

Ouestion- 252.1 (5.3.3.4) '

5.3.3.4 Inspection Requirements Nondestructive inspections should satisfy the ASME Code, Sections V and I XI. The methods of inspection, the sensitivity levels, and flaw evaluation criteria should be compatible with Section XI, and the results

  • of the preservice baseline inspection should be correlated with the results of later inservice inspections.

Resoonse 252.1 (5.3.3.4)

The preservice and inservice inspection of the reactor vessel are performed in accordance with Sections III and XI of the Code (see l

Sections 5.2.4.1, 5.3.1.3, and Table 5.21). Various nondestructive examinations are conducted in accordance with the requirements of Section' V of the code, as imposed by Section XI.

CESSAR DC, Section 5.3.1.3, will be revised to more clearly state the compatibility of preservice and inservice inspection methods, procedures, and requirements.

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Question 252.1 (5.3.3.7) 5.3.3.7 Inservice Surveillance The inservice surveillance of reactor vessel should satisfy the ASME Code,Section XI.

Resoonse 252.1 (5.3.3.7)

The inservice surveillance of the reactor vessel satisfiesSection XI of the ASME Code. CESSAR DC, Section 5.3.1.6, will be revised to state this.

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.. g Question 252.1 (5.4.1.1) 5.4.1.1 Pump Flywheel Integrity l

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The a>plicant specified a dynamic stress intensity factor of 100 ksi/in i for tie reactor coolant pump flywheelt however, SRP 5.4.1 specifies a  ;

critical stress intensity factor of 150 ksi/in. The applicant should  !

clarify its stress intensity factor. l The normal operating tempetature of the pump should be at least 100'F above the reference temperature, RTNDT, of the pump flywheel.

The fracture toughness, preservice inspection, flywheel design, overspeed test, and inservice inspection of the pump flywheel should satisfy Regulatory Guide 1.14.  :

Resoonse 252.1 (5.4.1.1)

This issue was discussed in the response to Question 251.11 (5.4.1.1). I Ouestion 252.1 (5.4.2) 5.4.2 Steam Generators The selection and fabrication of steam generator materials and steam generator design.should satisfy requirements of GDC 1,14,15 and 31 and Appendix B to 10CFR50. The materials used for components in the primary side and secondary side of the steam generator should be identified the l

same as shown in SRP 5 4.2.1, " Steam Generator Materials." i Resoonse 252.1 (5.4.2)

Conformance'with all the GDCs is discussed in CESSAR-DC, Section 3.1.

These discussions summarize the manner in which the design features meet L the individual General Design Criteria.

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The quality assurance program for the System 80+ Standard Design, l including steam generator design and material selection, complies with 10 CFR 50, Appendix B. This program is described in report CENPD-210 which is referenced in Chapter 17 of CESSAR-DC.

The materials used for components in the primary side and secondary side of the steam generator, shown in Table 5.2-2, are the same as shown in SRP 5.4.2.1.I.1 except for certain differences in terminology, e.g.,

" Channel Head" is called " Primary Head" for the System 80+ steam generators. '

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Page 13 of 14 i Question 252.1 (5.4.2.5) 5.4.2.5 Tests (nd Inspections The tests and insp,ections of steam generator tubes should satisfy Regulatory Guide 1.83 and the appitcable Standard Technical {

Specifications, NURIG-0212, latest Revision. ,

Resoonse 252.1 (5.4.2.5) '

i Section 5.2.4.1 states that the steam generators are capable of being l

examined in accordance with Regulatory Guide 1.83. The proposed l Technical Specifications will include a steam generator inspection  ;

program that will be consistent with Regulatory Guide 1.83 and *

NUREG 0212. That program will be similar to the program developed for the System 80 steam generators (Chapter 16 of CESSAR F which, for convenience, is currently presented in Chaptar 16 of CESSAR-DC) which was approved in Supplement 3 to the Safety Evaluation Report (NUREG 0852).  !

