ULNRC-03916, Forwards Util 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CCS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Supporting Info,Encl

Forwards Util 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CCS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Supporting Info,Encl

ML20195B514
Person / Time
Site:	Callaway
Issue date:	11/09/1998
From:	Passwater A UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20195B524	List: ML20195B526 ML20195B530 ML20195B549 ML20195B558 ML20195B565 ULNRC-03916, Forwards Util 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CCS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Supporting Info,Encl
References
GL-98-04, GL-98-4, ULNRC-03916, ULNRC-3916, NUDOCS 9811160153
Download: ML20195B514 (112)

Text

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Union Bectric One Ameren Plaza 1901 Chouteau Avenue PO Box 66149 di. Louis, MO 63166-6149 314 621.3222 November 9,1998 United States h(uclear Regulatory Commission Attention: Document Control Desk Washington DC 20555-0001 Gent lemen: ULNRC-03916 b

Docket Number 50-483 k Callaway Plant Union Electric Company License Number NPF-30 Response to Generic Letter 98-04 dated July 14.1998

References:

1) NRC Generic Letter 98-04: Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment

2) EPRI TR-109937: Guidelines on the Elements of a Nuclear Safety-Related Coatings Program, dated April 1998 On July 14,1998, the Nuclear Regulatory Commission issued the referenced generic letter addressing issues which have generic implications regarding the impact ofpotential coating debris on the operation of safety related systems, structures, and components (SSC) dunng a postulated design basis Loss of Coolant Accident (LOCA). Protective coatings are necessary inside containment to control radioactive contamination and to protect surfaces from (;

crosion and corrosion. Detachment of the coatings from the substrate may make hF the Emergency Core Cooling System (ECCS) unable to satisfy the requirement of 10 CFR 50.46(b)(5) to provide long-term cooling and may make the safety-related Containment Spray System (CSS) unable to satisfy the plant-specific licensing

.O r i asis ofcontrolling containment pressure and radioactivity releases following a oD LOCA. The generic letter requests information under 10 CFR 50.54(f) v evaluate the addressees' programs for ensuring that Service Level 1 protective coatings inside containment do not detach from their substrate during a design basis LOCA and interfere with the operation of the ECCS and the CSS.

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9811160153 981109 %

PDR ADOCK 050004832 p PM r a subsidiary of Amoren Corporation 0 \ \DD\

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I United States Nuclear Regulatory Commission N:vember 9,1998

Page 2 AM=chment I contains AmerenUE 120-day response to the specific information requested in Generie Letter 98-04. No new commitments are contained in this letter.

Should you have any questions or need additional information concerning this matter, please contact us.

Very truly yours,

" #PP Alan C. Passwater Manager, Corporate Nuclear Services BFH/jdg i Attachments: 1) Generic Letter 98-04 Requested Information ,

l 2) Callaway Standard Plant FSAR, Section 6.2

3) EDP-ZZ-03000 " Containment Building Coatings"  ;

4) Specification A-1003(Q)" Surface Preparation and Application for Field Coatings".

5) APA-ZZ-00400 " Procurement ofParts, Supplies, Materials and Services".

6) WEP-ZZ-00015 " Preparation of Procurement Documents".

7) QCP-ZZ-03003 " Material Receipt Inspection".

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l STATE OF MISSOURI )

) SS CITY OF ST. LOUIS )

l Alan C. Passwater, of lawful age, being first duly sworn upon oath says that he is Manager, Corporate Nuclear Services for Union Electric Company; that he has read the i foregoing document and knows the content thereof; that he i has executed the same for and on behalf of said company with l full power and authority to do so; and that the facts therein stated are true and correct to the best of his l knowledge, information and belief.

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By

/ 4 'i K/V. v W/24i Alan C. Passwater Manager, Corporate Nuclear Services i

l SUBSC day of BEDA67 7W77' !a d sworn to before , 1998. me this  ;

hdblY VYZ

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1. 1 PATRICIAL REYNOLDS IngerMaus-4WEOFAAIBOUM ST.IdXascoumy 3amtmannagrumgg g m i

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! cc:- M. H. Fletcher w/o Professional Nuclear Consulting, Inc.

19041 Raines Drive Derwood, MD 20855-2432

-Regional. Administrator U.S. Nuclear Regulatory Commission i Region IV ,

611 Ryan Plaza Drive l Suite 400 Arlington, TX 76011-8064 L Senior Resident Inspector Callaway Resident Office.

U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mel Gray (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738

Manager,. Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 i

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Attachment 1 Generic Letter 98-04 Requested Information (1) A summary description of the plant-specific program or programs implemented to ensure that Service Level 1 protective coatings used inside the containment are procured, applied, and maintained in compliance with applicable regulatory requirements and the plant-specific licensing basis for the facility. Include a discussion of how the plant-specine program meets the applicable criteria of 10 CFR Part 50, Appendix B, as well as information regarding any applicable standards, plant-specific procedures, or other guidance used for: (a) controlling the procurement of coatings and paints used at the facility, (b) the qualification testing of protective coatings, and (c) surface preparation, application, surveillance, and maintenance activities for protective coatings. Maintenance activities involve reworking degraded coatings, removing degraded coatings to sound coatings, correctly preparing the surfaces, applying new coatings, and verifying the quality of the coatings.

RESPONSE

AmerenUE has implemented controls for the procurement, application, and maintenance of Service Level 1 protective coatings used inside the containment in a manner that is consistent with the licensing basis and regulatory requirements applicable to Callaway Plant. The requirements of 10 CFR Part 50 Appendix B are implemented through specification of appropdate technical and quality requirements for the Service Level I coatings program, which includes ongoing maintenance activities.

For Callaway Plant, Service Level l' coatings are subject to the requirements of, Regulatory Guide 1.54, ANSI N 101.2, N 101.4, and ANSI N 5.12 as stated in Callaway Plant's Standard Plant Final Safety Analysis Report (FSAR) Section 6.2 (Attachment 2). Excluded from these requirements, as provided in FSAR Section 6.2, are coatings applied on small equipment such as transmitters, small instruments, valves and electrical equipment. Adequate assurance that the applicable requirements for the procurement, application, inspection, and maintenance are implemented is provided by procedures and programmatic controls, which are approved under Callaway's Quality Assurance program.

(a) Service Level I coatings used for new applications or repair / replacement activities are procured from vendors with quality assurance programs meeting the applicable requirements of 10 CFR Part 50 Appendix B. The applicable technical and quality requirements that the vendor is required to meet are specified by AmerenUE in Callaway Plant Specification A-1003(Q)" Surface Preparation and Application ofField Coatings" (Attachment 4), in procurement documents and procedures, APA-ZZ-00400 " Procurement of Parts, Supplies, Materials and Services" (Attachment 5) and WEP-ZZ-00015 " Preparation of Procurement Documents" (Attachment 6). Acceptance activities are conducted in accordance with

' Our response applies to Service level I coatings used in containment that are procured, applied and maintained by AmerenUE orits contractor 1

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I procedure QCP-ZZ-03003 " Material Receipt Inspection" (Attachment 7) that is consistent with ANSI N 45.2 requirements (e.g., receipt inspection, source surveillance, etc.). These l procedures and specifications establish the technical and quality requirements, which when combined with appropriate acceptance activities provide adequate assurance that the coatings received meet the requirements of the procurement documents.

(b) The qualification testing of Service Level I coatings used for new applications or j repair / replacement activities inside containment are controlled by Specification A-1003(Q)

" Technical Specification for Surface Preparation and Application of Field Coatings" (Attachment 4) which meets the applicable requirements contained in the standards and regulatory commitments referenced above. These coatings, including maintenance coatings, I have been evaluated to meet the applicable standards and regulatory requirements previously referenced. Maintenance coatings were applied over actual paint samples removed from Callaway's containment building and tested in accordance with ANSI N101.2 per

specification A-1003(Q)in order to demonstrate that these coatings were qualified for Service Level 1 application.

(c) The surface preparation, application and surveillance during installation of Service Level 1

coatings used inside containment are controlled by Callaway Plant Procedure EDP-ZZ-03000

( " Containment Building Coatings"(Attachment 3) and Specification A-1003(Q)(Attachment l 4). This procedure and specification control coating type, application conditions, surface i preparation, application, painter qualifications, QC Inspector qualifications and inspections l points to assure Service Level I coatings meet the applicable portions of the standards and l regulatory commitments referenced above. Documentation of completion of these activities is performed in accordance with EDP-ZZ-03000, consistent with the applicable requirements.

l 2) Information demonstrating compliance with item (i) or item (ii):

l (i) For plants with licensing-basis requirements for tracking the amount of unqualified coatings inside the containment and for assessing the impact of potential coating debris on the operation of safety-related SSCs during a postulated design basis LOCA, the following information shall be provided to demonstrate compliance:

(a) The date and findings of the last assessment of coatings, and the planned date of the next assessment of coatings.

(b) The limit for the amount of unqualified protective coatings allowed in the l

containment and how this limit is determined. Discuss any conservatism in the l method used to determine this limit.

(c) If a commercial-grade dedication program is being used at your facility for dedicating commercial grade coatings for Service Level 1 applications inside the containment, discuss how the program adequately qualifies such a coating i for Service Level I service. Identify which standards or other guidance are

currently being used to dedicate containment coatings at your facility; or, 2

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RESPONSE

l AmerenUE is not committed to track the amount of unqualified coatings inside containment.

l Callaway plant does, however, require the use of qualified safety related coatings inside l containment. The use of qualified coatings is controlled by plant procedure EDP-ZZ-03000 l (Attachment 3) and screening questions on " License Impact Reviews". When qualified coatings l are not practical, an engineering evaluation is performed to determine the impact the unqualified coatings would have on containment based on location within containment and flow path to the l sumps. All unqualified coatings are assumed to fail and are tracked on an " Unqualified Coatings l Log" To date we have a total of 6,318 square feet of unqualified coatings inside containment of l which 5,400 square feet was applied during plant construction. Each coating application was i evaluated on a case by case basis and the total amount ofunqualified coating installed was found to be acceptable.

l (ii) For plants without the (licensing-basis requirementsfor tracking the amount of unqualilled coatings inside the containment andfor assessing the impact ofpotential l coating debris on the operation ofsafety-related SSCs during apostulated design basis LOCA), information shall be provided to demonstrate compliance with the requirements of 10CFR50.46b(5),"Long-term cooling" and the functional capability l of the safety-related CSS as set forth in yourlicensing basis. If a licensee can l demonstrate this compliance without quantifying the amount of unqualified

! coatings, this is acceptable.

l l RESPONSE:

The following description and referenced materials describe the licensing basis for Callaway i Plant relative to conformance with 10 C.F.R. 50.46(b)(5), "Long-term cooling," specifically with regard to Callaway Plant's ability to provide extended decay heat removal including related assumptions for debris that could block containment emergency sump screens:

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• Callaway Plant's Standard Plant Final Safety Analysis Report (FSAR), Section 6.2 (Attachment 2) l l

• Callaway Plant's containment sumps conform to Regulatory Guide 1.82, Revision 0 " Sumps for Emergency Core Cooling and Containment Spray Systems," as described in Callaway l Plants Standard Plant FSAR, Table 6.2.2-1 (Attachment 2). Under this commitment l Callaway Plant has assumed that the systems that draw from the sumps for emergency core I cooling and containment spray systems may experience sump blockage of up to 50% of the effective sump area from debris generated as a result of a LOCA. At the time Callaway Plant was licensed, no distinction was drawn between the various potential sources for post-LOCA debris; these systems were intended to function, even with debris partially obstructing the sumps, from whatever source derived. The analyses submitted as part of the licensing basis

for Callaway Plant demonstrates that even with this blockage, the emergency core cooling i

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and containment spray systems will continue to provide sufficient cooling flow to fulfill the long-term cooling functions required to conform with 10 C.F.R. 50.46(b)(5).

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. The NRC accepted these analyses and these systems as meeting the requirements of 10 C.F.R

!- 50.46(b)(5) in the Safety Evaluation Report (SER) related to the operation of Callaway Plant, Unit No.1 (Docket STN 50-483) in Section 6.2.2.

The licensing basis for Callaway Plant, as accepted by the NRC, in the FSAR, provides both the regulatory and safety basis for safety system performance. Coatings are not treated separately in the licensing basis for Callaway Plant because the sump screen blockage assumption does not

' distinguish among the source terms for the LOCA-generated debris. Accordingly, a separate l demonstration of the regulatory and safety basis for safety nytem performance is not required. 1 The following information shall be provided

(a) If commercial-grade coatings are being used at your facility for Service Level 1 applications, and such coatings are not dedicated or controlled under your Appendix B Quality Assurance Program, provide the regulatory and safety basis for not i

controlling these coatings in accordance with such a program. Additionally, explain l why the facility's licensing basis does not require such a prograv.  !

RESPONSE

AmerenUE does not currently employ commercial grade dedication for Service Level I coatings

{ used inside containment at the Callaway Plant.

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fa ATTACHMENT 2

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CALLAWAY - SP l I CHAPTER 6.0

. ENGINEERED SAFETY FEATURES l-l Engineered safety features (ESF) are those safety-related systems and components 1 designed to directly mitigate the  ;

consequences of.a design bssis accident by: 1 l

a. Protecting the fuel cladding

b. Ensuring the containment integrity i

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c. . Limiting fission product releases to tdue environment  ;

within.the-guideline values of 10 CFR, Part 100 i The limiting design basis accidents which are discussed and analyzed in Chapter 15.0 and Section 6.3 are:

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a. Loss-of-coolant accident (LOCA)

b. . Main steam line break (MSLB) c.. Steam generator tube rupture

d. Fuellhandling accident

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2 The engineered safety. features consist of the following systems:

I a. Containment (Section 6.2.1) l-

b. Containment heat removal (Section 6.2.2) l

c. 'Contair:aent isolation (Sections 6.2.4 and 6.2.6)'

d. Containment combustible gas control (Section 6.2.5)

e. Emergency core cooling (Section 6.3)

'f. Fission product removal and control systems (Section 6.5)

g. Emergency HVAC and filtration (Section 9.4)

h. Control room habitability (Section 6.4)

. i. Auxiliary feedwater (Section 10.4.9) i ,

I The-containment is provided to contain radioactivity following a LOCA.

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( kr CALLAWAY - SP The containment spray system, in conjunction with the contain-ment fan coolers and the emergency core cooling system, is designed to remove sufficient heat from the containment atmos-phere following a LOCA or main steam line break inside the containment to rapidly reduce the containment pressure and temperature and maintain them at acceptably low levels.

The containment spray system is also designed to minimize the iodine and particulate fission product inventories in the containment atmosphere resulting from.a postulated LOCA.

Containment isolation is provided to minimize leakage from the containment. Steam line and feedwater line isolation is i provided to minimize the heat removal from the reactor coolant system and prevent excessive blowdown of a steam generator following a postulated main steam line rupture. Steam line isolation will also prevent excessive radioactivity release following a steam generator tube rupture. The containment purge isolation capability is provided to reduce the radio-iodine released following a fuel handling accident inside the containment.

Hydrogen recombiners prevent the accumulation of combustible mixtures of hydrogen and oxygen following a LOCA.

The emergency core cooling system (ECCS), consisting of accu-mulator tanks, safety injection pumps, RHR pumps, and centrif-ugal charging pumps, is provided for emergency core cooling to limit fuel damage following a LOCA or main steam line break.

An emergency exhaust system is provided to reduce the radio-iodine released following a fuel handling accident outside the containment and to filter ECCS leakage outside the containment following a LOCA.

The auxiliary feedwater system provides an adequate amount of feedwater into the steam generators to prevent a pressure transient which could cause a loss of reactor coolant through the pressurizer relief valves and a possible uncovering of the reactor core following a main steam line break or loss of the main feedwater system.

Other safety-related systems are identified in Section 3.2.

s Because of the importance of safety-related systems to the health and safety of the general public, special precautions are taken to ensure high quality in the components and in the

system design and to ensure reliable and dependable operation.

Rev. OL-0 6.1-2 6/86

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i- 6 '.1 ENGINEERED SAFETY FEATURE MATERIALS This section provides a discussion of the materials used in the fabrication of engineered safety feature components and of the material interactions that could potentially impair the operation of the ESF.

6.1.1 METALLIC MATERIILS 6.1.1.1 Materials Selection and Fabrication Information on the selection and fabrication of the materials in the engineered safety features of the plant, such as the emergency core cooling systems, thc containment heat removal systems, the containment combustible gas control system, and the containment spray system, is provided below. Materials for Et use in the ESF are selected for their compatibility with the 2:q reactor coolant system and containment spray solutions, as required by Section III of the ASME Boiler and Pressure Vessel 4 Code, Articles NC-2160 and NC-3120.

, 6.1.1.1.1 Specifications for Principal Pressure-Retaining

, Materials

.=3 39 All pressure-retaining material in the engineered safety

., feature systems' components complier with the corresponding

, ,j material specification permitted by'ASME Section III, Division 1.

- i The material specifications for pressure-retaining material in each component of the engineered safety feature systems will meet the requirements of Article NC-2000 of ASME Section III, Class 2, for quality group B and Article ND-2OOO of ASME Section III, Class 3, for quality group C components. Contain-ment penetration materials will meet the requirements of Article NE-2300 of ASME Section III, Division I. Table 6.1-1 includes the specifications for the principal pressure-retaining components.

6.1.1.1.2 Engineered Safety Feature Materials of Construction

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The engineered safety feature materials that would be exposed to the emergency core cooling water and containment sprays following a LOCA are indicated in Table 6.1-1. These materials are chosen to be compatible with the core cooling and spray solutions. Additional information concerning metallic ma-terials' compatibility with post-LOCA conditions is provided in Reference 1.