Ouestion 252.1 (6.1) '

6.1 Engineered Safety Feature Materials The engineered safety feature materials should satisfy GDC 1, 4, 14, 31, 35, 41, and Appendix B to 10 CFR 50. A discussion on ferritic steel -

welding should be added to this section. (see SRP 6.1.1)

Resoonse 252.1 (6.1)

Conformance with all the GDCs is discussed in Section 3.1. These discussions summarize the manner in which the design features meet the individual General Design Criteria.

The quality assurance program for the System 80+ Standard Design complies with 10 CFR 50, Appendix B. This program is described in report CENPD-210 which is referenced in Chapter 17 of CESSAR-DC.

Section 6.1.'.1.5 is being added to CESSAR-DC to provide statements on ferritic steel welding.

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. i Question 252.1 (6.6)

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6.6 In-service Inspection of Class 2 & 5 Components j

This section should be written according to the format of SRP 6.6. The ,

following seven subsections need to be discussed in the ISI program:  :

o Examination Categories and Methods '

o Inspection Intervals i o Evaluation Results . '

o System Pressure Tests i o Augmented ISI to Protect Against Postulated Piping Failures +

o Code Exemptions ,

<- o Relief Requests l

A pre-service inspection program should also be discussed in this i

i. section.

Response 252.1 (6.6)

! o Examination Categories and Methods o Inspection Intervals o Evaluation Results o System Pressure Tests An amendment will be made to Section 6.6 to cover the above four items..

Augmented ISI to Protect Against Postulated Piping Failure o

o Code Exemptions '

l o Relief Requests '

.These items are not applicable to System 80+ Standard Design. '

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p Enclosure II to LD-90-008 Page 1 of 20 l

PROPOSED REVISIONS TO THE COMBUSTION ENGINEERING STANDARD SAFETY ANALYSIS REPORT -

DESIGN CERTIFICATION l y L

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3.s.1.3 Turblae Missiles The probability of turbine missile generation and adverse impact  !

, effects on Seismic Category I systems and components is assured  :

l to be acceptably low by a combination of the following measures:  !

A. eliglgtup overspeedprotectionprovisions(seefe.d'en

.r pe. ,. *[.[ S. *9 .f,e.v [s Y /# Y /./[I C. placement and orientation of the turbine generator (duc Ll L/,h. -

D. Consideration of the protection provided by plant structures not explicitly designed as barriers that may limit missile j penetrating capabilities to less than the capability of j

[ ismic Category I structures.

3.5.1.4 Missiles Generated by Matural Phenomena Tornado-generated missiles are the limiting natural hazard and, as such, are a part of the design basis for Seismic Category I structures and components. Table 3.2-4 (IATER) lists those structures, shields and barriers that will be designed for ,

missiles considered in the design are given !_

T C '. : 0.0 ;  :: r;;/L; tornadoin^f missile

:Mi r n S, tle M-L effects.

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  • 3.5.1.5 Missiles Generated by Rvents Maar the Site Justification will be provided in the site-specific SAR.

3.5.1.6 Aircraft Razards 4 Justification will be provided in the site-specific SAR.

3.5.3 STRUCTURES, SYSTEMS, AND COMPONENTS TO EE PROTECTED FROM ERTERNALLY GENERATED MISSILES Tornado missiles are the design basis missiles from abternal sources. All safety related systems, equipment and components -

required to safely shut the reactor down and maintain it in a safe condition are housed in Category I structures designed as tornado resistant (see Section 3.5.1.4) and as such are .

considered to be adequately protected.

3.5.3 EARRIER DESIGN PROCEDURES Missile barriers, whether steel or concrete, are designed with sufficient strength and thickness to stop postulated missiles and 1

Amendment D s

3.5-4 September 30, 1988

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The turbine generator placement and orientation for the System 80+ Standard Design and the corresponding low trajectory missile strike zones are I illustrated in Figure 3.5 1. The placement and orientation of the turbine generator provides protection against low trajectory turbine missiles by excluding safety related structures, systems, and components from the low trajectory turbine missile strike zones in accordance with the guidelines of Regulatory Guide 1.115.