In order to keep materiais within the containment that are subject to corrosion to a minimum, the following restrictions Rev. OL-O 6.1-3 6/86

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.are placed on the use of zinc, aluminum, and mercury in the containment:

a. Aluminum is severely attacked by the alkaline con-tainment spray solution. This reaction may result in the loss of structural integrity and the generation of gaseous hydrogen. The use of aluminum in the con-tainment is minimized.

b. Boric acid reacts with zinc, oxidizing it and libera-ting hydrogen gas. The use of zinc (galvanized 1

materials and paint) in the containment is minimized to reduce the generation of hydrogen.

c. The use of mercury and mercuric compounds is minimized inside the containment because of its corrosive effects on stainless steel, NiCrFe alloy 600, and The amount of mercury

'" alloys containing copper.

-associated with plant lighting and control switches, etc., is negligible.

Figure 6.2.5-2 shows the maximum allowable quantities of zinc 2-and aluminum inside the containment building. Corrosion. rates for zinc and aluminum are given in Table 6 2.5-4. Use of-aluminum and zinc inside containment is minimized to the extent practicable.

f For other materials which could come in contact with contain-ment sprays, tests have been performed and are~ detailed in

Reference 2. These tests have shown that no significant amount

! of. corrosion products will be produced from these materials.

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 in significant quantities are demonstrated to withstand the design basis accident conditions and are designed l- to meet the criteria given in ANSI N101.2 (1972), " Protective L Coatings (Paints) for Light Water Nuclear Reactor Containment l

Facilities," and are in compliance with Regulatory Guide 1.54, L " Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," as indicated in Table 6.1-2. Some small items may be painted or coated using common -

industrial practice but the paint / coating will not be in l

sufficient quantity to cause any clogging problems for the sump screens. Any precipitation of appreciable size that occurs ,

either settles out prior to reaching the sump screens or is trapped by the sump filter screen. The screen opening size

'(1/8 inch) is smaller than the line piping, the RHR heat Rev. OL-5 6.1-4 6/91 ,

, , CALLAWAY - SP exchanger tubes, the spray nozzles,'and clearances in the reactor core. Therefore, particles which could potentially i cause blockage are filtered out. Refer to Section 6.2.2.1 for  !

a discussion of the sump design and consideration given to screen clogging. For each containment component, a complete  ;

list of the surface coatings, the dry film thickness, and the l surface area covered is presented in Table 6.1-3. i 6.1.1.1.3 Integrity of Safety-Related Components

. The following information is provided to demonstrate that the integrity of the safety-related components is maintained during all stages of component manufacturing:

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a. Regulatory Guide 1.44, Control of the Use of

, Sensitized Stainless Steel, is complied with to the extent specified in Table 6.1-4 for the purpose of )

-4 avoiding significant sensitization and stress 4 corrosion cracking in austenitic stainless steel components of the engineered safety features. )l 1

J~ b. Cleaning and contamination protection of austenitic 5, stainless steel components of the engineered safety

'. features complies with Regulatory Guide 1.44, Control l S' of the Use of Sensitized Stainless Steel, as described

1. Il in Table 6.1-4. Regulatory Guide 1.37, Quality s Assurance Requirements for Cleaning of Fluid Systems

'" and Associated Components of Water-Cooled Nuclear L.;. -) Power Plants, is complied with to the extent specified Si in Table 6.1-5. ,

5 c. Cold worked austenitic stainless steel material with s

0.2-percent offset yield strengths greater than 90,000  :

j' psi are not used in components that are part of the t 1.. engineered safety features.

d. The selection, procurement, testing, storage, and installation of all nonmetallic thermal insulation assure that the leachable concentrations of chloride, fluoride, sodium, and silicate are in accordance with Regulatory Guide 1.36, Nonmetallic Thermal Insulation

. j for Austenitic Stainless Steel, with clarifications as j/ discussed in Table 6.1-6.

e. With regard to the preheat temperature used for welding low alloy steels, the recommendations of Regulatory Guide 1.50, Control of Preheat Temperatures l for Welding of Low Alloy Steel, were followed, as 1 discussed in Table 6.1-7.

f. The recommendations of Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility, are ,

followed as discussed in Table 6.1-8.

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g. In order to determine the RT NDT for the steam and feedwater system materials, the guidelines in NRC Branch Technical' Position MTEB 5-2 Section 1.1, i Article 4 were followed. .

The applied test methods and acceptance criteria for j all materials used in the steam and feedwater systems, 1 with the exception of the steam generators, comply completely with ASME Code Section III, Article NC-2310 of the Winter 1974' Addenda for fracture toughness of ferritic materials used in Class 2 components. The .

applied test methods and acceptance criteria for all Class 2 steam generator materials comply with the requirements of ASNE Code Section III 1971 Edition ,

l through Summer 1973 Addenda.

6.1.1.1.4 Control of Stainless Steel Welding Regulatory Guide 1.31, Control of Stainless Steel Welding, is complied as ,

t supplemented by Branch Technical Position MTEB 5-1, with to the extent specified in Table 6.1-9 for the purpose of l avoiding fissuring in.austenitic stainless steel welds that are part of the engineered safety features.

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6.1.1.2 Comoosition. Comoatibility. and Stability of Containment and Core Soray Coolants The information given below is provided on the composition, ,

compatibility, and stability of the core cooling water and the  :

containment sprays on the engineered safety features.

6.1.1.2.1 Control of pH During a Loss-of-Coolant Accident

• A description of the method of establishing containment spray and recirculation sump pH following a LOCA is included in  !

Sections 6.2.2 and 6.5. The resultant basic equilibrium pH, which is greater than or equal to 7.1, is not conducive to 1

stress-corrosion cracking in austenitic stainless steels. i' 1 Hydrogen evolution is discussed in Section 6.2.5, Combustible Gas Control in Containment.

6.1.1.2.2 Engineered Safety Feature Coolant Storage i The borated water supply for the containment sprays and emergency core cooling system is drawn from the refueling water i storage tank. As described in Section 6.3, the refueling water storage tank is fabricated of stainless steel and is not  ;

subject to significant corrosive attack by the tank's contents.

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. CALLAWAY - SP 1, The accumulator tanka which store borated water for the j accumulator safety injection system are made of carbon steel

's and are clad with sta,inless steel to ensure that they are

p -resistant to corrosion.

J' 6.1.2 ORGANIC MATERIALS -

m. Use of organic material inside the containment is kept to a minimum.

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ce The amount of lubricants inside the containment which is subject to being released to the containment is listed in Table 6.1-10. The lubricants, such as those needed for the reactor i coolant pumps and hydraulic snubbers, are, however, totally enclosed and not open to the containment atmosphere.

Table 6.1-3 is a coating schedule for the containment which

,. indicates the type of paint and compliance with Regulatory ,

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Guide 1.54.

[ All protective coat ngs covered by Regulatory Guide 1.54 which are applied to surfaces within the containment have been tested to demonstrate that they will remain intact during postulated LOCA conditions. The tests are performed by an independent laboratory and show that no significant decomposition, radiolytic or pyrolytic failures will occur during a DBA.

s Where the surface area and application type do not dictate j

) special coatings, the coatings are evaluated by generic-type '

and formulation information. Paint chip formation is controlled by limiting the thickness of nonqualified coatings 1 to a point where there is insufficient tensile strength in a 4 removed film to farm a chip.

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6.1.3 POST-ACCIDENT CHEMISTRY Following a main steam line break or design basis LOCA, 1

j trisodium phosphate and boric acid solutions will be present in l the containment sumps. Table 6.5-5 indicates the quantities of I trisodium phosphate and boric acid that will be present in the  !

containment after an accident. The pH control reduces the probability of chloride stress corrosion cracking on stainless

, ) steel and attack on aluminum fittings. The long term, equilibrium pH of the sump fluid will be greater than or equal to 7.1 following complete dissolution of the stored trisodium phosphate.

6.

1.4 REFERENCES

1. Whyte,_D. D. and Picone, L. F., " Behavior of Austenitic Stainless Steel in Post Hypothetical Loss-of-Coolant Environment," WCAP-7798-L (Proprietary), November 1971 and WCAP-7803

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(Non-Proprietary), December 1971.

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2.  ; Picone, L. F., " Evaluation of Protectiv$ Coatings for use.in Reactor Containment," WCAP-7198-L

' (Proprietary), April 1968 and WCAP-7825-(Non-Proprietary), December 1971.-

3. Caplan,' J. S., "The Application of Preheat Temperatures after Welding Pressure Vessel Steels,"

WCAP-8577-(Non-Proprietary), September 1975.-

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• TABLE 6.1-2 i DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATURY CUIDE 1.54 REVISION O, DATED JUNE 1973, TITLED " QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS" '

Regulatory Guide Position en' Position on 1.54 Position Non-NSSS Components; . NSSS Components

1. ANSI N101.4-1972 should be 1.' Complies. 1, 2, 3 and 4. NSSS equipment located '

used in conjunction with ANSI '

in the containment building is separated N45.2-1971, " Quality Assurance into four categories to identify the Program Requirements for Nuilear applicability of this regulatory guide a Power Plants." to various types of equipment. These categories of equipment are as follows:  :

2. Subdivision 2.7 of ANSI N101- 2. Complies.

4-1972 states that when references

• are made to other standards, these t references shall imply the most a. Category 1 - Large equipment recent or current editions of the b. Category 2 - Intermediate equipment '

referenced standards. The specific c. Category 3 - Small equipment applicability or acceptability of -- d. Categcry 4 - Insulated / stainless  ;

referenced standards will be steel equipment  !

covered separately in other regula- "

k tory guides, where appropriate. _ A discussion of each equipment category I follows:

3. Subdivision 1.1.2 of ANSI 3. Complies, except N101.4-1972 states that quality that for certain a. Category 1 - Large Equipment assurance, as covered by this applications within?, .

standtrd, comprises all those  !

the containment. The Category 1 equipment consists planned and systematic actions where the coating - '

of the following:

• necessary to provide specified is not necessary documentation and adequate con- for the protectionc (1) Reactor coolant system supports  :

fidence that shop or field of the component, (2) Reactor coolant pumps (motor and I coating work for nuclear a quality assurance. motor stand)  !

facilities will perform program is not (3) Accumulator tanks  !

satisfactorily in service. i applied. In those . (4) Refueling machine This statement shcr.ld not applications, the v be interpreted as implying coating is reviewed._ Since this equipment has a large that the end product of to assure that -surface area and is procured from [

quality assurance actions is there are no long- . only e few vendors, it is possible  !

the production of specified term detrimental e to implement tight controls over t documentation. The term e ffects . - -

. these items. Stringent requirements t

" quality assurance," as used -'

are specified for protective coatings ,

in ANSI N101.4-1972, should on this equipment through the use of  !

be considered to comprise a painting specification in the '

all those planned and sys- - -

procurement documents. This specification  !

tematic actions necessary to defines requirements for-provide adequate confidence s '

that shop or field coating (1) Preparation of vendor procedures j work for nuclear facilities *

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Use of specific coatings systems in Subdivisions 7.4 7.8, records and documents listed through which are qualified to ANSI N101.2 i

^ and included.in the (3) Surface preparation standard, only. are suggested forms Alternate consistent with the documentation require- (4)

Application of the coating systems ments of Appendix B to 10 CFR in accordance with the paint Part 50 is also acceptable. considered manufacturer's instructions (5)

4. Inspections and nondestructive Sections 3 and 4 of ANSI examinations N101.4-1972 4.

, assurance requirements fordelineate quality Complies (6) coating materials and surface Exclusive of certain materials j preparation of substrates. (7) i Cleaning materials used with (8)

Identification of all nonconformances stainless steel would not be Certifications of compliance compounded from or treated with chemical compounds con- The vendor's procedures are subject to taining elements that could review by engineering personnel, and the i contribute to corrosion, vendor's implementation of the specification intergranular cracking, or requirements is monitored during quality stress corrosion cracking. assurance surveillance activities.

Examples of such chemic compounds are those conal This system of controls provides assurance i taining chlorides that the protective coatings will properl 1

lead, zinc, copper,, fluorides, sulfur, adhere to the base metal during prolongedy or mercury where such elements exposure to a post-accident environment are leachable or where they present within the containment building.

! could be released by breakdown b.

of the chemical compounds under Category 2 - Intermediate Equipment

[ expected environmental condi-tions (e.g., by radiation). The Category 2 equipment consists of the following:

(' This limitation is not in- '

tended to prohibit the use of (1) trichlorotrifluoroethane Military Specification which (2) Seismic platform and tie rods (3) Reactor internals lifting rig MIL-C-81302b for cleaning or degreasing of austenitic (4) Head lifting rig Electrical cabinets stainless steel provided adequate removal is assured. Since these items are procured from a large number of vendors, and individual surface areas,ly have very small it is not practical ~"

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Regulatory Guide Position on Position on 1.54 Position Non-NSSS Components NSSS Components to enforce the complete set of stringent requirements which are applied to Category 1 items.

Another painting specification is used in these procurement docu-ments. This specification defines to the vendors the requirements for:

(1) Use of specific coating systems which are qualified to ANSI N101.2 (2) Surface preparation (3) Application of the coating systems 4 in accordance with the paint manu-facturer's instructions The vendor's compliance with the requirements is also checked during quality assurance surveillance activities in the vendor's plant.

These measures of control provide a high degree of assurance that the protective coatings will adhere properly to the base metal and with-stand the postulated accident environ-ment within the containment building.

c. Category 3 - Small Equipment Category 3 equipment consists of the following:

(1) Transmitters $

(2) Alarm horns (3) Small instruments (4) Valves (5) Heat exchanger supports These items are procured from-several different vendors and are painted by the vendor in accordance with conventional industry practices. Because-the total exposed surface l area is very small, Westinghouse does not specify further requirements.

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. 1.54 Position ._. Position on Non-NSSS Components Position on ..

.NSSS Components s-

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d.

-Category 4 - Insulated or Stainless Steel Equipment.

Category 4 equipment consists

-v -

of the following:

(1) Steam generators - covered with wrapped insulation-(2)' ~

Pressurizer wrapped ~ covered with insulation-(3) Reactor pressure vessel -

covered insulationwith - rigid reflective (4) Reactor cooling piping <

stainless steel (5)

Reactor coolant pump casings -

stainless steel Since Category 4 equipment is insulated '

or is- stainless steel, no painted surface Therefore, areas are exposed within the containment.

applicable for Category 4 equipment.this regulator l

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TABLE 6.1-3 CONTAINMENT COMPONENTS - COATING SCHEDULE o

s j Uncoated

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a

, oc<m n e m:o d# M c-* Estimated - Estimated Item / Type / 9O

-5 dM $$

d j Q" Total Film Area Generic - . Thickness ' Shop Field .(Square Category Description 84 C8 I: M O Type (1)

H (mils) Applied Applied Feet)

Carbon steel Containment - dome X Inorganic zinc 2-4 X Touch-up 31,000 liner plate Containment - walls ;X Inorganic zinc 2-4 X Touch-up- '59,000' Structural steel ~

Heavy support steel X Inorganic zinc 2-4 X Touch-up 182,300 Miscellaneous steel X Inorganic zine 2-4 .X Touch-up 16,500 Gratings X 43,700 .

, t 3 '

Elevator Metal siding X 8,50'O.

Tanks and pools Accumulator tanks X Epoxy 4-5 X ' Touch-up 5,200 Refueling pool X N/A Reactor coolant A N/A drain tank e Carbon steel pipe, Pipe X X .

N/A <  :

hangers, valves, Pipe X Inorganic zinc 2-4 X Touch-up 9,100- .t

[ Pipe supports X Inorganic zinc I2-4 X Touch-up 25,500 Valves and valve X Alkyd / red oxide 2.5-4 X Touch-up 3,500 .

actuators Mechanical Polar crane X Inorganic zinc 4-7 X .36,700 equipment Pumps (RCPs) X Epoxy 2-4 X Touch-up 3,000 (including Tang Nnd fan hous- X Epoxy 7.5-11 X 1,200 driver) ings (carbon steel) X Epoxy 7.5-11 X 400 t HVAC ducting X 14,000 (6)

HVAC ducting X '

N/A Steam generators X ., 15,200 ,

Hydrogen recombiners X g N/A y Containment coolers X Epoxy 10 , <

X. 5,400 l Containment coolers :X _, 1,100 Heat exchangers X Epoxy . 2-4 X Touch-up 300 5

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.E a M >g$ Estimated Total Film Estimated Area Item / Type / gu d j Q E Generic Thickness Shop Category Description Field (Square 00 E m o M Type (1) (mils) Applied Applied Feet)

Electrical Motor control centers X Alkyd / red oxide 1-2.5 X 500 Terminal boxes X Control panels 600 X Epoxy 1.75-3 X . 1,000 Raceways, conduit, X cable trays, and 38,400 (6) supports Concrete and Floor, cove, and X Epoxy (2 )- 12 masonry X 12,900 (3) wainscot NOTES:

(1) Generic coating systems acceptable for containment use are selected from suppliers who are prequalified to Bechtel standards and test criteria. Other coating systems may be shown to be acceptable based on individual analyses.

(2) Concrete, if painted, will be painted with epoxy surfacer or epoxy coating system.

(3) The wainscot extends 12 inches above the floor and is painted the same as described in Note 2, then top coated with 8 to 10 mils of epoxy-based paint.

(4) Deleted (5) Deleted 4

(6) Estimated area includes a limited amount of unqualified touch-up coating.

Rev. OL-0 6/86

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TABLE 6.1-4  ;

DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.44, REVISION O, DATED MAY 1973, TITLED " CONTROL OF THE USE OF SENSITIZED.