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. CESSAR.tBib l";/fD !! .

l A. For- systems and parts of systems located inside the containment (RCS and connected systems, Engineered Safety

Feature systems), appropriate missile barrier design i procedures are used to ensure that the impact of any potential. missile will not lead to a loss-of-coolant-accident or preclude the systems from carrying out their l specified safety functions. 1 1

B. For systems and equipment outside containment, appropriate I design procedures (e.g., proper turbine orientation, natural '

separation, or missile barriers) are used to ensure that the D i impact of any potential missile does not prevent the system I

. or equipment. from carrying out its specified safety function. <

C. For.all systems and equipment, appropriate design procedures ,

are used to ensure that the impact of any potential missile does not. prevent the conduct of a safe plant shutdown, or '

prevent the plant from remaining in a safe shutdown condition.

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CISSILE DESCRIPTIONS (kg) (Ibs) (ma ) (in a) (m/sec) (ft/sec) (m/sec) (ft/pec) 52 115 0.027 41 83 272 58 198 l A tIood Plank 0.092 x 0.299 x 3.66 34 52 171 36 119 0.168D x 4.58 130 287 0.022 i s G," Sch. 4C Pipe 0.00051 0.79 51 167 51 167 0.0254D x 0.915 4 8.8

! C 1" Steel Rod 0.092 143 55 188 39 126 D Ct111ty Pole 0.3430 x 10.68 510 1124 j

i 125 47 154 33 108  ;

12" Sch. 40 Pipe 0.32D x 4.58 340 750 0.000 E t 2.60 4030 59 194 41 136 1810 3990 F Automobile 2 x 1.3 x 5 l

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j Cissiles A, B, C, and E are to be considered at all elevations and missiles D and F et elevations up to 30' grade levels within 1/2 mile of the structure.

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- Enclosure 11 1CtSSARIB h Weld heat affected sono sensitised austenitic stainless steel (which will fail in the strauss Test, ASTN A708) is avoided in control element drive mechanism structural components by careful control of A. Weld heat input to less than 60 AT/in B. Interpass temperature to 350'F maximum

c. Carbon content to s 0.065 lD 4.5.1.4 control of Balta Perrite in Austamitie stainians steel Walem The austenitic stainless steel, primary pressure retaining welds in the control element drive mechanism structural components are consistent with the recommendations of Regulatory Gu ,de 1.31 as follows:

The delta ferrite content of A-No.8 (Table 2W-442 of the ASME Code,Section IX) austenitic stainless steel welding materials is controlled to 5FN-20FN.

1 The delta ferrite determination is carried out using methods D specified in the ASME Code,Section III, for each heat, lot or heat / lot combination of weld filler material. For the submerged arc process, the delta ferrite determination for each wire / flux comb:, nation may be made on a production or simulated (qualification) production veld.

4.5.1.5 cleanina and contamination Protection Procedures The procedure and practices followed for cleaning and contamination protection of the control element drive mechanism Q structural components are in compliance with the recommendations 273,l of Regulatory Guide 1.37 and are d%scribed L Gwc.lwI; below d M A -2.- MSN AA M /AS,r

' specific requirements for cleanliness and contamination protection are included in the equipment specifications for components fabricated with austenitic stainless steel. The provisions described below indicate the type of procedures utilised for components to provide contamination control during fabrication, shipment, and storage.

Contamination of austenitic stainless steels of the Type 300 series by compounds that can alter the physical or metallurgical structure and/or properties of the material is avoided during all stages of fabrication. Painting of Type 300 series stainless

( steels is prohibited. Grinding is accomplished with resin or Amendment D 4.5-5 September 30, 1988

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- Enclosure If l CESSAR !!n%.m. "'S' 8 '2 i

i k fbThe reactor;2coolant pumps may be disassembled, if necessary, for inspection.5 ,

Additional provisions for access, and details of the inservice spection program are included in the site-specific SAR.

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5.3.5 REACTOR COOLENT PRESSURE SOUNDARY (RCPS) tungE DST 30 TION SYSTEMS ,

Means for the detection of leakage from the Reactor Coolant  :

Pressure Boundary are provided to alert operators to the  ;

existence of leakage above acceptable limits, which may indicate i an . unsafe condition for the facility. The leakage detection

systems are sufficiently diverse and sensitive to meet the criteria of Regulatory Guide 1.45 for leaks from identified and unidentified sources. See Section 5.1.4 for interface requirements.

l 5.3.5.1 Leakace Detection Methods j 1  :

l 5.2.5.1.1 Unidentified Leakage 4 See site-sp'cifice SAR for details of determining unidentified g leakage. Interface requirements are contained in Section 5.1.4.

In addition to the methods for detecting unidentified leakage discussed in the site-specific SAR, a method for detecting large lB volume leaks, which is available as an integral part of the RCS and CVCS, is the reactor coolant inventory method. Leakage from ,

the Reactor Coolant System can be determined by not level changes in the pressurizer and in the volume control tank since the Reactor Coolant System and the Chemical and Volume Control System are closed systems. Since letdown flow and the reactor coolant l pump seal controlled bleedoff flow are collected and recycled back into the Reactor Coolant System by the Chemical and Volume control System.(CVCS), the not inventory in the Reactor Coolant System and CVCS under normal operating conditions will be '

constant. Transient changes in letdown flow rate or Reactor coolant System inventory can be accommodated by changes in the volume control tank level. Makeup flow rate provides a means of l detecting leakage from the Reactor Coolant System through measurement of the not amount of makeup flow to the system. The net makeup to the system under no-leakage steady state conditions should be zero. The makeup flow rates and the integrated makeup flow rates from the Reactor Makeup System and the refueling water tank are continuously monitored and recorded. Analysis of the integrated makeup flow recorders over a period of steady state operation can provide detection of abnormal leakage. An increasing trend in the amount of makeup required will indicate an abnormal leak which is increasing in rate. Leaks occurring

' Amendment B 5.2-21 March 31, 1988

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INSERT B .

5.2.4.2 Examination Cateaories and Methods The examination categories and methods are in accordance with Articles IWA 2000 and IWB-2000 of Section XI of the ASME Code. The acceptance standards are given in Article IWB-3000 of Section XI of the ASME Code.

5.2.4.3 Insoection Procram F

The scheduling of the inspection program complies with Subarticle IWA 2400 of l

Section XI of the ASME Code.

5.2.4.4 Hydrostatic Pressure Tests The pressure retaining Code Class I component leakage and hydrostatic pressure test program agrees with the requirements of Article IWB-5000 of Section XI of the ASME Code. {

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c l TABLE 5.2-2 i i

(Sheet 1of5)

REACTOR COOLANT SYSTEN MATERIALS C: r.ent Material $necification Reactor Vessel Forgings SA 508 C1 ass 2 and 3 Cladding Weld deposited austenitic stainless steel ,

with 5FN-18FN delta ferrite or Nicrfe 0  ;

L N, ule. 5;8e. End,, ..... ,, , , alloy (equivalent to SB 168)  ;

  1. CS tr h Reactor vessel head $8 166 CEDM Nozzles l Vessel internals (*) Austenitic Stainless Steel and NiCrFe alloy Feel cladding (a) Zircaloy 4 . i l

( Instrument nozzles SB 166 Control element drive mechanism housings .

Lower Type 403 stainless steel according to Code Case N 4-Il with end fittings to be l0

, SB 166 and/or SA 182 Type 348 stainless l steel Upper SA 479 and SA-213 Type 316 stainless steel ,

I with end fitting of SA 479 Type 316 and vent valve seal of Type 316 and vent

! valve seal of Type 440 stainless steel seat Closure head bolts SA 540 B24 or B23 Pressurizer Shell SA 533 Grade A or B Class 1 or SA-508 Class 3 1 Cladding')I Weld deposited austenitic stainless steel

  • with 5 FN-18FN delta ferrite or NiCrFe 0 alloy (equivalent to SB 166) l l

l Amendment D l September 30, 1988 L

Enclosure 11 Pap H M 20 CESSAR tinhno.

5.3.1.3 Special Methods for Mondestructive Rxamination Prior to, during, and after fabrication of the reactor vessel, i nondestructive tests based upon Section III of the ASME Boiler and Pressure Vessel Code are performed on all welds and forgings kg as indicated.

includin The nondestructive calibration methods, examination requirements instrumentation, sensitivity,

Ml reproduc bility of data, and acceptance standards are in accordance with requirements of the ASME B&PV Code,Section III.

n (see Table 5.2-1).