STAINLESS STEEL" Regulatory Guide Position on Non- Position on-1.44 Position NSSS Components NSSS Components Unstabilized, austenitic stain-less. steel of the AISI Type 3XX series used for com-ponents that are part of (1)the reactor coolant .

pressure boundary. (2) systems required for reactor shutdown, (3) systems required for emergency core cooling, and (4) reactor vessel internals that are relied upon to permit adequate core cooling for any mode of normal ,

operation or under credible 'b t postulated accident conditions should meet the following:

1. Material should be suitably 1. Complies. 1. Complies, as discussed i protected against contaminants in Section 5.2.3.4.1.

capable of causing stress corro-sion cracking during fabrication, i*

shipment,' storage, construction, '

testing, and operation of compo- ,

nents and systems. r l

2. Material from which compo- 2. Complies. 2. Complies, as discussed  !

nents and systems are to be fabri- . in Section 5.2.3.4.2.

icated should be solution heat ,

treated to produce a nonsensi-tized condition in the material.

3. All austenitic stain- 3. Complies,:as discussed in  !

3. Nonsensitization of the r material should be verified less steels are - Section 5.2.3.4.3.

using ASTM A 262-70, " Rec- furnished in the solu-commended Practices for Detecting tion annealed and water-Attack in Stainless Steel," quenched condition.

Practices A or E, or another Since susceptibility to method that can be demonstrated stress corrosion crack-to show nonsensitization ing in this condition is minimal, testing is  ;

in austenitic stainless steel.

Test specimens should be selected not performed. +

from material subject to each different heat treatment practice and from each heat.  ;

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4. NSSS Components _ Position on Material subjected to sen- NSSS Components _

temperature in the range 4.

During fabrication and of 800 to 1500 F, subsequent to installation, austenitic 4. Complies, as discussed solution heat treating in accor- stainless steels are not in Section 5.2.3.4.4.

dance with Subparagraph C.2. pemitted to be exposed above and testing in accordance to 800-1500 temperatures in the range with Subparagraph C.3, above, of F, except for should be L Grade material; welding. Welding practices that is, it should not have a are controlled to minimize carbon content greater than sensitization, as discussed 0.03 percent. Exceptions are: in Position 5 below.

a. Material exposed to reactor coolant which has a

' ~

controlled concentration of less than 0.10 ppm dissolved oxygen at all temperatures above 200 F during normal operation, or b.

Material in the form of castings or weld metal with a percent; 5 ferrite content or of at least c.

Piping in the solution annealed condition whose exposure to temperatures in the range of 800 weldingto 1500 F has been operations, limitedit to provided is of sufficiently small diameter so iblethat in the event postulated failure of aofcred-the piping during norinal reactor operation, the reactor can be shut down orderly and cooled down in an manner, is provided by the reactorassuming makeup coolant makeup system only.

5. Material subjected to sensitizing temperatures in the 5.

Heat treatment of auste.nitic range of 800 to 1500 F during heat stainlass steel in the 5. Complies,. as discussed treating or processing other temperature range 800 to in Section 5.2.3.4.5.

than welding, subsequent to 1,500 F is not permitted.

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Regulatory Guide Position on Nsn 4 Poaitica en NSSS Components' .z 1.44 Position. NSSS Components

-s.ccordance with Subparagraph performed at the' solution C.2. above, and testing in annealing temperature, cccordance with Subparagraph followed by an immediate-C.3. above, should be retested water quenching. If hot in accordance with Subparagraph bending is performed.at **

C.3 above, to demonstrate that some temperature.other it is not susceptible to than the solution annealing the pipes intergranular attack, ext.ept . are temperature, re-solution annealed ~and that retest is not required for:

water quenched. Since Cast metal or weld sensitization is avoided,

a. testing to determine metal with a ferrite content susceptibility to in-of 5 percent or more: or tergranular attack is . , .

b. Material with a carbon not performed.

content of 0.03 percent or less that is subjected to tempera-tures in the range of 800 to 1500 F for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or

c. Material exposed to special processing, provided the processing is properly controlled to develop a uniform product and provided that adequate doc-umentation exists of service experience and/or test data to demonstrate that the processing will not result in increased susceptibility to intergranular stress corrosion.

Specimens for the above retest should be taker. from each heat of material and should be subjected to a thermal treat-ment that is representative of , ,

the anticipated thermal con-i ditions that the production .

material will undergo.

6. Welding practices are con- 6. Complies, as discussed

6. Weldingmaterial practices and,i-if trolled to minimize sensitiza- in Sections 5.2.3.1, 5.2.3.2.2, necessaryld compos be controlled to tion in the heat-affected zone 5.2.3.3.2 and 5.2.3.4.

tion shou of unstabilized austenitic avoid excessive sensitization stainless steels, as described of base metal heat-affected below.

zones of weldments. An 1 Rev. OL-5 6/91

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NSSS Components -

on Non -

such as specified in Sub-intergranular corrosion test, Position on paragraph C.3 above NSSS Comoonents

-be performed for eac,hshould a. Weld _ Heat Input' twelding procedure to be ' Heat used for 0.03 percent. input during weld-the size of electrodes foring is controlled by limiting

' the shielded metal are and gas

'welding.

tungsten arc processes and thebead thickness for submer are not permitted,Other welding processes b.

iInterpass Temperatures-

. welding are-controlled so asInterpass temperatures during not.to exceed 350 F.

c.

Ferrite Content Stainless steel welding materials are in the furnished range ofwith 8 to a25ferrite content percent for

- t p e 308 and 308L welding materials and 5309 316L to 15 percent for type 316, rials,.

Additional discussionand 309L welding mate-regarding compliance to Replatory

-Guide 1.31 is provided in Table 6.1-9.

d. -

Postweld Heat Treatment Postweld heatintreatment temperatures excess of at: '

350 F are not permitted i

unless a full-solution anneal and water quench are performed. }

The above welding practices  ;

.are sufficient to ensure-that unacceptable s

the base metal sensitization of does not occur; heat therefore, affected the  !

is not performed.intergranular corrosion' testing E

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$b p 6.2 CONTAINMENT SYSTEMS 3 .

-The containment systems include the containment, the contain-ment heat removal systems, the containmentfisolation system, I

and the containment combustible gas control system.

The design basis accident (DBA) is defined as the most severe of a spectrum of hypothetical loss-of-coolant accidents (LOCA) .

l The ability of the containment systems to mitigate the con-sequences of a DBA depends upon the high reliability of these systems. This section provides the design criteria and evalua-l tions to demonstrate that these systems function within the i

t ,- specified limits throughout the unit operating lifetime.

j 6.2.1 CONTAINMENT FUNCTIONAL DESIGN I

A physical description of the containment and the design criteria relating to construction techniques, static loads, and seismic loads is provided in Section 3.8. This section pertains to those aspects of containment design, testing, and evaluation

( that relate to the accident-mitigation function.

6.2.1.1 Containment Structure

• l 6.2.1.1.1 Design Bases The safety design basis for .the containment is that the contain-l (Ns_ ment must withstand the pressures and temperatures of the DBA without exceeding the design leakage rate, as required by 10 CFR 50, Appendix A, General Design Criterion 50, and that, in conjunction with the other containment systems and the other engineered safety features, the release of radioactive material i subsequent to a DBA does not result in doses in excess of the l guideline values specified in 10 CFR 100. -The radiological l consequences of the DBA are presented in Section 15.6.

a. Assumed Accident Conditions I

For the purpose of determining the design pressure "f 'f -"~Q requirements for the containment structure and the containment internal structures, the following simul-taneous occurrences are assumed:

1. The postulated reactor coolant system pipe rupture, as listed in Table 6.2.1-1, is assumed to be concurrent with the loss of offsite power and the worst single active failure. Ne two pipe breaks are assumed to occur simultaneously ce consecutively.

For design loadings on the systems used to mitigate the consequences of a postulated reactor coolant, system pipe rupture, a safe shutdown earthquake is f"4 assumed.

Rev. OL-0 6.2.1-1 6/86

.. o CALLN/ LAY - SP

2. The postulated secondary system pipe rupture, as identified in Section 6.2.1.4, is assumed con-current with the loss of offsite power and the worst single active failure. No two pipe breaks are assumed to occur simultaneously or consecutively.

3. The postulated inadvertent operation of a contain- l ment heat removal system is considered a low I probability event and is not considered to be concurrent with any other event.

i The calculated maximum containment structure internal and  !

external pressures are listed in Table 6.2.1-2. These calcu- '

lated pressures are based on the conservative analyses described

,,j in Section 6.2.1.1.3 and demonstrate that a substantial margin exists (approximately 20 percent on internal pressure and 10 l percent on external pressure) between the calculated maximum pressure and the design pressures.

. The calculated maximum pressures on the containment internal r 1

C~ structures (e.g. primary and secondary shield walls) are listed i n in Table 6.2.1-2. These pressures are based on the conservative analyses described in Section 6:2.1.2. The loads on the internal structures are calculated using the differentials-between the maximum calculated subcompartment pressures and 14.7 psia, the pressure of the containment atmosphere at the time of peak subcompartment pressure.

b. Sources and Amounts of Mass and Energy Released The sources and amounts of mass and energy released for the postulated reactor coolant system pipe rup-tures and secondary system pipe ruptures are discussed in Sections 6.2.1.3 and 6.2.1.4, respectively.

c. Effects of the ESFs as Heat Removal Systems The effects of the ECCS as an energy removal system are discussed in the determination of the mass and

', 71 energy release data provided in Section 6.2.1.3. The l only additional effect of this system considered is the long-term heat removal capability of the residual heat removal heat exchangers. In performing the containment design evaluation in Section 6.2.1.1.3, j single failures of the ECCS are assumed to be con-sistent with the mass and energy release data assump-

.,e tions for the break analyzed. '

! The effects of the containment heat removal systems, I as active energy removal systems, have been considered in the containment design evaluation in Section p"

6.2.1.1.3. The most stringent single active failure l

l Rev. OL-0 l 6.2.1-2 6/86 l

l

.- o CALLAWAY - SP f.~ of these systems is assumed to be consictent with the

( mass and energy release data assumptions for each break analyzed. The total heat removed by each of the containment heat removal systems up to the time of the calculated peak containment pressure is listed in Table 6.2.1-8. The design bases of the containment heat removal systems are discussed in Section 6.2.2.

The functional performance of the containment and the ECCS also rely upon the operation of the containment isolation system, as described in Section 6.2.4.

Required isolation operations are assumed for purposes of the containment design evaluation in Section "N 6.2.1.1.3.

d. Parameters Affecting Capability for Post-Accident Pressure Reduction The principal parameters which affect post-accident pressure reduction are 1) the heat absorbed by the heat sinks inside the containment, 2) the heat removed by the containment air coolers, and 3) the heat transferred to the containment, sump by the containment spray system.

A conservative amount of heat sink material has been calculated, and its heat absorption capability has been considered in the containment design evaluation C- in Section 6.2.1.1.3. The parameters describing the heat sinks credited with heat absorption are provided in Table 6.2.1-4.

The pressure reduction capability of the containment air coolers and the containment spray systea consider .

the parameters provided in Table 6.2.1-3. The assumed start time of these active heat removal systems considers a diesel start time of 12 seconds, load sequencing times, and the maximum startup time of the l systems.

b

I supp rt of case c, large break LOCA (DECLG CD" * '

Maximum SI with 12-second diesel generator start) of

[J( ')

Table 15.6-10, an evaluation of the assumptions used in

} the LOCA and MSLB containment p.essurization calculations, with respect to the full functioning 3' times of the containment spray system and the containment air coolers, was performed. The evaluation

' shows that the containment pressurization calculations for both LOCA and MSLB provided sufficient margin so that a 12-second diesel generator start time does not

~

' A:c,", change the assumed full functioning times of the RTM y containment spray and the containment air coolers.

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! pressurization calculations are not required for case c

.of Table 15.6-10, since this case is bounded by the previously performed containment calculations.

e. Parameters Affecting Heat Removal from the Containment Heat is transferred f t ;m the containment to the outside environment Tia the fan coolers and residual heat removal heat exchangers through the component cooling water and essential service water systems and released to the ultimate heat sink. A small amount of I heat is also transferred through the containment wall and dome to the outside atmosphere.

'1 d The component cooling, water system is described in Section 9.2.2, the essential service water system is described in Section 9.2.1, and the ultimate heat sink is described in Section 9.2.5.

.lu Single failures in systems which remove energy from l' c, the containment are considered to be consistent with M the single failures assumed in the development of the mass and energy release data. The energy removal capability of the containment air coolers, the

~

containment spray system, and the residual heac removal system consider the parameters provided in Table 6.2.1-3. The long-term energy inventories and total heat transferred to the various containment heat removal mechanisms, as a function of time, are diagrammed in Figures 6.2.1-25 and 6.2.1-26 for the double-ended pump suction guillotine (DEPSG) break with minimum safety injection and DEPSG break with maximum safety injection cases, respectively.

I f. Bases for Containment Depressurization Rate i l

To meet the containment safety design basis of  !

limiting the release of radioactive material l

J subsequent to a DBA so that the doses are within the i

[. ,. g-y guideline values specified in 10 CFR 100, the t ' " p ? ". containment pressure is reduced to less than 50 percent of the containment design pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after on m 4 dent. Chapter 15.0 contains the assumptions- the analysis of the offsite j l

radiological consequences 6f the accident. l I

g. Bases for Minimum Containment Pressure Used in ECCS l Y'

1. Performance Studies

'[ The minimum containment pressure transient used in the

"# # analysis of the emergency core cooling system's

f. .. l .I capability is based on the conservative overestimated lMuns-y "; Rev. OL-0 HTph ' 6.2.1-4 6/86 T

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.- o l CALLAWAY - SP J heat removal capability and pressure reduction capa-k bility of the containment structures and the contain-

~

• ment systems and on the conservative reactor coolant l

system thermal analysis provi8ed in Section 15.6. The determination and evaluation of the minimum containment pressure transient are provided in Section 6.2.1.5.

6.2.1.1.2 Design Features The principal containment and containment subcompartment design parameters are provided in Table 6.2.1-2. General arrangement drawings for the reactor containment are provided in Figures

_s 1.2-9 through 1.2-18. Simplified arrangement drawings illustrat-ing the nodalization model used for the containment subcompart-(V ul ment analyses are provided in Figures 6.2.1-27 through 6.2.1-33, 6.2.1-43 through 6.2.1-55, and 6.2.1-76. j

a. Missile and Pipe Whip Protection Missile shield considerations are described in Section 3.5. The structural design of the containment and the containment subcompartments is discussed in Section .

I 3.8. The designed structhral strength considers the effects of pipe whip and jet forces, as discussed in Section 3.6.

b. Codes and Standards

(

The codes, standards, and guides applied in the design of the containment structure and the containment internal structures are identified in Section 3.8.

c. Inadvertent Operation of the containment Spray System The design external pressure doad on the reactor containment is provided in Table 6.2.1-2. The lowest calculated internal pressure is also provided in Table 6.2.1-2, and is the result of an assumed inadvertent

, pCy actuation of the containment spray system. The y jfel analysis performed to determine the lowest calculated internal pressure following an inadvertent actuation of the containment spray system is provided in Section 6.2.1.1.3. At least a 10-percent margin exists between the lowest calculated internal pressure and the design external pressure load.

d. Entrapment of Recirculation Water Locations within the reactor containment which may entrap spray water and subtract from the water in-

[

ventory considered to be available in the containment

• a4 sump are identified in Section'6.2 2.1. The effect "ofh 49 h "S Jo l

Rev. OL-0 6.2.1-5 6/86

-. . - . . - - - - - . - - - - ~ ~ . - - . - - - . - . _ . ~ .. . . - . . .- -

e. o CALLAWAY - SP this potential water loss is considered in determining l the net positive suction head available to the RHR and containment spray pumps. Any special provisions which reduce the amount of the entrapped water are discussed in Section 6.2.2.1.

e. Normal Operation of Systems Which control the Con-tainment Environment The functional capability and frequency of operation of the systems provided to maintain the containment

and subcompartment atmospheres within prescribed pressure, temperature, and humidity limits during i normal operation are discussed in Sections 6.2.2.2 and d 9.4.6.

6.2.1.1.3 Design Evaluation

a. Analysis of Postulated Ruptures

,~

'c -o uf In the event of a LOCA in the containment, much of the

^ 'EC released reactor coolant will flash to steam. This j release of mass and energy raises the temperature and i pressure of the atmosphere within the containment.

The severity of the temperature and pressure peaks depends upon the nature, size, and location of the postulated rupture.

In order to identify the worst case, the spectrum of  !

hypothetical accidents listed in Table 6.2.1-1 has been analyzed. The analytical model and computer code  ;

designed to predict containment pressure and tempera- i ture responses following the accidents are described in item b. of this section.

A summary of the resu. ts of the containment pressure

• and temperature analysis for this spectrum of postu- The l

lated accidents is provided in Table 6.2.1-8.

' ' lid

- peak containment pressure calculated resulted from the ng Mj assumed (DEPSG) break with minimum safety injection 1 F'^ and with the worst single failure being the loss of one emergency diesel.

The calculated containment pressure and temperature responses as a function of time for the spectrum of postulated accidents are illustrated in Figures (w?thpi 6.2.1-1 through 6.2.1-6 and 6.2.1-7 through 6.2.1-12, L*  ;

respectively. -

9 f., e-- a ,

.w V d3&$di irp MCan"'t k &"wA!)

"pdge% ' @

-$;,'7M [.q ,n b Rev. OL-0 L'

f 6.2.1-6 6/86 b'

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r e' e m CALLAWAY - SP Computer Codes for LOCA Analyses (s

'i b.

The spectrum of hypothetical accidents has been analyzed by the COPATTA computer cqde, which is designed to predict the pressure and temperature transients in the containment following a rupture. The mass and energy release data used by COPATTA are developed and described in Section 6.2.1.3. The  :

analytical model is described in the following references: 1 Refer to Sections 3.2 and 3.2.1 of Reference 1.

~';, THERMODYNAMIC ASSUMPTIONS - Refer to Section 3.2.2 of j Reference 1.

w ATMOSPHERE AND SUMP REGIONS - Refer to Section 3.2.3 of Reference 1.

REACTOR VESSEL REGION - Refer to Section 3.2.4 of Reference 1.

-i

' EMERGENCY CORE COOLING SYSTEMS,- Refer to Section 3.2.5 of Reference 1.

CONTAINMENT SPRAY SYSTEM - Refer to Section 3.2.6 of Reference 1.

/

HEAT TRANSFER SURFACES, HEAT SINK MODELING, AND HEAT

(~ TRANSFER COEFFICIENT - Refer to Section 3.2.7 of Reference 1.

ATMOSPHERE FAN COOLER UNITS - Refer to Section 3.2.8.1 of Reference 1.

Normally, the heat removal due to containment air cooler operation is simulated in the COPATTA code by specifying input values from a curve of heat removal rate versus containment atmosphere temperature. This curve is based upon the cooling coil thermal-physical design and is given in Figure 6.2.1-15. The fan

' coolers assumed start time is provided in Tables

{... , 6.2.1-6 and 6.2.1-7 for the DEPSG breaks with minimum safety injection and with maximum safety injection, respectively. This start time is based upon the diesel start time of 12 seconds, the loading sequence, l and the startup time of the system. The parameters describing the containment air coolers are given in Table 6.2.1-3.

A w a

( .

Rev. OL-9 6.2.1-7 5/97

e. (,

CALLRWAY - SP SUMP WATER RECIRCULATION HEAT EXCHANGERS - Refer to Section 3.2.8.2 of Reference 1.

c. Initial conditions Initial conditions used for the containment analysis are listed in Table 6.2.1-5.

The initial containment conditions were selected based on the range of the normal expected conditions within the containment with consideration given to maximizing the calculated peak. containment pressure. Parametric studies have been performed to uetermine the effects of varying these initial containment conditions (Ref.

1). The results of these studies showed that varying a the initial containment conditions over a wide range l

i of values changes the calculated peak pressure by less than 1 percent. Therefore, the initial containment conditions are relatively unimportant parameters with respect to the containment pressure temperature analysis.

us Q j *i, ' il '

The conservatisms in the assumptions made with respect to the containment heat removal systems'and the emergency core cooling system operability are discussed in Sections 6.2.2 and 6.3, respectively.

l d. Results of.the Failure Mode and Effects Analysis Single active failures have been considered in' the

! emergency core cooling system and in the containment heat removal nystems with respect to maximizing energy release to the containment and minimizing the heat removal frcn the containment. The criteria used to-determine the worst single active failure was the calculated peak containment pressure. Therefore, single active failures in the containment heat removal systems were considered consistent with the mass and energy release data determined by the corresponding

, common mode failure in the emergency core cooling

{

l

• Pp system.

Y^ The worst calculated peak containment pressure was the result of a double-ended pump suction guillotine break with. minimum performance of the emergency core cooling system and the containment heat removal systems.

e. Containment Design Parameters awsq The principal containment design parameters are provided in Table 6.2.1-2.

3

! b.,rY.. ,

9.

Rev. OL-0 6.2.1-8 6/86 l

l s.

CALLAWAY - SP l

s-

f. Engineered Safety Features Design '.'arameters i l

The engineered safety features design parameters used in the containment analysis are listed in Table 1 6.2.1-3. The parameters identified as full capacity

were used when no failure was assumed to affect the operation of that system, and the parameters identified of minimum capacity were used when a single failure was assumed to affect the operation of that system.

l The containment air cooler duty curve per air cooler used in the analysis is given in Figure 6.2.1-15.

j

g. Results of Postulated Accidents Analyzed l

h%a,/ A summary of the results of the containment pressure temperature analyses for the spectrum of postulated accidents is tabulated in Table 6.2.1-8.

h. Secondary System Pipe Rupture Containment Analysis A complete discussion of secondary system pipe rup-tures inside the containment with respect to the containment pressure and temperature responses is provided in Section 6.2.1.4.

1. Containment Passive Heat Sinks With respect to the modeling of the containment l

passive heat sinks for the heat transfer calculations used in the containment pressure temperature analysis, Reference 1, Section 3.2.7, provides the justification for 1) the computer mesh spacing used for concrete, steel, and steel-lined concrete heat sinks, 2) the steel-concrete interface resistance used for the steel-lined concrete heat-sinks, and 3) the heat transfer correlations used in the heat transfer calculations.

The specific passive heat sinks considered in the containment pressure temperature analysis and their parameters are listed in Table 6.2.1-4. Figures 6.2.1-13 and 6.2.1-14 show the condensing heat transfer coefficient as a function of time for the DEPSG with minimum safety' injection and DEPSG with maximum safety injection cases, respectively.

Zero heat transfer is specified at the outside surface of the containment exposed to the earth, and between the containment sump liquid and the containment atmosphere within the containment.

I (

I Rev. OL-O 6.2.1-9 6/86

CALLAWAY - SP .s ' ..

j. Analysis of Inadvertent Operation of a Containment Heat. Removal System ,

V Inadvertent actuation of the containment spray system results in the lowest calculated containment internal pressure.

As discussed in Soction 6.2.2.1, the containment spray l system can only ba~ actuated in two ways, either i automatically upon receipt of two-out-of-four contain-ment high pressure signals or manually from the control room.

w Section 7.3.8 discusses the engineered safety l features actuation system and demonstrates that the }~

system design precludes a single active or passive failure from inadvertently actuating the containment spray system. Manual actuation of the containment ,

spray system can only be accomplished by the reactor  ;

operator deliberately switching on.two switches on the main control board. The main control board is designed with physical separation of these switches to prevent accidental actuation of the spray system.

Thus, inadvertent actuation of the sprays is precluded by design, and only a deliberate actuation l of the containment spray system could result in the  !

i reactor building being sprayed. s Although precluded by design, inadvertent actuation of  ;

the containment spray system has been assumed, and the i resultant reduction in the containment pressure has .

been calculated. The postulated inadvertent actuation of the containment spray system is assumed, concurrent .

with the following conservative containment and environmental conditions:

Summer Winter I

Initial containment temperature, F 120 100

);

N .

Initial containment pressure, psia 14.7 14.7 Initial containment relative humidity, % 100 100 I Containment spray flow rate, gpm 3,900*  !

3,900*

(per train) ,

RWST water temperature,*F 60 37 I l ~

l

• Runout flow rates for the containment spray system.

l ss )

l l

J Rev. OL-4 6.2.1-10 6/90

, s, CALLAWAY - SP Actuation of the containment spray system could be postulated under any set of containment and environ-mental conditions. However, no consistent set of realistic conditions can categorically be selected as the worst case initial condition to be used in the containment pressure analysis. These assumed initial conditions are defined as limiting in that these conditions 1) represent the largest differences in the containment ambient temperature and the RWST tempera-ture and 2) the 100-percent humidity case maximizes the amount of mass transferred out of the containment atmosphere.

F Using Henry's law of partial pressures and the Ideal

")"5 .. Gas Law and assuming that the inadvertent operation of the containment spray system will reduce the contain-ment vapor temperature to coincide with that of the RWST water being sprayed, the maximum reduction in the containment pressure is provided in Table 6.2.1-2.

The containment design external pressure load is provided in Table 6.2.1-2, and shows a minimum of 10-percent margin above the maximum reduction in the containment pressure calculated by the above-described method. Thus, corrective action by the operator is

~ not required to ensure that containment integrity is maintained.

The control room operator will be notified that the containment spray system is operating through the following means:

1. The containment spray actuation annunciator light will be on, and an audible alert alarm will be sounded.

2. The running status light of the containment spray pumps will be on.

s

3. The open status lights of the containment spray system isolation valves will be on.

4. The containment normal sump and the incore instru-mentation tunnel level indicators and level alarms will be actuated.

5. The flow indicators for the discharge of the containment spray pumps will indicate flow in the containment spray pumps.

6. The balance-of-plant computer will audibly alert

, and visually inform the operator that the contain-ment spray system is actuated.

Rev. OL-0 6.2.1-11 6/86

CALLA';!AY - SP

k. Accident Chronology ~N

\

")

The chronology of events occurring after a DEPSG break with minimum safety injection is given in Table 6.2.1-6.- The chronology of events after a DEPSG break with maximum safety injection is given in Table 6.2.1-7.

1. Mass and Energy Balances A mass and energy balance for the reactor coolant system, steam generators, and the safety injection system is provided in Section 6.2.1.3.2 and shows the distribution of energy prior to the accident, at the ')

end of the blowdown phase, at the end of the core reflood phase, and at the end of the post-reflood phase.

A mass and energy balance for the reactor and con-tainment systems for the DEPSG break with minimum safety injection and DEPSG break with maximum safety injection are provided in Tables 6.2.1-9 and 6.2.1-10, respectively. These tables provide the distribution of energy at the following times:

1. Prior to the accident

)

2. Blowdown peak pressure

3. End of blowdown

4. Peak containment pressure

5. End of reflood i

6. Approximately one day after recirculation

m. Long-Term Cooling Following a LOCA The long-term system behavior during various LOCAs has  ;

l been evaluated to verify the ability of the ECCS and -

l the containment heat removal systems to keep the I reactor vessel flooded and maintain the containment below design conditions for all times following a This evaluation is based on the conservative l LOCA.

l predictions of the performance of these engineered safety features consistent with the single failures l

assumed for each accident analyzed. The heat genera-l tion rate from shutdown fissions, heavy isotope decay, and fission product decay is provided in Figure 6.2.1-16.

)

Rev. OL-0 6.2.1-12 6/86

. . - . - - _= . -. - - . _

, 'y CALLAWAY - SP The containment pressure and temperature transients

',; for the DEPSG break with minimum safety injection up to 10' seconds are shown in Figures 6.2.1-1 and 6.2.1-7, I respectively-. These figures demonstrate the containment systems' capability of rapidly reducing the containment pressure and temperature and maintaining those parameters to acceptably low values. The containment pressure and temperature transients for the DEPSG break with maximum safety injection up to 10' seconds are shown in Figures 6.2.1-2 and 6.2.1-8, respectively. For all other accidents analyzed, the pressure and temperature transients are provided for 108 seconds. These tran-sients demonstrate similar characteristics to the large DEPSG break transients discussed above and,

.u

~'

h) since the performance of the containment heat removal systems should be similar, long-term cooling is I ensured, l

The sump temperature transients for the DEPSG break with minimum safety injection and the DEPSG break with maximum safety injection are provided in Figures l

6.2.1-17 and 6.2.1-18, respectively.

l The energy removal rates for the containment fan coolers, the RHR heat exchangers, and the containment passive heat sinks for the DEPSG break with minimum

/ safety injection and the DEPSG break with maximum s '

safety injection as a function of time are shown in Figures 6.2.1-19 through 6.2.1-24.

l The containment system energy inventory as a function i of time is plotted for the DEPSG break with minimum safety injection and the DEPSG break with maximum safety injection in Figures 6.2.1-25 and 6.2.1-26, respectively. All mechanisms of energy removal from and transfer within the containment are addressed in l

these figures. Included are the vapor energy, sump energy, energy contained in heat sinks, total energy removed from the containment by fan coolers and by the g residual heat removal system, and net energy trans-I Q ferr'ed by sprays from the containment vapor to the sump.

For the DBA at the time of the calculated peak con-tainment pressure, the vapor energy is 286.76 x 10 8 Btu, the energy deposited in the sump is 92.31 x 10' Btu, the containment passive heat sinks have absorbed l 76.09 x 10' Btu, 4.27 x 10 8 Btu have been removed by the containment fan coolers, 5.60 x 10' Bcu have been transferred from the containment vapor to the sump via the containment sprays, and no energy has been removed by the RHR system. Safety injection is switched to i

Rev. OL-O 6.2.1-13 6/86

J r CALLAWAY - SP

'the recirculation mode at.1,509 seconds, and the ~ h.

containment sprays are switched to the recirculation mode at 3,227 seconds after the accident,

~

n. Accumulator Nitrogen Release

~ Table'6.2.1-11 provides the nitrogen release rate from the accumulators following the discharge of their liquid volumes. The added mass and associated energy of this nitrogen release are accounted for' in the LOCA analysis.

o. Normal Containment Ventilation System Evaluation The functional capability of the normal containment ventilation systems to maintain the temperature, pressure, and humidity in the containment and con-tainment subcompartments is discussed in Sections 6.2.2.2 and 9.4.6.

p. Post-Accident Monitoring Instrumentation for post-accident monitoring is discussed in Section 7.5.

6.2.1.2 Containment Subcompartments '}

6.2.1.2.1 Design Basis Subcompartments within the containment, principally the reactor cavity,_the steam generator loop compartments, and the pres-surizer compartment, are designed to withstand the transient differential pressures and jet impingement forces of a postu-lated pipe break. Venting of these chambers maintains the differential pressures within the structural limits. In addition, restraints on the reactor coolant pipes, reactor vessel,-steam generators, etc., are. designed so that neither pipe whip nor vessel upset forces threaten the integrity of the '

subcompartments or of the ;ontainment structure.

\'

Analysis of the pressure transients in the steam generator l compartment and pressurizer compartment has been performed to verify the adequacy of the structural design of these structures under accident conditions. The following is a synopsis of the pipe breaks analyzed:

.j

)

Rev. OL-2 6.2.1-14 6/88

>aan.'

j ". .

t L

CALLAWAY - SP l Pf/

l

a. For the steam generator loop compartments, the design basis break is a steam generator inlet elbow longitu-dinal split with a break flow area of 763 square inches, a double-ended steam generator outlet nozzle break restrained to a break flow area of 436 square inches, and a double-ended reactor coolant pump outlet nozzle break restrained to a break flow area of 236 square inches.

b. The pressurizer compartment is divided into two compartments: 1) the pressurizer vault and 2) the pressurizer surge line compartment.

l

• i The design basis break for these subcompartments is the double-ended pressurizer surge line break. In addition to this break, the pressurizer spray line break and the three break cases from the steam genera-tor loop compartment analysis.were considered in the selection of the. design analysis break. In all cases, l

the pressures in the pressurizer compartment were substantially lower than those resulting from the pressurizer surge line break.

l 6.2.1.2.2 Design Features l

All design features provided for alleviating pressure buildup within the subcompartments are discussed in the subcompartment design evaluation in Section 6.2.1.2.3. Reference 2 describes the design features which limit the movement of the pipe after the postulated break.

6.2.1.2.3 Design Evaluation j .a . Mass and Energy Release Rate Transient Model l

l The computer programs used to develop the mass and l energy release transients for subcompartment pressuri-zation analyses are described in Reference 3. Tables 6.2.1-12 through 6.2.1-16 provide tabulations of the mass and energy release rates versus time for the spectrum of breaks analyzed.

l b. Subcompartment Pressure Analyses Model The COPDA computer code (Ref. 4) employs a finite difference technique to solve the time dependent equations for the conservation of mass, energy, and momentum to perform the subcompartment analyses. This I

l.

Rev. OL-2 6.2.1-15 6/88

CALLAWAY - SP .' .-

~

code and the assumptions inherent to it are described /., .

fully in Reference 5.

. 1. Reactor Cavity Lipture Analysis -

On May 31, and October 26, 1984, Union Electric submitted Westinghouse topical reports (WCAP's

-10500, -10501, -10690 and 10691) to the NRC in order to demonstrate compliance with the revised GDC-4, which provides for the application of

" leak-before-break" technology to eliminate protective devices against dynamic loads resulting from postulated ruptures of primary coolant loops. By letter dated October 28, 1986, 'hl the NRC confirmed its finding that, based on the .ip'f a

UE submittals, Callaway is in compliance with the revised GDC-4. Based upon the revised GDC-4, the water bags were deleted from the design. A permanent reactor. cavity seal / neutron shield, as described in Section 3.8.3.1.4, has been installed. i L

s-l i

'h

~

i i

l m.

Rev. OL-5 6.2.1-16 6/91

---w -

- , , , - . - - +- - -- - -,r,- , , , -

CALLAWAY - SP

<a, 1 0

i M

) Sheet 6.2.1-17 has been deleted.

7>

s. >

i l

l

\ i Y l l

l l

1 l

l Rev. OL-2 6.2.1-17 6/88

,o s CALLAWAY - SP ,

. v.

2. Steam Generator Loop Compartments The steam generator loop compartment is subjected to double-ended breaks of the pump suction line, the cold leg, the hot leg, a longitudinal split of the hot leg, and double-ended branch line breaks.

'All double-ended breaks are mechanically restrained so that the largest breaks in the hot leg, cold leg, and pump suction are 763 in.2, 236 in.', and 436 in.', respectively. These three breaks envelope all postulated breaks within the steam generator loop compartment. These breaks were analyzed, using the same 59-node model, to determine the ')

maximum pressures on the walls of the compartment and on the enclosed equipment, i.e., the steam generator, the reactor coolant pump, and the pressurizer. The blowdown data for the three breaks are given in Tables 6.2.1-13 through 6.2.1-15. The nodalization model for the analyses is given in Figures 6.2.1-43 through 6.2.1-55.

Only breaks in loop 4 were analyzed, since this loop has the smallest vent area directly to the remainder of the containment due to the presence of the pressurizer, and thus results in the L highest pressures.

To ensure conservative design of the loop compart-ment walls and the equipment supports, the loads calculated for loop 4 were applied to the other three steam generator loop compartments by appro-priate translation and rotation of the force vector axes. The volumes of the subcompartments, as well as the initial conditions prior to the transient, are given in Table 6.2.1-22.

As with the reactor cavity analysis, the node boundaries were selected wherever significant

-restrictions to flow occurred. A sensitivity study was performed in which the number of nodes s in the steam generator compar~c ment was varied.