  • strict quality control is maintained in '

V critical areas such as calibration of test instruments.

All full-penetration, pressure-containing welds are 100%

radiographed to the standards of Section III of the ASME Boiler and Pressure Vessel Code. Wald preparation areas, back-chip areas, and final weld surfaces are magnetic-particle or <

dye-penetrant examined. Other pressure-containing welds, such as used for the attachments of nonferrous nickel-chromium-iron

.I mechanism housings, vents, and instrument housings to the reactor '

vessel and head, are inspected by liquid-penetrant tests of the root pass, the lesser of one-half of the thickness or each 1/2-inch of weld deposit, and the final surface. Additionally, ,

the base metal weld preparation area is magnetic-particle T

, examined prior to overlay with nickel-chromium-iron weld metal. ,

l All forgings are inspected by ultrasonic testing, using longitudinal beam techniques. In addition, ring forgings are tested using shear wave techniques.

All carbon-steel and low alloy forgings and ferritic welds are also subjected to magnetic-particle examination after stress relief.

All vessel bolting material receives ultrasonic and magnetic-particle examination during the manufacturing process.

The bolting material receives a straight-beam, radial-scan, ultrasonic examination with a search unit not exceeding 1 square-inch area. All hollow material receives a second ultrasonic examination using angle-beam, radial scan with a search unit not exceeding i square inch in area. A reference specimen of the same composition and thickness containing a notch (located on the inside surface) 1 inch in length and a depth of 3% of nominal section thickness, or 3/8-inch, whichever is less, is used for calibration. Use of these techniques ensures that no l materials that have unacceptable flaws, observable cracks, or sharply defined linear defects are used.

The magnetic-particle inspection is performed both before and after threading of the studs.

5.3-2

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Enclosure II Page 12 of 20 INSERT D These methods, procedures and requirements are compatible with Section XI of l the ASME Code so that results of preservice inspections can be correlated with

'l inservice inspections.

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. Enclosure II

CESSAR M e. .

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5.3.1.6 Reactor vessel Material surveillance. Program The irradiation surveillance program for System 80+ will be conducted to assess the neutron-induced changes in the RT  :

(reference temperature) and the mechanical properties of N  :

reactor vessel materials. Changes in the impact and mechanical properties of the material will be evaluated by the comparison of pre- and post-irradiation test results. The capsules containing  :

the surveillance test specimens used for monitoring the neutron-induced property changes of the reactor vessel materials will be irradiated under conditions which represent, as closely as practical, the irradiation conditions of the reactor vessel. '

ASTM E-185-82, Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and 10 CFR 50 B Appendix H, Reactor Vessel Material Surveillance Program Requirements, present criteria for monitoring changes in the

(/( toughnesspropertiesofreactorvesselbeltlinematerialsthroughl0 surveillance programs. The System 80+ reactor surveillance program adheres to all of the requirements of ASTM vessel

,l E-185-82 and satisfies the intent of 10 CFR 50, Appendix H. D j 5.3.1.6.1 Test Material selection

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Materials selected for the surveillance program are those judged _

most likely to be controlling with regard to radiation g embrittlement according to the recommendations of Regulatory Guide 1.99, Revision 2.

Surveillance test materials are prepared from the actual materials used in fabricating the beltline region of the reactor pressure vessel. The test materials are processed so that they are representative of the materials in the completed reactor i vessel. Specimens are prepared from three metallurgically different materials, including base metal, weld metal and "

heat-affected zone (HAZ) material.

In addition, material is included from a standard heat of ASME SA-533 Grade B Class 1 manganese-molybdenum-nickel steel made D available by the USNRC sponsored Heavy Section Steel Technology (HSST) Program. This standard reference material (SRM) is used as a monitor for Charpy impact tests, permitting comparisons among the irradiation data from operating power reactors and experimental reactors. Compilation of data generated from l

post-irradiation tests of these correlation monitors will be carried out by the HSST program.

Base metal test material is from a section of the shell course forging selected from the beltline of the reactor vessel.

Selection shall be based on an evaluation of initial toughness Amendment E 5.3-4 December 30, 1988

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Enclosure II

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This surveillance program'also satisfies the ASME Code.Section XI.  !