)

The resulting forces on the. compartment walls and on the equipment in all cases were less than the forces calculated with the 59-node model. There-fore, it was assumed that the nodalization employed in the original model was both adequate and l

l r

l N

l l- m.

Rev. OU-2 6.2.1-18 6/88

,r r 4 - -- -

. b CALLAWAY - SP I

q conservative. All major obstructions, such as

) columns, pumps, tanks, grating, and the steam generators, were considered in the calculation of the subcompartment volumes and vent areas. In addition, the values for volume were reduced by 5 percent to allow for minor obstructions, such as cable trays, supports, and various piping. The principal obstructions within the steam generator loop compartments were the reactor coolant pumps and the steam generatore. Flow through the reactor cavity was neglected. The flow coeffi-l cients associated with the flow paths were calcu-lated in the same manner as for the reactor

) cavity. The head loss coefficients used in the s

~'

.) calculation of the flow coefficients, as well as the vent areas and 1/a's for each flowpath, are listed in Table 6.2.1-23. i l

The fluid flow from one subcompartment to another was calculated, using the homogeneous frozen flow option in the analysis. The peak pressures for I each subcompartment are listed in Table 6.2.1-22.

The complete pressure histories for those subcom-partments near the break for each of the three break cases analyzed are shown in Figures 6.2.1-56, 6.2.1-57, 6.2.1-61, and 6.2.1-69. When the I

)/ subcompartment pressures were applied to their projected areas on the steam generator and the  ;

reactor coolant pump, the forces were determined on these pieces of equipment. The forces on the reactor coolant pump and the steam generator are shown in Figures 6.2.1-58, 6.2.1-59, 6.2.1-62 through 6.2.1-67, and 6.2.1-70 through 6.2.1-74.

The coeffici'ents used to calculate the forces are

! given in Tables 6.2.1-24 and 6.2.1-25. l The component and resultant forces on the steam  ;

generator and reactor coolant pump for the three i breaks analyzed are illustrated in Figures 6.2.1-60, j 6.2.1-68, and 6.2.1-75.  ;

Lxe '

( 3. Pressurizer Vault l The pressurizer vault is subjected to a pres- l surizer spray line break, a pressurizer surge line  !

break, and a reactor coolant loop break. The pressurizer surge line compartment located directly below the pressurizer vault is subject to a l pressurizer surge line break and reactor coolant pipe break within the steam generator compartment

~

adjacent to the pressurizer vault. Analyses showed that the worst postulated break for both l Rev. OL-0 6.2.1-19 6/86

J ..

CALLAWAY - SP the pressurizer vault and the surge line com- T <

partment was the double-ended pressurizer surge )

line break. The mass and energy release data for this case are given in Table 6.2.1-T6.

w In the model, the pressure is relieved through large vents in the top of the pressurizer vault, and through the surge line compartment, out into the steam generator loop compartment and then up to the remainder of the containment. Figure 6.2.1-76 provides a simplified elevation view of the pressurizer vault, and Figure 6.2.1-77.shows a schematic diagram of the flow model.

The subcompartment volumes along with the peak ..)

calculated pressures and the design pressures in the pressurizer vault and the surge line compart-ment are given in Table 6.2.1-26. The pressure histories of those subcompartments directly below the pressurizer are given in Figure 6.2.1-78.

Table 6.2.1-27 summarizes the head loss coeffi-cients used to calculate the flow coefficients and the vent areas and 1/a's for all of the flow paths.

c. Nodalization Model Adequacy i

The determination of nodalization models used for the SNUPPS subcompartment analysis is adequate and based on the following criteria:

i

a. The models are' physically representative of the geometry investigated.

b. Themodelsareofadequatedeta[1toconsider all significant obstructions and flow losses.

c. The selection of nodal boundaries and volumes reflect the conservative theoretical thermo and fluid dynamic application.

.. a A determination that these criteria are met is based on previously performed developmental SNUPPS subcompartment analysis, Bechtel experience in the performance of other PWR subcompartment analyses, and comparisons with information in the public domain (such as NUREG/CR-1199, and NUREG-0609).

i b*

)

Rev. OL-0 6.2.1-20 6/86

- - _- .- . . - - _ = _ - - - - - - . - - - - _ . - . - _ _ - .

% t. CALLAWAY - SP

, 6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents The containment system receives mass and energy releases  !

following a postulated rupture of the reactor coolant system ' '

l (RCS). These releases continue through blowdown and post-blowdown. The release rates are calculated for pipe failure at 1 three distinct locations: 1) hot leg, 2) pump suction, and 3) i cold legs. Because of the pressure in the RCS before the  !

postulated rupture, the mass and energy flows rapidly from the RCS to the containment. As the water exits from the rupture, a portion of it flashes into steam due to the pressure and temperature in the containment, compared to the pressure and d" temperature of the RCS. The blowdown reduces the pressure in y.; the RCS. .

1 During the reflood phase, these breaks have the following l different characteristics. For a cold leg pipe break, all of l the fluid which leaves the core must vent through a steam '

generator and becomes superheated. However, relative to breaks 1 at other locations, the core flooding rate (and, there fore, the i rate of fluid leaving the core) for cold leg breaks is low  :

because-all the core vent paths include the resistance of the {

reactor coolant pump. For a hot leg pipe break, the vent path j resistance is relatively low, which results in a high core 1 flooding rate, but the majority of the fluid which exits the I) core bypasses the steam generators in venting to the contain-

\f ment. The pump suction break combines the effects of the '

relatively high core flooding rate, as in the hot leg break, l and steam generator heat addition, as in the cold leg break. I As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period. The spectrum of breaks analyzed includes the largest cold and hot leg breaks, reactor inlet and outlet / respectively, and a range of pump suction breaks from the largest to a 3.0 ft 2 break.

Because of the phenomena of reflood, as discussed above, the pump suction break location is the limiting case, with the

. double-ended pump suction break being the most limiting. This

\** conclusion is supported by studies of smaller hot leg breaks which have been shown on similar plants to be less severe than the double-ended hot leg. Cold leg breaks, however, are lower both in the blowdown peak and in the reflood pressure rise.

Thus, an analysis of smaller pump suction breaks is representa-tive of the spectrum of break sizes.

The LOCA analysis calculational model is typically divided into three phases, which are: 1) blowdown, which includes the period from accident occurrence (when the reactor is at steady state full power operation) to the time when zero break flow is first calculated, 2) refill, which is from the end of blowdown to the time the emergency core cooling system (ECCS) fills the vessel l

'Rev. OL-0 6.2.1-21 6/86

CALLAWAY - SP lower plenum, and 3) reflood, which begins when water starts moving into the core and continues until the end of the tran- 0:s cient. For the pump suction break, consideration is given to a l possible fourth phase; that is, froth. boiling in the For steam a

I generator tubes after the core has been quenched. . l l

description of the calculational model used for the mass and l cnergy release analysis, see Reference 3.

6.2.1.3.1 Mass and Energy Release Data

a. Blowdown Mass and Energy Release Data Tables 6.2.1-28 through 6.2.1-32 present the calcu-lated mass and energy releases for the blowdown phase 3 of the various breaks analyzed with the corresponding /

break size.

b. Reflood Mass and Energy Release Data The lower vessel plenum is assumed to refill immediately following blowdown, hence the refill phase is skipped.

Tables 6.2.1-33 through 6.2.1-38 present the calculated mass and energy releases for the reflood phase of the l various breaks analyzed along with the corresponding

( safeguards assumption (maximum or minimum).

c. Dry Steam Post-Reflood Mass and Energy Release Data }

The calculated mass and energy releases for the post-reflood phase with dry steam are provided in the reflood mass and energy release tables (Tables 6.2.1-35 through 6.2.1-38) after the end of the 10-foot entrain-ment occurs. These tables correspond to the hot leg, cold leg, and small pump suction breaks analyzed,

d. Two-Phase Post-Reflood Mass and Energy Release Data Tables 6.2.1-39 and 6.2.1-40 present the two-phase (froth) mass and energy release data for a double-ended pump suction break, using minimum and maximum safeguards assumptions, respectively. The following x; ,')

procedure was followed to account for the depressuri-zation, equilibration, and decay heat mass and energy releases to 108 seconds.

1. Depressurization Energy Release The froth mass and energy release was initially tabulated based on a reference temperature for heat stored in the steam generator metal and secondary fluid of saturation at the containment design backpressure of 60.0 psig. Additional

)

J Rev. OL-0 6.2.1-22 6/86-

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{

1 i

two-phase mass and energy releases become avail-able due to the energy within the steam generators, as the containment depressurizes to atmospheric conditions.

2. Depressurization (Two-Phase Mixture)

I Tables 6.2.1-41 and 6.2.1-42 show the available I depressurization energy of the steam generators above atmospheric pressure (14.7 psia).

This energy is brought out in two stages. .In the first, the sources above are brought into equili-l ) brium with_the actual containment pressure. The (3}) rate for this phase is set by the froth calcula-tion models. In the second, the sources give up additional energy as the containment pressure decreases. The rate.for this stage is set by the l containment depressurization rate.

l l The depressurization mass and energy release rates can be determined if the depressurization time is

known. The depressurization time was estimated by l choosing a conservative.'.y low value that would

! maximize the depressurization mass and energy l

release rates to the containment (3,600 seconds l for normal dry containment). First, a containment

\/ pressure calculation was performed, neglecting the depressurization energy release. For this case, the containment will depressurize faster and, l hence, a conservative depressurization time is I calculated. The second containment pressure l calculation is made, utilizing depressurization time with the procedure for calculating depres-

! surization mass and energy release rates described in this section.

The steam generator depressurization mass and energy release rate from the broken and intact loops were calculated and added to the initial '

mass and energy releases, which were based on a )

containment back pressure of 60.0 psig, described l l below. I

3. Broken Loop Steam Generator - Equilibration Stage The amount of energy in the steam generator .8 directly proportional to pressure and, hence, the fraction to be brought out equals the difference between reference pressure (60.0 psig) and the

, actual containment pressure divided by reference pressure. Since the broken loop steam generator t

Rev. OL-0 6.2.1-23 '6/86 l

i

CALLAWAY - SP is in equilibrium with the reference pressure of 60.0 psig prior to the beginning of froth, a conservative value for steam generator heat release was assumed. A rate of 100,000 Btu /sec would release all of the available energy in 209 seconds. This value is conservative.

4. Broken Loop Steam Generator - Depressurization Stage The amount of energy to be brought out is the original amount of energy remaining in the broken loop steam generator given in Tables 6.2.1-41 and 6.2.1-42, less what is brought out to reach g equilibrium. The heat addition rate is this ,"*,c/

amount divided by the assumed depressurization time. The mass boiloff rate is this rate divided by latent heat. The energy addition rate is the boiloff rate times saturated vapor enthalpy.

5. 7.ntact Loop Steam Generator - Equilibration Stage The same procedure as for the broken loop is used here. However, metal and core energy is lumped with the steam generator energy for this calcula-tion. The fraction to be brought out to attain 'g equilibrium equals the difference between the )

reference containment pressure and the actual containment pressure divided by the reference value. The rate of addition to the containment is 90.0 lb/sec at 1,034 seconds. This cools the steam generator and metal at 37,576 Btu /sec.

Thus, the duration of the extension of the post-reflood table is the fraction times the available energy divided by the rate of cooling. This was not extended beyond recirculation because the continued condensation effect is-implicit in these numbers and should change after recirculation.

6. Intact Loop Steam Generator - Depressurization )

s .-

Stage Again the procedure used here is the same as the broken loop case except that the decay heat should be added to the heat addition rate, which was not included in the initial post-reflood tables. The amount of energy to be brought out is the original energy remaining given in Tables 6.2.1-41 and 6.2.1-42, less what is brought out to reach equilibrium. The heat addition rate is this amount divided by the depressurization time. The mass boiloff rate is this rate divided by latent

)

Rev. OL-0 6.2.1-24 6/86

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4 *. CALLAWAY - SP 1

l 39 heat. The energy addition rate is the boiloff. J l

.'g"["j rate times saturated vapor enthalpy. Beyond the equilibration stage, the mass boiloff rate due to decay heat is added to the depressurization mass ,

boiloff rates. This rate is the. decay heat rate '

divided by latent heat, including ECCS water i subcooling prior to recirculation. The corresponding energy addition rate is the boiloff  !

rate times saturated vapor enthalpy.

The continued condensation benefit is not implicit in these numbers, and thus this' calculation may 1

i - extend beyond recirculation, k

bi 7. Decay Heat Figure 6.2.1-16 presents the decay heat which is used for the depressurization calculation.

8. Post-Recirculation Energy Release ,

l Following initiation of the recirculation phase, the energy release rates are determined by the COPATTA computer code discussed in Section 6,.2.1.1.3 and Reference 1.

Recirculation for the maximum safety injection case occurs at 849 seconds, . which is during the j broken loop depressurization stage. During this i

stage, the energy release is a function of the l recirculated safety injection water sensible heat, l the reactor decay heat (Figure 6.2.1-16), the broken loop depressurization heat release, and the

l. actual containment pressure. The intact loop equilibration and depressurization releases are accounted for in the same way.

Recirculation for the minimum safety injection

(, case occurs at 1,509 seconds, which is during the

, intact loop depressurization stage. During this

! stage, the energy release is a function of the recirculated safety injection water sensible heat, the reactor decay heat, the broken and intact loop depressurization heat releases, and the actual containment pressure.

l--

Following the end of broken an.d intact loop depressurization to the end of'the transient, the energy release is a function of decay heat and sensible heat'only. E occurs at 3,772 and 3,nd 775ofseconds depressurization for the maximum and minimum safety injection cases, respectively, i

Rev. OL-0 6.2.1-25 6/86

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• t.

CALLAWAY - SP .

6.2.1.3.2 Energy Sources The sources of energy considered in the LOCA mass and energy ~y release analysis are given in the energy balance tables (Tables 6.2.1-43 through 6.2.1-48). These energy sources are:

a. RCS, accumulators, and pumped safety injection sensible heat

b. Decay heat

c. Core stored energy

d. Thick and thin metal energy

}

e. Steam generator energy The energy balance tables show the initial energy distribution and the energy distribution at end-of-blowdown (EOB), end-of-entrainment (EOE), end-of-froth (EOF), and end-of-froth intact loops (EOFIL) for the two-phase post-reflood analyses. For the dry steam post-reflood analyses, the energy distribution at an aosumed recirculation time of 1,500 seconds is given instead of EOF and EOFIL.

The methods and assumptions used to release the various energy s sources are given in Reference 3.

The following items ensure that the core energy release is conservatively analyzed for maximum containment pressure. j

a. Core power level of 3,636 MWt (102 percent of core power level)

b. Allowance in temperature for instrument error and dead band (+4 F) l

c. Margin in volume (1.4 percent)

d. Allowance in volume for thermal expansion (1.6 percent)

]

e. Margin in core power associated with use of engineered j safeguards design rating (ESDR)

f. Allowance for calorimetric error (2 percent of ESDR)

g. Conservatively modified coefficients of heat transfer

h. Allowance in core-stored energy for effect of fuel l

densification l 1. Margin in core-stored energy (+15 percent)

)

Rev.. OL-3 6.2.1-26 6/89

i

'S **

CALLAWAY - SP l .mq j .

Maximum calculated operating temperature (627.3 F) l //py with above assumption H6.2'1.3.3

. bescriptionofBlowdownModel ,

f l A description of the model used to determine the mass and

! energy released from the RCS during the blordown phase of a )

l postulated LOCA is provided in Reference'3. All significant l correlations.are discussed.

l l l 6.2.1.3.4 Description of Core Reflood Model 1 yp? A description of the model used to determine the mass and

{

L 4t , energy released from'the RCS during the reflood phase of a

'45 postulated LOCA is provided in Reference 3. All significant l correlations are discussed. Tran-tents of the principal l parameters during reflood are given in Tables 6.2.1-49 and j

6.2.1-50 for the limiting case pump suction breaks with maximum and minimum 1 safeguards.

J 6.2.1.3.5 Description of Long-Term Cooling Model l The calculational procedure used to determine the mass and j energy released during the post-reflood phase of a postulated j l LOCA is described in Reference 3. '

6.2.1.3.6 Single Failure Analysis The effect of single failures of various ECCS components on the mass and energy releases is included in these data. The two analyses for the DEPSG breaks bound this effect.

No single failure is as,sumed in determining the mass and energy l

releases for the maximum safeguards case. For the minimum safeguards case, the single failure assumed is the loss of one emergency diesel. This failure results in the loss of one pumped safety injection train. The analysis of both maximum

,,. and minimum safeguards cases ensures that the effect of all credible single failures is bounded.

/U e :

6.2.1.3.7 Metal-Water Reaction In the mass and energy release data presented here, no Zr-H 0 reaction heat was considered because the clad temperature did not rise high enough for the rate of the Zr-H 0 reaction to be of any significance.

6.2.1.3.8 Reactor Coolant System Mass and Energy Balance l Reactor coolant system mass and energy balances are tabulated for hot leg, cold leg,.and pump suction breaks in Tables

[*.. 6.2.1-43 through 6.2.1-48.

I L Rev. OL-0 1

6.2.1-27 6/86

, i.--. y,.-,.

l CALLAWAY - SP .* ra 6.2.1.3.9 Additional Information Required for Confirmatory j

Analysis

System parameters and hydraulic characteristics needed to l perform confirmatory analysis are provided in Tables 6.2.1-51 through 6.2.1-55. -

l 6.2.1.4 Mass and Energy Release Analysis fcr Postulated Secondary Pipe Ruptures Inside Containment Steam line ruptures occurring inside a reactor containment structure may result in significant releases of high energy fluid to the containment environment, possibly resulting in l high containment temperatures and pressures. The quantitative T

'I l

nature of the releases following a steam line rupture is j dependent upon the many possible configurations of the plant

! steam system and containment designs as well as the plant operating conditions and the size of the rupture. These variations make a reasonable determination of the single absolute " worst case" for both containment pressure and tem-perature evaluations following a steambreak difficult. This section describes the methods used in determining the contain-

! ment responses to a variety of postulated pipe breaks encom-

! passing wide variatione in plant operation, safety system performance, and break size.