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fluence of the test specimens is expected to be approximately j j' 1.15 times higher than that seen by the adjacent vessel vall, so ,

i the measured changes in properties of the surveillance materials l E l will be able to predict the radiation induced changes in the D reactor vessel beltline materials.

The capsule assemblies are placed in capsule holders positioned

. circumferentially about the core at locations which include the regions of maximum flux. Figure 5.3-4 presents the typical exposure locations for the capsule assemblies.

All capsule assemblies are inserted into their respective capsule M 2*/

holders during the final reactor assembly operation. 4 5.3.1.6.6 Withdrawal Schedule The capsule assemblies remain within their holders- until the specimens contained therein have been exposed to predetermined D levels of fast neutron fluence. At that time, the capsule assembly is removed and the surveillance materials are evaluated.

The capsule assembly removal schedule and the associated target fluence are presented in Table 5.3-7.

The target fluence levels for the surveillance capsules are (e determined at the a::imuthal locations for the recommended E withdrawal schedule'of ASTM E-185-82, extended to a design life of 60 years (48 EFPY). The fluence values in Table 5.3-7 are accurate within +30%, -30%. The uncertainty is composed of errors in the calculational method and errors in the combined radial and axial power distribution.

Withdrawal schedules may be modified to coincide with those refuelingoutagesorplantshutdownsmostcloselyapproachingthelE withdrawal schedule. The two standby capsules are provided in the event they are needed to monitor the effect of a major core change or annealing of the vessel, or to provide supplemental toughness data for evaluating a flaw in the beltline.

Irradiation Effects Prediction Basis 5.3.1.6.7 Irradiation induced RT shift and reduction of upper shelf energy are predicted bayed on Regulatory Guide 1.99 Revision 2, p

" Effects of Residual Elements on predicted Radiation Damage to Reactor Vessel Materials." Predicted changes in RT and upper shelf energy are used to select the surveillan N materials (Section 5.3.1.6.1) and to formulate the initial heatup and cooldown limit curves for plant operation. Once actual post-irradiation surveillance data become available for each reactor vessel, these data will be used to adjust plant operating

(. limit curves.

Amendment E 5.3-11 December 30, 1988

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  • Enclosure II
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- Page 16 of 20 INSERT C The design also permits the remote installation of replacement capsule assemblies. The capsule holders are welded to the vessel cladding on the ,

inside surface, and the welds are subject to inspection according to the j requirements for permanent structural attachments as given in the ASME Code, ,

Sections !!! and XI. l

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. Enclosure-II 7CESSAR WWim.

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Internal surfaces of completed components are cleaned to produce an ites that is clean to the extent that grit, scale, corrosion products, grease, oil, wax, gun, adhered or embedded dirt, or extraneous material are not visible to the unaided eye.

Degreasing solvents, acetone or isopropyl alcohol, may be used on metallic surfaces. Water used for cleaning is inhibited with 30-100 ppa.hydraaine.

The specification for water quality ist Halides chloride, ppa <0.60 Fluoride, ppa <0.40 Conductivity, pahos/cm <5.0

. pH 6.0-8.0

-Visual clarity No turbidity, oil or sediment To prevent halide-induced intergranular corrosion which could occur in an aqueous environment with significant quantities of dissolved oxygen, flushing water is inhibited via additions of hydrazine. Many experiments conducted by C-E have proven this '

inhibitor to be affective.

Onsite and pre-operational cleaning of ESF components is in accordance with the recommendations of Regulatory Guide 1.37, 6.1.1.1.3.3 Cold-Worked Stainless steel Cold-worked austenitic stainless steel is not utilized for components of the ESF.

6.1.1.1.3.4 mon-Metallic Insulation All non-metallic insulation materials installed on stainless steel piping and equipment of the ESF are made of calcium silicate, expanded pearlite, fiberglass fiber, or similar materials (ASTM C533, C547, C553, C610, C612) and are consistent with the recommendations of Regulatory Guide 1.36.

6.1.1.1.4 Wald Fabrication and Assembly of Stainless Steel E8F Components The recommendations of Regulatory Guide 1.31 for the ESF compenents of the NSSE are followed as discussed below.

The delta ferrite content of A-No. 8 (Table QW-442 of the ASME Code,Section IX) austenitic stainless stwel welding materials, D except SFA-5.4 Type 16-8-2 and welding materials for weld metal Amendment D 6.1-4 September 30, 1988

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1 'e Enclosure II i Page 18 of 20

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overlay. cladding, used in the fabrication of components of the engineered safety features system, is controlled to 5FN-20FN.

l_ The delta ferrite content of each lot and/or heat of weld filler l

-metal used for welding of austenitic stainless steel code ,

components shall be determined for each process to be used in production. Delta ferrite determinations for consumable inserts, electrodes, rod or wire filler metal used with the gas tungsten

  • D are welding process, and deposits made with the plasma arc

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welding process may be determined by either of the alternative

i. methods of magnetic measurement or chemical analysis described in "

Section III of the ASME Code. Delta ferrite verification should be made for all other processes by tests using the magnetic measurement method on undiluted weld deposits described by i Section III of the ASME Code. The_ average ferrite content shall l- meet the acceptance limits of 5FN to 20FN for weld rod or filler i i

( metal.

For submerged arc welding processes, the delta ferrite determination for each wire / flux combination may be made on a production or simulated (qualification) production weld. lD

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6.1.2 ORGANIC MATERIALS

'5.1.2.1 Protective coatinus Many coatings which are in common industrial use may deteriorate in the post-accident environment and contribute substantial quantities of foreign solids and residue to.the containment sump.

Consequently, protective coatings used inside the containment, excluding components limited by size and/or exposed surface area, are demonstrated to withstand the design _ basis conditions and meet the intent of ANSI N101.2 (1972), " Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities,"

and recommendations of Regulatory Guide 1.54, " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants." Any chemical precipitation of appreciable size that does occur is trapped by the sump filter screen. The screen size is smaller than the line piping diameter, the shutdown cooling heat exchanger tube diameter, and the spray nozzle openings so that particles that could potentially block the system are filtered out. A list of surface coatings used D inside containment and their applicable conditions is given in Table 6.1-3.

6.1.2.2 Other Materials i

A listing of other materials in the containment is included in Table 6.1-4. The materials listed are not protective coatings applied to surfaces of nuclear facilities.

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Amendment D 6.1-5 September 30, 1988

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.. ~ Enclosure II

.o-- Page 19 of 20 INSERT F' 6.1.1.1.5 Ferritic Steel Welding The recommendations of Regulatory Guide 1.50, ' Control of Preheat Temperature for Welding of Low Alloy Steel' and Article D 1000, Appendix D,Section III of the ASME Code are followed.

Moisture control on low hydrogen materials shall conform to the requirements of the ASME Code and/or AWS D1.1, ' Structural Welding Code'.

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.4ES S AR 4!!!%.no. RN E%10 ]I 6.6 IM* SERVICE INSPECTION OF CLhSS 2 & 3 COMPONENTg  ;

6.6.1 00NPONENTS SUBJECT TO IIANIMATION See the site-specific SAR. D 6.4.3 &cCESSIBILITY  !

ASME code class 2 & 3 components are provided with access to-enable the performance of in-service examination, in-service-testing, and to meet the pre-service examination requirements set forth in Section XI.of the ASME Boiler and Pressure Vessel Code, as applied to the construction of the particular component.

46.3 Ep(AmtNA y{oe4 t.AT ssostIES - Anf D NE7 HODS

%e e.x. min.4<an Aorres and -bds are. In accgrdanc4. .

w$ Ah'cles TW C. ~ woo as 4 TW D-poo , indsd; ; 4 .

inspechen pr g m . % acc yw a dada-J.s are. $ W in )h.kc.las IWC-3 are and W0-3e of 6*bM E N A5mE h.

4 (o,(o.4 GVSTEm PRESSUR5 TESTS n.e. .sysk gessswee 4tsks & Cass % a.na 3 press ue<.

<ehmin iny ay a. w % Kstre.p'e m fr of Article.twc-swo anJ zwo-s* of see.h XI o f %

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Amendment D

/ 6.6-1 September 30, 1988 l

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