Table 6.2.1-56 lists the 16 casee that were analyzed to h determine the worst case containment pressures and /

temperatures following a main steam line break. Out of these cases, the following four cases were reanalyzed to determine the impact of plant uprating to 3579 MWt:

' 1. Full double-ended rupture at 102 percent uprated power (Case 1). .

2

2. 0.60 ft double-ended rupture at 102 percent uprated j

power (Case 2).

2 i 3. 0.80 ft split rupture at 102 percent uprated power (Case 3). ]

~~

4. Full double-ended rupture at 102 percent uprated power L

.(Case 16) assuming failure of MSIV.

l The 8 percent revaporization of condensate as allowed by NUREG 0588 was modeled. The analysis was based on new mass and energy release data provided by Westinghouse. This mass and energy release data included the effects of superheated steam and no credit was taken for entrainment.

1 2

The analysis for the 0.8 ft split rupture was an iterative process. For the uprated condition, the time to reach Hi-1 set pressure of 6.0 psig and Hi-2 set pressure of 20 psig were

)

Rev. OL-3 6.2.1-28 6/89

' ~ '

j CALLAWAY - SP i

l found to be 16.6 seconds and 61.6 seconds, respectively. This

.f{;R igj f information was provided to Westinghouse to btain the total mass and energy release data for the 0.8 ft 2 split break case, Tables 6.2.1-57 and 6.2.1-58 have been revised to reflect the f

l effects of plant uprating, as given in Reference 10.

l 6.2.1.4.1 Significant Parameters Affecting Steam Line Break i Mass and Energy Releases l

There are four major factors that influence the release of mass

, and energy following a steam line break: steam generator fluid I

, inventory, primary to secondary heat transfer, protective l '

-3 system operation, and the state of the secondary fluid blow-l down. The following is a list of those plant variables which j determine the influence of each of these factors: '

l a. Plant power level

b. Main feedwater system design

c. Auxiliary feedwater system design

d. Postulated break type, size, and location

e. Availability of offsite power

f. Safety system failures r

! g. SG reverse heat transfer and reactor coolant system l

metal heat capacity The following ,is a discussion of each of these variables.

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3 , CALLAWAY - SP 6.2.1.4.1.1 Plant Power Level Steam line breaks can be postulated to occur with the plant in any operating condition ranging from hot standby to full power.

Since steam generator mass decreases with increasing power level, breaks occurring at lower power will generally result in a greater total mass release to the plant containment. However, because of increased energy storage in the primary plant, increased heat transfer in the steam generators, and the additional energy generation in the nuclear fuel, the energy release to the containment from breaks postulated to occur during power operation may be greater than for breaks occurring with the plant in a hot standby condition. Additionally, steam pressure and the dynamic conditions in the steam generators s change with increasing power and have significant influence on both the rate of blowdown and the amount of moisture entrained

.n the fluid leaving the break following a steambreak event.

Because of the opposing effects of changing power level on steam line break releases, no single power level can be singled out as a worst case initial condition for a steam line breuk event. Therefore, several different power levels spanning the operating range as well as the hot standby condition have been analyzed.

6.2.1.4.1.2 Main Feedwater System Design t

The rapid depressurization which occurs following a rupture may result in large amounts of water being added to the steam generators through the main feedwater system. Rapid closing isolation valves are provided in the main feedwater lines to limit this effect. Also, the piping layout downstream of the isolation valves affects the volume in the feedwater lines that cannot be isolated from the steam generators. As the steam generator pressure decreases, some of the fluid in this volume will flash into the steam generator, providing additional secondary fluid which may exit out the rupture.

The feedwater addition which occurs prior to closing of the

feedwater line isolation valves influences the steam generator blowdown in several ways. First, the rapid addition increases the amount of entrained water in large-break cases by lowering the bulk quality of the steam generator inventory. Secondly, because the water entering the steam generator is subcooled, it lowers the steam pressure, thereby reducing the flow rate out of the break. Finally, the increased flow rate causes an increase in the heat transfer rate from the primary to secondary system, resulting in greater energy being released out the break. Since thece are competing effects on the total mass and energy release, no " worst case" feedwater transient can be f

V Rev. OL-0 6.2.1-29 6/86 1

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CALLAWAY - SP defined for all plant conditions. In the results presented, the worst effects of each variable have been used. For example, moisture entrainment for each break is calculated, assuming conservatively small feedwater additions so that the entrained water is minimized. Determination of total steam generator inventory, however, is based on conservatively large feedwater additions, as explained in Section 6.2.1.4.3.2.

The unisolated feedwater line volumes between the steam genera-tor and the isolation valves serve as a source for additional high energy fluid to be discharged through the pipe break.

This volume is accounted for in the mass and energy release data presented in Section 6.2.1.4.3.2.

6.2.1.4.1.3 Auxiliary Feedwater System Design Within the first minute following a steam line break, the auxiliary feed system is initiated on any one of several protection system signals. Addition of auxiliary feedwater to the steam generators increases the secondary mass available for release to the containment, as well as increases the heat transferred to the secondary fluid. The effects on steam generator mass are maximized in the calculation described in Sect!on 6.2.1.4.3.2 by assuming full auxiliary feed flow to the faulted steam generator starting at time zero and continuing until manually stopped by the plant operator.

6.2.1.4.1.4 Postulated Break Type, Size, and Location

a. Postulated Break Type Two types of postulated pipe ruptures are considered in evaluating steam line breaks. ,

First is a split rupture in which a hole opens at some point on the side of the steam pipe or steam header but does not result in a complete severance of the pipe. A single, distinct break area is fed uniformly by all steam generators until steam line isolation occurs. The blowdown flow rates from the individual ',

steam generators are interdependent, since fluid coupling exists between all steam lines. Because flow limiting orifices are provided in each steam generator, the largest possible split rupture can have an effec-tive area prior to isolation that is no greater than the throat area of the flow restrictor times the number of plant primary coolant loops. Following isolation, the effective break area for the steam generator with the broken line can be no greater than the flow restrictor throat area.

s-

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Rev. OL-0 6.2.1-30 6/86

"., s, CALLAWAY - SP The second break type is the double-ended guillotine l

rupture.in which the steam pipe is completely severed and the ends of the break displace from each other.

Guillotine ruptures are characterized by two distinct break locations, each of equal area but being fed by different steam generators. The largest possible guillotine rupture can have an effective area per steam generator no greater than the throat area of one steamline flow restrictor.

The type of break influences the mass and energy releases to containment by altering both the nature of the steam blowdown from the piping in the steam plant and the effective break area fed by each steam genera-l (gq tor prior to steam line isolation. For example, a double-ended rupture in a pipe having a cross-sectional area of 2.4 square feet would appear as a 1.4-square-foot rupture to a single steam generator feeding one end of the break, but would appear as a 0.8-square-foot rupture to each of the steam generators feeding the other end of the break.

l b. Postulated size

, Break area is also important when evaluating steam

! line breaks. It controls the rate of releases to the l

l s

)f containment as well as exerts significant influence on the steam pressure decay and the amount of entrained water in the blowdown flow. The data presented in this section include releases for three break areas at l each of five initial power levels. Included are two double-ended and one split rupture, as follows:

l 1. A full double-ended pipe rupture downstream of the

! steam line flow restrictor. For this case, the

! actual break area equals the cross-sectional area

of the steam line, but the blowdown from the steam l generator with the broken line is controlled by l the flow restrictor throat area (1.4 square feet).

The reverse flow from the intact steam generators l

is controlled by the smaller of the pipe cross l section, the steam stop valve seat area, or the l total flow restrictor throat area in the intact l loops. The reverse flow has been conservatively l assumed to be controlled by the flow restrictors j in each of the intact loop steam generators.

l Actually, the combined flow from the three steam l generators must pass through an 18-inch (1.42

! square feet) line, which would greatly restrict l the flow.

~

I Rev. OL-0 l 6.2.1-31 6/86 L

l

e CALLAWAY - SP

2. A small double-ended rupture having an area just ~h larger than the area at which water entrainment '

~/

ceases. Entrainment is assumed in the forward direction only. Dry steam blowdown is assumed to occur.in the reverse direction.

3. A split break that represents the largest break which will neither generate a steam line isolation signal from the primary protection equipment nor result in moisture entrainment. Steam and feed-water line isolation signals will be generated by high containment pressure signals for these cases.

Being a split rupture, the effective area seen by the faulted steam generator will increase by a j factor of 4, following steam line isolation. c Conceivably, moisture entrainment could occur at that time. However, since steam line isolation for these breaks will generally not occur before 20-60 seconds, it is conservatively assumed that the pressure has decreased sufficiently in the affected steam generator to preclude any moisture carryover.

4. A break representing the largest double-ended rupture for which only dry steam blowdown occurs need not be presented. Studies (Ref. 7) have shown that this break size is typically smaller than the largest split break (no entrainment) for which blowdown for the split rupture will be more severe than the no-entrainment DER at any given power level.

c. Postulated Break Loc,ation Break location affects steam line blowdowns by virtue of the pressure losses which would occur in the length of piping between the steam generator and the break.

The effect of the pressure loss is to reduce the effective break area seen by the steam generator.

Although this would reduce the rate of blowdown, it i would not significantly change the total release of x>

energy to the containment. Therefore, piping loss effects have been conservatively ignored in all blowdown results, except in the small double-ended ruptures in which moisture entrainment occurs. The effects of pipe friction are conservatively assumed to be sufficiently large in this case to prevent moisture entrainment in the reverse flow, thus minimizing water relief to the containment.

)

Rev. OL-0 6.2.1-32 6/86 l

d '- CALLAWAT - SP C 6.2.1.4.1.5 Availability of Offsite Power The effects of the assumption of the availability of offsite power has been enveloped in the analysis. Loss of offsite power has been assumed where it delays the actuation of the containment heat removal systems (i.e., containment sprays and containment air coolers) due to the time required to start the emergency diesel generators. Offsite power has been assumed to be available where it maximizes the mass and energy released from the break due to 1) the continued operation of the reactor coolant pumps which maximizes the energy transferred from the reactor coolant system to the steam generators and 2) continued operation of the feedwater pumps and actuation of the auxiliary feedwater system which maximizes the steam generator inventories x, ' available for release.

6,2.1.4.1.6 Safety System Failures In addition single activetofailures assumingwerea considered:

loss of offsite power, the following

a. Loss of one emergency diesel

b. Failure of one main steam isolation valve

c. Failure of one main feedwater isolation valve

\ / The loss of one diesel results in the loss of one train of each of the containment heat removal systems. As discussed in Section 6.2.1.4.3.3, this is the most severe single active l failure.

The effect of a main steam isolation valve failure is to provide additional fluid which may be released to the con-tainment via the break. This results from the blowdown of all the steam piping between the break location and the isolation valves in the intact loops.

'N The failure of a main feedwater isolation valve will result in additional fluid being released to the containment following a main steam line break. The additional fluid to be released will be the volume between the isolation valve and the feed-water control valve.

6.2.1.4.1.7 Steam Generator Reverse Heat Transfer and Reactor Coolant System Metal Heat Capacity Once steam line isolation is complete, those steam generators in the intact steam loops become sources of energy which can be transferred to the steam generator with the broken line. This energy transfer occurs via the primary coolant. As the primary plant cools, the temperature of the coolant flowing in the 4

l Rev. OL-0

6.2.1-33 1 '6/86

b CALLAMAY - SP steam generator tubes drops below the temperature of the ~h secondary fluid in the intact units, resulting in energy being returned to the primary coolant. This energy is then available

'/

to be transferred to the steam generator with the broken steamline.

Similarly, the heat stored in the metal of the reactor coolant piping, the reactor vessel, and the reactor coolant pumps will be transferred to the primary coolant as the plant cooldown progresses. This energy also is available to be transferred to the steam generator with the broken line.

The effects of both the reactor coolant system metal and the reverse steam generator heat transfer are included in the 'T

/

results presented in this document.

6.2.1.4.2 Description of Blowdown Model A description of the blowdown model used is provided in Refer-ence 6. This reference is the basis for the tables contained in Reference 7.

6.2.1.4.3 Containment Response Analysis The COPATTA computer code (Ref. 1), which is discussed in Section 6.2.1.1.3, was used to determine the containment responses following the postulated main steam line breaks. The i following assumptions were made to obtain these responses.

6.2.1.4.3.1 Initial Conditions The initial containment conditions are the same as those used in the containment response analysis for the postulated reactor coolant system pipe ruptures"(see Table 6.2.1-5).

6.2.1.4.3.2 Mass and Energy Release Data The tables contained in Reference 7 present the mass and energy release data used to determine the containment pressure-temperature responses for the spectrum of breaks analyzed. The basis for )

these tables is provided in Reference 6, along with a discussion /

of the methods used to modify the data to reflect the specific plant design. The specific plant design input which was assumed is provided for each case in Table 6.2.1-57. Tables 6.2.1-57A and 6.2.1-57B provide the mass and energy release data for the cases which resulted in the highest temperature and pressure, respectively.

The rate of auxiliary feedwater addition represents the maximum runout flowrate to a fully depressurized steam generator. The value given for mass added by feedwater pumping assumes that no reduction in feedwater pump turbine speed occurs following a

)

Rev. OL-0 6.2.1-34 6/86

'. '. '. CALLAWAY - SP C- 11 MSLB and prior to main feedwater isolation. Feedwater isolation for the full and partial double-ended ruptures is g dependent on signals generated by the primary protection system, which results in isolation times ranging between 7.0 and 8.9 seconds for these cases. Feedwater isolation for the split breaks was based on the time required to reach the containment pressure setpoint which generates the isolation signal. Determination of feedwater flowrates prior to isolation assumed that the feedwater control valve in the broken loop goes wide open while those in the intact loops remain in their pre-break positions.

6.2.1.4.3.3 Containment Pressure-Temperature Results

. J. c Figures 6.2.1-79 through 6.2.1-82 provide curves of the resul-

\l: tant containment pressure-temperature transients for the cases producing the highest peak containment pressure and temperature.

Table 6.2.1-58 summarizes the results of all the cases analyzed and indicates the times at which dryout occurs and the various containment pressure setpoints are reached. The sequence of events following a postulated main steam line break is listed in Tables 6.2.1-59 and 6.2.1-60 for worst pressure and tempera-ture cases, respectively.

The worst single active failure is the loss of an emergency diesel. This is evident by comparing the results given in Table 6.2.1-58 for case 1, which assumes the loss of an emer-gency diesel, and case 16, which assumes the failure of a main steam isolation valve. Both cases assume a full double-ended rupture at 102-percent power. The failure of the main feed- l water isolation valve was not specifically analyzed. The  !

additional fluid which would be released is that contained in l the volume between the main feedwater isolation valve and the l feedwater control valve. This volume is only 50 cubic feet and l would not significantly affect the containment pressure- l temperature response. The loss of an emergency diesel has been assumed for the spectrum of break sizes and power levels analyzed, w As illustrated in Figure 6.2.1-79, case 12, 0.66 ft8 split at l

( 25-percent power, results in a peak pressure of 48.1 psig.

%~ This case represents the peak calculated containment pressure for the spectrum of breaks analyzed. The condensing heat transfer coefficient versus time for this case is provided in Figure 6.2.1-83.

It is important to note that the peak calculated pressure is coincident with the termination of the auxiliary feedwater flow to the affected steam generator, which was assumed to occur at 1,800 seconds (30 minutes). Actual termination of auxiliary feedwater flow to the affected steam generator due to operator

, action is expected to occur prior to 600 seconds (10 minutes),

b)I Rev. OL-4 6.2.1-35 6/90

- - . - . - - ~ . _ . - . - - . - ~ . _ . _ -

CALLAWAY - SP .! ' .'

as discussed in Section 10.4.9. The peak calculated pressure t for the spectrum of postulated breaks,' assuming 10-minute In all cases, the peak' 'A operator' action, would be 42.8 psig. =-

calculated containment pressure demonstrates considerable margin below the containment design pressure.

As illustrated in Figure'6.2.1-82, case 6, 0.84'ft* split at l 75-percent power, results in a peak vapor temperature of 384.9 F. This case represents the peak calculated containment vapor temperature for the spectrum of breaks analyzed. The condens-ing heat transfer coefficient versus time for this case is provided in Figure 6.2.1-84.

For the spectrum of breaks analyzed, the calculated containment- , ,

vapor. temperature for some cases exceeds the specified contain- ,}'

ment design temperature of 320 F for a short period of time, e The 320 F containment design temperature is the design tempera-ture for safety-related equipment and instrumentation located withir. the containment and not the maximum temperature allowed for the containment atmosphere vapor. l Figure 6.2.1-85 provides plots of surface temperature versus time for various representative materials within the contain-ment. These curves are based on the IEM model discussed in Reference 8, used in conjunction with COPATTA for the case i resulting in the highest material surface temperatures. These figures clearly show that the actual equipment temperatures, s following a postulated secondary' system break, are well below }

their design temperatures and are, in fact, approximated more closely by the containment' vapor saturation temperature.

l Cables _ocated inside the containment are qualified to higher

! ' temperatures (340 to 385 F) than their surfaces are expected to experience as shown in Figure 7A of the NUREG-0588 submittal.

The' calculated temperature for each type of cable is below the qualification temperature; however, due to the low mass to surface area ratios for cables, the calculated jacket / cable surface temperatures exceed the containment vapor saturation temperature. .

6.2.1.4.3.4 Energy Inventories

\,

Mass and energy balances are provided in Tables 6.2.1-61 and l

6.2.1-62 for the most severe secondary pipe ruptures, based on

[ the highest peak calculated containment pressure and tempera-L ture, respectively.

6.2.1.4.~4 Results of Postulated Feedwater Line Breaks Inside Containment The effects of a postulated feedwater line break on the con-l tainment is not as severe as the MSLB because the initial break

/

i

,J Rev. OL-4 L 6.2.1-36 6/90 l

a, CALLAWAY - SP offluent during a fondwater lina break is at a lower specific

( enthalpy.

6.2.1.4.5 Additional Information Required for Confirmatory Analysis No additional information is deemed necessary for the performance of confirmatory analyses.

6.2.1.5 Minimum Containment Pressure Analysis f_Qr Performance Canability Studies on Emercency Core Cooling System s

The containment backpressure used for the limiting case j (CD - 0.6) double-ended cold leg guillotine break for the ECCS analysis presented in Section 15.6.5 is presented in Figure I

6.2.1-86. The containment backpressure is calculated, using the methods and assumptions described in Appendix A of Reference 9. Input parameters, including the containment  ;

initial conditions, net free containment volume, passive heat sink materials, thicknesses, and surface areas, and starting time and number of centainment cooling systems used in the 3 analysis, are described in the following paragraphs.

s. J Mass and Energy Release Data 6.2.1.5.1 The mass and energy releases to the containment during the '

blowdown and reflood portions of the limiting break transient are presented in Tables 6.2.1-63 and 6.2.1-64.

The mathematical models shich calculate the mass and energy releases to the containment are described in Section 15.6.5 and conform to 10 CFR Part 50, Appendix K, "ECCS Evaluation Models." A break spectrum analysis is performed (see references in Section 15.6.5) that considers various break sizes, break locations, and Moody discharge coefficients for the double-ended cold leg guillotines which do not affect the

'. mass and energy released to the containment. This effect is considered for each case analyzed. During refill, the mass and  ;

energy released to the containment is assumed to be zero, which l minimizes the containment pressure. During reflood, the i effect of steam water mixing between the safety injection water and the steam flowing through the reactor coolant system intact loops reduces the available energy released to the co'.tainment vapor spaces and therefore tends to minimize containment pressure.

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CALLAWAY - SP ,

'6.2.1.5.2 -Initial Containm:nt Internal Conditions The following initial values were used in the analysis: _

a .- A containment pressure of 14.7 psia. .

b. A containment temperature of 90 F. P

c. A refueling water storage tank temperature - of 37 F.

d. An outside temperature of -30 F. l

e. A relative humidity of 99 percent.

These containment initial conditions are representatively low i values anticipated during normal full power operation.

Containment Volume 6.2.1.5.3 The volume used in the analysis was 2.7 x 106 fg3 ,

6.2.1.5.4 Active Heat Sinks L

.The containment spray system and containment air coolers operate to remove heat from the containment.

Pertinent data for these systems which were used in the }

j analysis are presented in Table 6.2.1-65.

The sump temperature was not used in the analysis because the maximum peak cladding temperature occurs prior to initiation of the recirculation phase for the containment spray system. In l

addition, heat transfer between the sump water and the containment vapor space was not considered in the analysis.

6.2.1.5.5 Steam-Water Mixing l

l l' Water spillage rates from the broken loop accumulator are l-determined as part of the core reflooding calculation and are included in the containment code (COCO) calculational model. .

6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis, with their thermophysical properties, are given in Table 6.2.1-66. The passive heat sinks and thermophysical properties were derived in compliance with Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation.*

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6.2.1.5.7. Hnct Transfer to.Pacaive Hnnt Sinks A.

M~ ~

The condensing. heat transfer-coefficients used for heat

. transfer to the steel containment structures are given in Figure 6.2.1-87 for the limiting break. The containment pressure transient for.the limiting break is shown in Figure

6.2.1-86.

6.2.1.6- Tests and Inspections Refer to Sections 6.2.6 and 6.6 6.2.1.7 Instrumentation Reauirements-u

) Instrumentation.is provided to actuate the engineered safety features and to monitor the containment temperature, pressure, and sump level. Design details and logic of the' instrumentation are discussed in Sections 7.1, 7.2, 7.3, and

7.5.

6.2.

1.8 REFERENCES

~

1. Bechtel Power Corporation, " Performance and Sizing of Dry Pressure Containments," Topical Report No. BN-TOP-3, (Rev.

4), October 1977.

2. " Pipe Breaks for the LOCA Analysis of the Westinghouse V/

} Primary Coolant Loop," WCAP-8082-P-A (Proprietary) and WCAP-8172-A (Non-Proprietary), January 1975.

3. Shepard, R.M., et al, " Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8264-P-A, Rev.

1, (Proprietary) and WCAP-8312-A, Rev. 2

-(Non-Proprietary), August 1975.

4. Bechtel Power Corporation "COPDA, Compartment Pressure Design Analysis," (Bechtel Computer Code), 1973.

5. Bechtel Power Corporation, "Subcompartment Pressure and Temperature Transient Analysis," Topical Report No.

.r. BN-TOP-4, (Rev. 1), October 1977.

.i

6. Land, R. E., " Mass and Energy Releases Following a Steam Line. Rupture," WCAP-8822 (Proprietary) and WCAP-8860 (Non-Proprietary) , September 1976.

7. Letter - SNP-2035 (P) Proprietary, Westinghouse (Rawlins)

-to Bechtel (Tudera), September 12, 1978.

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8. Letter - Dockst 50-368, " Main...Steamlins Break Accident 7g Environmental Qualifications," John F. Stolz (NRC) to William Cavanaugh III (Arkansas Power and Light Co.), i-)

April,14, 1978. _

9.' Bordelon,'F. M., Massie, H. W.,

Jr., Zordon, T. A., .

" Westinghouse Emergency Core Cooling System Evaluation Model Summary", WCAP-8339, June 1974.

10. ULNRC-1471 "Callaway Plant Uprating Submittal",

March 31, 1987.

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Rev. OL-3 6.2.1-40 6/89

CALLAWAY - SP 6.2.2 CONTAINMENT EEAT REMOVAL SYSTEMS The functional performance objective of the containment heat removal system, as an engineered safety features system, is to reduce the containment temperature and pressure following a LOCA or rain steam line break (MSLB) accident, by removing thermal energy from the contairment atmosphere. These cooling systems also serve to limit of f site radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby diminishing the driving force for the leakage of fission products from the containment to the environment. The containment heat removal systems include the residual heat removal system discussed in Sections 5.4.7, 6.2.1, and 6.3, the containment spray system

,j (CSS) discussed in Section 6.2.2.1, and the containment 13 cooling system discussed in Section 6.2.2.2.

6.2.2.1 Containment Sprav System 6.2.2.1.1 Design Bases 6.2.2.1.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The CSS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, or external missiles (GDC-2).

SAFETY DESIGN BASIS TWO - The CSS is designed to remain functional after a SSE or to perform its intended function following the postulated hazard of a pipe break (GDC-3 and 4) .

SAFETY DESIGN BASIS THREE - Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC-38).

SAFETY DESIGN BASIS FOUR - The active components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection of components at appropriate times specified in the ASME Boiler and Pressure Vessel Code,Section XI (GDC-39 and 40).

g. i SAFETY DESIGN BASIS FIVE - The CSS is designed and' fabricated sof to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide

1. 3 2 ..

SAFETY DESIGN BASIS SIX - The capability of isolating components or piping is provided so that the CSS safety function will not be compromised. This includes isolation of components to deal with leakage or malfunctions (GDC-38).

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CALLAWAY - SP ,e f.'

SAFETY DESIGN BASIS SEVEN - Th@ containment isolation valves in the system are selected, tested, and located in accordance with the requirements of GDC-54 and 56 and 10 CFR 50, Appendix f J, Type A testing.

SAFETY DESIGN BASIS EIGHT - The CSS, in conjunction with the containment fan cooler system and the emergency core cooling system, is designed to be capable of removing sufficient heat and subsequent decay heat from the containment atmosphere following the hypothesized LOCA or MSLB to maintain the containment pressure below the containment design pressure.

Section 6.2.1 provides the assumptions as to sources and amounts of energy considered and the analysis of the containment pressure transient following a LOCA or MSLB j accident inside the containment (GDC-38).

SAFETY DESIGN BASIS NINE - The CSS remains operable in the I accident environment.

SAFETY DESIGN BASIS TEN - The containment spray water does not contain substances which would be unstable in the thermal or radiolytic environment of the LOCA or cause extensive corrosive attack on equipment.

SAFETY DESIGN BASIS ELEVEN - The CSS is designed so that adequate net positive suction head (NPSH) ex3ctc at the suction of the containment spray pumps during all operating phases, in accordance with Regulatory Guide 1.1.

SAFETY DESIGN BASIS TWELVE - The CSS is designed to prevent debris which could impair the-performance of the containment spray pumps, valves, eductors, or spray nozzles from entering the recirculation piping. Design is in accordance with Regulatory Guide 1.82, as discussed in Table 6.2.2-1.

6.2.2.1.1.2 Power Generation Design Bases The CSS has no power. generation design bases.

6.2.2.1.2 System Design 6.2.2.1.2.1 General Description -q The CSS, shown schematically in Figure 6.2.2-1, consists of two separate trains of equal capacity, each independently capable of meeting the design bases. Each train includes a containment spray pump, spray header and nozzles, spray recirculation path, valves, and the necessary piping, instrumentation, flushing connections, and controls.

)

l Rev. OL-8 6.2.2-2 11/95

, CALLAWAY - SP

. ,l '

l The rsfueling water etorage tank cupplies borated injection .

l f water to the containment spray system. Each train takes suction from separate containment recirculation sumps during l

the~ recirculation phase.

l l The CSS provides a spray of cold'or subcooled borated water from the upper regions of the containment to reduce the l containment pressure and temperature during either a LOCA or MSLB inside the containment.

Each CSS pump discharges into the containment atmosphere through an independent spray header. The spray headers are located in the upper part of the reactor building to allow maximum time for the falling spray droplets to reach thermal equilibrium with the steam-air atmosphere. The condensation of the steam by the falling spray results in a reduction in containment pressure and temperature. Each spray train provides adequate coverage to meet the design requirements with respect to both containment heat removal and iodine removal. Further discussion of the iodine removal function of l the CSS is provided in Section 6.5.2.

In the CSS, only the containment recirculation sumps, the trisodium phosphate baskets, the spray headers, nozzles, and i associated piping and valves are located within the containment. The remainder of the system is located within the auxiliary building, separated from that portion in the containment by motor-operated isolation valves. During the

~5) recirculation phase, leakage outside of the containment will be detected with the auxiliary building radiation indicators and alarms, temperature alarms, and auxiliary building sump alarms. The motor-operated isolation valves in each train assure train isolation capability in the event of leakage during the recirculation phase. Leakage detection within the auxiliary building is discussed in Section 9.3.3.

Following a large break LOCA, the containment spray during the injection phase will be a boric acid solution having a pH of about 4.5. The desired pH level is greater than 7.0 to assure iodine retention in the sumps, to limit corrosion and the associated production of hydrogen, and to limit chloride

, induced stress-corrosion cracking of austenitic stainless steels. To adjust the sump solution pH into the desired range, a minimum of 9000 pounds of trisodium phosphate dodecahydrate

( Na PO,

• 12 H O e 1/4 NaOH ) is stored in two baskets, one within 3 2 the confines of each containment recirculation sump, which will be submerged after a LOCA. This amount of trisodium phosphate is sufficient to assure that the equilibrium sump solution pH will be greater than or equal to 7.1.

l The baskets are stainless steel with mesh sides and bottoms to permit a large surface to be exposed to the solution, thus maximizing the rate of dissolution into the sump. During the

\"f recirculation phase, the fluid mass released to the containment is screened through a 1/8-inch wire' mesh before Rev. OL-8 6.2.2-3 11/95

_ - . . ~ . . .. . . . _ . _ _ . _ _ . . . _ _ . . . _ _ . _ _ _ _ . _ . . _ _ . _ _ . _

CALLAWAY - SP j' ,.

cntering.the racirculation cumps to ba pumpad back through-the spray nozzles. . Trisodium phosphate'(TSP-C), . stored in baskets -

within the confines of the recirculation sumps at an elevation that will be flooded post-LOCA, dissolves in the sump solution >]b thereby raising the' sump solution pH' to enhvice materials compatibility and retention of. iodine in tie sump fluid.

6.2.2.1.2.2 Component Description Mechanical components of the CSS are described in this section. Component design parameters are given in Table 6.2.2-2.

Each component in the CSS.is designed and manufactured to withstand the environmental effects, including radiation, )

found in Table 3.11(B)-2.

CONTAINMENT SPRAY PUMPS - The two CS pumps are the vertical centrifugal type, driven by electric induction motors. The  ;

motors have open drip-proof enclosures and are provided with adequate insulation which will allow continuous operation of r

,m.

s

\ .. ,)

l Rev. OL-8 6.2.2-3a 11/95 t

,, CALLAWAY - SP J .

c 100-perc nt-rated load at 50 C ambient. Power for theso motors is supplied from the Class IE 4,160-Volt busses.

Power supply availability is discussed in Section 8.3.

The pump motors are specified to have the capability of starting and accelerating the driven equipment, under load, to a design point running speed within 4 seconds, based on 75 percent of the rated motor voltage. The pumps are designed to withstand a thermal transient from 37'F to 300*F occurring in 10 seconds, which exceeds the severity of the transient occurring when pump suction is switched from the RWST to the containment sump.

The shaft seals on the pumps are reliable, easy to maintain.

! and compatible with the fluids to be circulated. They are designed to operate at a temperature of 300 F, which exceeds the maximum temperature to which they will be exposed following an accident.

The containment spray pumps are designed to handle the runout flow associated with the startup transient, when minimal discharge head is applied.

CONTAINMENT SPRAY HEADER AND NOZZLES - Each containment spray header contains 197 hollow cone nozzles, each capable of the design flow and differential pressure given in Table 6.2.2-2.

These nozzles have a 7/16-inch spray orifice. The nozzles produce a drop size distribution, as described in Figure 6.5-2, at system design conditions. Special tests performed on the spray nozzles are discussed in Section 6.5.2.2.2. The spray solution is completely stable and soluble at all temperatures of interest in the containment and, therefore, will not precipitate or otherwise interfere with nozzle performance. The nozzles of each header are oriented to provide greater than 90-p'ercent area coverage at the operating deck of the reactor building. The area coverage at the operating deck (based on the calculated post-LOCA-containment saturation temperature) is provided in Table 6.5-2 for various nozzle orientations. The containment spray envelope reduction factor as a function of post-LOCA containment saturation

~% temperature is provided in Figure 6.5-4. The spray header

(,) design, nozzle spacing, and crientation are shown in Figure 6.2.2-2. The containment spray header and nozzles are designed to withstand the impulse of a water hammer at the commencement of flow.

CONTAINMENT RECIRCULATION SUMPS - The two containment recirculation sumps are collecting reservoirs from which the containment spray pumps and the residual heat removal pumps separ6;ely take suction after the contents of the refueling water storage tank have been expended. The sumps are located as far as feasible from the reactor coolant system piping and

, components which could become sources of debris. Thermal

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insulation uced incido containment will not be a significant

,. source of debris. The majority of insulation is NUKON which is discussed in Reference 2. Limited quantities of other types of insulation are used in widely dispersed locations. A

• design basis accident will not degrade a sufficient quantity of this insulation to adversely affect the performance of the sump. A baffle arrangement consisting of grating, coarse screening, and fine screening completely surrounds the sumps to prevent floating debris and high-density particles from entering. The grating is placed as the outermost barrier and is used to keep large debris off the screens. The next barrier is coarse industrial wire mesh screening which has a 1/2-inch maximum opening. The inner barrier is fine industrial wire mesh screening which has a 1/8-inch maximum

) opening. The attachment of the screening to the support structure is designed to keep debris from bypassing the screening. The sump baffle arrangement is shown in Figure 6.2.2-3.

Sources of debris, as indicated above, are physically remote from the recirculation sumps. Debris generated as a result of a LOCA will either be retained in an area such as the reactor cavity or refueling pool or must follow a tortuous path to reach the recirculation sump screens. Therefore, no appreciable debris will reach the recirculation sump screens to cause any significant blockage. In addition, as demonstrated in Figure 6.2.2-3, the recirculation sumps are T covered with concrete pads supporting the accumulator tanks; J thus, debris cannot fall directly upon the screening structure. However, the screens have been sized per Regulatory Guide 1.82, assuming 50-percent blockage, as discussed in Section 6.2.2.1.3, Safety Evaluation Twelve. To limit any possible vortexing, vortex breakers are placed horizontally over each containment recirculation sump at the 2000-foot level and in the suction lines from containment sumps to the containment spray pumps. The suction pipe from the sump is horizontal to limit any possible vortexing and has sufficient submergence to ensure continuous intake flow. The suction lines from the containment sumps to the containment spray pumps are sloped to assure switchover capability. These lines, up to and including the isolation valve, are encased in

_) guard piping.

%)

REFUELING WATER STORAGE TANK - The refueling water storage tank (RWST) is an austenitic stainless steel tank containing borated water at a concentration of 2,350-2,500 ppm boron.

The design parameters are given in Table 6.2.2-2.

The tank is an atmospheric storage tank vented directly to the atmosphere. Thermal insulation and heating are provided to prevent the tank contents from freezing. A manway is provided for tank internal inspection. Tank level indication and high and low level alarms are also provided. Additional

%)

) information is provided in Section 6.3.

Rev. OL-4 6.2.2-5 6/90

,{l

~CALLAWAY - SP  ; ,

2<>. VALVES'- CSS motor-oparated valves' are capable of being

• )

l operated from the control room. All valve seats are capable

~

of' limiting through leakage to less'than 2 cubic centimeters ~]/ l Gate and globe per hour per. nominal inch of pipe diameter. l valves:are provided with backseats. i l

i Encapsulation - The. containment spray system suction lines from the containment recirculation sumps are each provided ,

with a single remote manual' gate isolation valve outside the  ;

. containment. The piping from the sump up to and including the valve and.its motor operator is. enclosed in an encapsulation arrangement which is leaktight at the containment design pressure. A seal is provided so that the encapsulation is not .:

connected directly to the containment sump ~or containment atmosphere. A single passive or active failure in the sump }

f lines or in the encapsulation arrangement will not provide a ,

path for leakage to the environment.

~

PIPING - The piping of each spray header contains a test i connection. ' Air can be introduced into this connection to '

verify spray nozzle flow. Check valves immediately upstream of each spray ring header prevent system contamination due to pressurization in the containment and provide containment isolation backup protection.

A containment spray pump test l b e seen the pumps'

' discharges and the RWST is insteilt or periodic testing.

6.2.'2.1.2.3 ' System Operation )

The CSS'has two phases of operation, which are initiate'd sequentially following system actuation; they are the ,

injection phase and the recirculation phase. I INJECTION PEASE - The CSS,is actuated either manually from the l control room or on the coincidence of two-out-of-four

. containment Hi-3 pressure signals.

Both containment spr'ay pumps start and the motor-operated spray ring header isolation valves open to begin the injection phase. A summary of the accident chronology for the ,

, containment spray system is provided in Table 6.2.2-3 for the j injection' phase of a LOCA and MSLB inside the containment. /

The containment spray pump inlet nozzle, located at El. 1,970,  ;

takes suction from the RWST, located at El. 2,000'-6", through -

locked open valves. More than 95 percent of the pump discharge is directed to the containment spray ring headers.

'These headers are located at elevations up to 2,201 feet, the highest practical level to maximize iodine removal (discussed l

l

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t Rev. OL-8 6.2.2-6 11/95

' CALLAWAY - SP 1 ..

in Ssction 6.5.2). Tha headora are located outside of and i above the internal containment structures which serve as '

4 /

) missile barriers and are thereby protected from missiles generated during a LOCA or MSLB. The remaining portion of the l

l containment spray pump discharge is recirculated.

l l On coincidence of two-out-of-four low level signals from the RWST level transmitters, the emergency core cooling system i (ECCS) pumps switch suction to the containment recirculation sump, as described in Section 6.3.2. The low-low-1 level i

setpoint indicates that 104,080 useable gallons remain in the l RWST. Switchover for the spray pumps is manually initiated I when the low-low-2 level in the RWST is reached. The l low-low-2 level indicates imminent depletion of the RWST.

Switchover initiated at the time of the low-low-2 level alarm ensures that the system piping remains full of water and that adequate NPSH for the spray pumps is maintained. The RWST low-low-2 level alarms and level indicators inform the operator of the need to make this switchover.

The time length of the containment spray injection phase is given in Table 6.2.2-4. These times are based on the minimum RWST volume and are given for credible combinations of minimum and maximum containment spray and ECCS operation and runout l flow rates of these pumps. l RECIRCULATION PHASE - The recirculation phase is initiated by l

,f T the operator manually shifting containment spray pump suction from the RWST to the containment recirculation sump. The accident chronology for the containment spray system for the recirculation phase of a LOCA is provided in Table 6.2.2-3.

The RWST suction line valves remain open during the switchover to the recirculation phase to preclude the loss of supply to the containment spray pumps in the highly unlikely event that the isolation valve in the recirculation line is delayed in opening. The operator then remote manually closes the motor-operated valves in the RWST suction lines l N

U d

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CALLAWAY - SP .'.

l The suction line from the containment recirculation sump to the spray pump is a sloped line which precludes air from entering the system. The single valve in the containment sump recirculation line for the containment spray pump is encapsulated and located outside the containment. The flow paths from the spray pumps are the same as in the injection phase. Check valves are provided in the recirculation sump I suction lines to prevent the establishment of a flow path between the RWST and the containment sump.

Containment spray in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculation sump water. The length of time that the CSS operates during the recirculation phase is determined by the )

operator. The spray cannot be terminated until completion of the injection phase.

6.2.2.1.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design basis.

SAFETY EVALUATION ONE - The safety-related portions of the CSS are located in the reactor and auxiliary buildings. These buildings are designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, Sections 3.3, 3.4, and other appropriate natural phenoment.

3.5, 3.7 (B) , and 3.8 provide the basis for the adequacy of the structural design of these buildings.

SAFETY EVALUATION TWO - The safety-related portions of the CSS are designed to remain functional after a SSE. Sections 3.7 (B) .2 and 3.9 (B) provide the design loading conditions that were considered. Section.3.6 provides the hazards analysis to assure that the system performs its intended function.

SAFETY EVALUATION THREE - There are two spray system trains with complete redundancy of active components. Each train is capable of providing full design flow and cooling. In the event of the failure of a pump, valve, actuation system, or any other

. component in one train, the other train would be unaffected. To -;

assure that a single failure will neither initiate a spurious s containment spray nor prevent the activation of a necessary component, the containment spray pumps and containment header valves are actuated by the independent containment spray actuation signal (CSAS). The refueling water storage tank l (RWST) is Rev. OL-8 6.2.2-8 11/95

,,- CALLAWAY - SP common to the two trains and in used only during the injection

'~. phase following a LOCA. Redundant level indication for this tank is provided. No power-operated valve is installed in the common suction header from the RWST so that it is impossible for an active failure to disable both trains during the injection phase. Single failure analysis for the CSS is given in Table 6.2.2-5. l The emergency power supply pump room cooling and control and instrumentation systems serving one train are independent of comparable supporting systems for the other train. All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0. Minimum availability of the CSS is discussed in the Callaway Technical Specifications.

SAFETY EVALUATION FOUR - The CSS was initially tested with the l program given in Chapter 14.0, Functional testing is done in accordance with Section 6.2.2.1.4.

Section 6.6 provides the ASME Boiler and Pressure Vessel Code,Section XI requirements that are appropriate for the CSS.

SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-related portion of this system and supporting systems.

Section 6.2.2.1.2.2 shows that safety-related components meet the design and fabrication codes given in Section 3.2. All the power supplies and the control functions necessary for the safe function of the CSS are Class 1E, as described in Chapters 7.0 and 8.0.

SAFETY EVALUATION SIX - Section 6.2.2.1.2.1 describes provisions made to identify and isolate leakage or malfunction and to isolate the nonsafety-related portions of the system.

SAFETY EVALUATION SEVEN - Sections 6.2.4 and 6.2.6 provide the safety evaluation for the system containment isolation arrangement and testability.

SAFETY EVALUATION EIGHT - As shown by the containment analysis s and the description of the analytical methods,and models given

, ) in Section 6.2.1, the containment spray system, in conjunction s,. / with the emergency core cooling system and the containment fan coolers, is capable of removing sufficient heat energy and subsequent decay heat from the containment atmosphere following the hypothesized LOCA and MSLB inside the containment to maintain the containment pressure below the design pressure. Curves showing sump temperature, heat generation rates, heat removal rates of the containment heat removal systems, and containment total pressure, vapor pressure, and temperature as a function of time for minimum engineered safety features performance are also given in Section 6.2.1.

J, Rev. OL-8 6.2.2-9 11/95

CALLAWAY - SP

.' , f During the injection phase, all pressure transient analyses take credit for a spray system capable of delivering borated .

100 F spray water at the design flow rate. For the design basis LOCA and MSLB accident, credit is taken for spray flow initiation within 60 seconds.*

l An assured water volume of 394,000 gallons is available in the RWST to ensure that, after a LOCA, sufficient water is injected for emergency core cooling and forInrapidly reducing addition, this the containment pressure and temperature.

volume ensures that sufficient water is available in the containment sump to permit recirculation flow to the core and  !

the containment and to meet the NPSH requirements of the l residual heat removal and containment spray pumps and assurre i that a sufficient water volume is available in the RWST to allow for manual switchover of the containment spray pumps.

For the recirculation phase, while the safety injection system pumps are still operating after a LOCA, containment pressure transient analysis in Section 6.2.1 assumes residual heat removal by heat exchangers, as described in Section 5.4.7.

Credit is taken for heat removal from heat exchangers during the recirculation phase based on a tube side inlet temperature equal to the recirculation sump temperature, which is given in Section 6.2.1 as a function of time after the accident.

Each spray header train provides a minimum of 90-percent area coverage at the operating deck, as demonstrated in Figure 6.2.2-4. Area coverage by these spray nozzles varies as a function of saturation temperature. The design basis coverage for the nozzles at various orientations is provided in Table 6.5-2 and is based on the calculated containment saturation temperature. Figure 6.5-4 provides the curve of the containment spray envelope reduction factor to determine the design basis coverage. The minimum of 90-percent area coverage.at the operating dock is used as a layout guide for the location of the spray nozzles on the containment spray headers to assure 100-percent volumetric coverage above the operating floor of the containment. Physical obstructions, such as the containment polar crane, are not considered to

~

impede the spray coverage due to the extreme turbulence created by the hydrogen mixing fans, containment air coolers, j the spray within the containment, and the blowdown resulting from the postulated rupture. Thus, the header layout coupled with the extreme turbulence assures the validity of a

~

one-region model above the operating deck for accident dose calculations (see Chapter 15.0).

Discussion of the volume of containment covered by the sprays is provided in Section 6.5.2.

• LOCA case-4 spray flow initiation within 70 seconds.

Rev. OL-4 6.2.2-10 6/90

CALLAWAY - SP SAFETY EVALUATION NINE - That part of the CSS located inside the containment is designed to remain operable in the containment accident environment described in Section 3.11(B) .

The material compatibility of the containment spray system in contact with the post-accident recirculation fluids is discussed in Section 6.1. That part of the CSS located in the l auxiliary building is designed to remain operable in the auxiliary building accident environment described in Section i 3.11(B). l I

SAFETY EVALUATION TEN - The borated spray solution is stable l under the anticipated LOCA thermal and radiolytic conditions.

The borated solution is chemically compatible with components I with which it may come into contact. The use of materials

) which react to release hydrogen (principally n. gnesium, zinc, I and aluminum) has been minimized in equipment located inside the containment. An analysis of hydrogen generation following a LOCA is given in Section 6.2.5.

h SAFETY EVALUATION ELEVEN - System piping size and layout will provide adequate NPSH to the containment spray pump during all I anticipated operating conditions, in accordance with '

Regulatory Guide 1.1. In calculating available NPSH, the conservative assumption has been made that the water in the containment sump after a design basis LOCA is a saturated liquid, and no credit has been taken for anticipated subcooling. That is, although NPSH = elevation head + l (containment pressure - liquid vapor pressure) - suction line losses, the (containment pressure - liquid vapor pressure) term has been assumed to be zero. Calculated NPSH exceeds required NPSH by at least 10 percent. The recirculation piping penetrating the containment sumps is nearly horizontal to minimize vortexing. In addition, a vortex breaker is provided in the inlet of the piping from the sump.

In calculating the water level within the reactor building t which contributes to the NPSH available to the containment I spray pumps at the beginning of its recirculation phase, l consideration has been given to the potential mechanisms of water loss within the reactor building. These water loss l mechanisms include water present in the vapor phase, water

,) loss to compartments below El. 2,000, water loss above El.

s/ c 2,000, and water loss due to wetted surfaces. Tables 6.2.2-6 and 6.2.2-6a identify each water source which releases water to the reactor building and its associated mass and each potential water loss mechanism and the volume of water not assumed to contribute to the water level within the containment for a large LOCA and a MSLB, respectively. The static head available to contribute to the NPSH of the pump, suction line losses, and the minimum NPSH available are also given in Table I 6.2.2-7. The CSS pump NPSH versus flow is shown in Rev. OL-8 6.2.2-11 11/95

CALLAWAY - SP j' ,-

Figure 6.2.2-5. The reduction in water level dua to potential water loss mechanisms is considered in the calculated NPSH available.

SAFETY EVALUATION TWELVE - Recirculation sump construction provides screening down to 1/8-inch mesh to prevent entrained particles in excess of that size from entering the containment recirculation sump and containment spray system suction piping. Restrictions in the reactor core channels are the minimum restrictions and, therefore, the basis of the mesh opening size.

Since the containment spray pumps are designed to operate with entrained particles up to 1/4 inch in diameter and the minimum constriction size in the spray nozzles is 7/16 inch, this screening is adequate to assure proper system op"erability.

Each screening barrier has supports which are designed to withstand the differential pressure which would exist if the screens were 100 percent clogged. Both the screens and the ,

grating are designed to withstand the differential pressure of 100-percent clogging and sufficient screen area exists to allow over 50-percent clogging both screens without degrading spray pump NPSH. The sump baffle arrangement is shown in Figure 6.2.2-3.

The sump baffle arrangement does not allow flow into the sump below 6 inches above the concrete floor level surrounding the sump. This arrangement leaves ample depth for buildup of )

high-density debris without affecting sump performance.

Additionally, the velocity of recirculated fluids approaching the trash rack will be between 0.01 and 0.08 fps for all modes of operation following a LOCA or MSLB, and thus a low velocity I

settling region for high-density particles is provided. Table t 6.2.2-9 provides flow velocities at several times and l locations for a large LOCA and an MSLB.

Any debris which eludes the baffling, screens, and settling region passes into the sump through the 1/8-inch screen and

, will be drawn into the suction piping for the containment

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I spray.and residual heat removal systems. Such debris is small ,

enough to pass through any restriction in either system or the \

reactor vessel channels, and will eventually be pumped back l

into the containment.

l A comparison of the containment recirculation sump design features with each of the positions of Regulatory Guide 1.82,

" Sump for Emergency Core Cooling and Containment Spray Systems," is provided in Table 6.2.2-1.

l Rev. OL-0

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CALLAWAY - SP i , .

6.2.2.1.4 Tc:ts and Insp ctions l

Testing and inspection of components of the CSS are discussed I in this section.

Each containment spray pump has a shop test to generate complete performance curves. The test includes verifying total developed head (TDH), efficiency, and brake horsepower '

for various flow rates. An NPSH test for various flow rates 1 was performed on one pump. A shop thermal transient analysis, i from ambient temperature to 350 F in 10 seconds, has been performed on the CSS pump. Results of that analysis assure l that the design is suitable for the switchover from the  ;

injection to the recirculation phase.

The screening configuration on the containment recirculation sumps is shop tested to verify that all design requirements are adequately met.

The spray nozzles' design parameters were verified with prototype tests in the vendor's shop. Results of those test are provided in Section 6.5.2.2.2.

PREOPERATIONAL TESTING - Instruments are calibrated prior to system preoperational testing. Alarm functions are checked for operability and limits during preoperational testing. The flow paths and flow capacities of all components are verified d'uring preoperational tests.

The functional test of the ECCS, described in Section 6.3, demonstrates proper transfer to the emergency diesel generator power source in the event of a loss of power. A test signal simulating the containment spray signal is used to demonstrate the operation of the spray system up to the isolation valves on the pump discharge. The isolation valves are closed for the test. These isolation valves are functionally tested i separately. l l

The spray header nozzle performance is verified during the preoperational testing by blowing air through the nozzles and l observing the movement of the te11 tales.

)

The objectives of preoperational testing are to:

a. Demonstrate that the system is adequate to meet the design pressure and temperature conditions.

Components are tested in conformance with applicable codes.

b. Deme"* rate that the spray nozzles in the containment spray header are clear of obstructions by passing air through them, utilizing test connections.

Rev. OL-8 6.2.2-13 11/95 l

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CALLAWAY - SP .

c. Verify that the proper atquencing of valven and pumps

! occurs on initiation of the CSS and demonstrate the .

_ proper operation of remotely operated valves. .

d. Verify the operation of the spray pumps. Each spray pump is operated at full flow to verify that it meets.

the design curve generated during shop testing. Both design point and runout flow rates are uti21 zed to l verify that the pump performance is within design.

In addition, each spray pump is operated at minimum 1 '

flow, which is directed back to the refueling water l storage tank. A flow orifice is provided to regulate l

minimum flow to that required for routine testing.

! Verification of vortex control and acceptable pressure drops 't e '

a through the sump screens has been accomplished by hydraulic -

i .model testing (discussed previously in Appendix 3A, 1

Conformance with Regulatory Guide 1.79). The tests replicated the spray pumps and RHR pumps taking suction at full flow rates from the sumps for a variety of approach flow conditions, screen blockages, water levels, and pump operation

combinations. Data from these tests together with known 4 -

pressure drops across suction lines and valves (determined i using standard engineering calculations) verified that the

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available net positive suction head is adequate.

I Further details of each preoperational test to be performed are discussed in Chapter 14.0. s

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OPERATIONAL TESTING - The CSS is designed to permit periodic determination of proper system operability,.as specified in

.the Callaway Technical Specifications. The objectives of

. operational testing are to:

a. Verify that the proper sequencing of valves and pumps occurs on initiation of the containment spray signal and demonstrate the proper operation of remotely operated valves.

b. Verify the operation of the spray pumps. That each pump is run at a minimum flow and the flow is directed back to the RWST. p, ;

To assure the structural and leaktight integrity of components, the operability and performance of the active components, and

the operability of the system as a whole, the system is periodically tested up to the last isolation valve before the ,

containment penetration. The testing is accomplished by using a recirculation line (sized to take :UD percent of the design flow) back to the RWST.

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Rev. OL-8 6.2.2-14 11/95

.,.. ., CALLAWAY - SP ls '

1 All in3trumentation will also bn pariodically checked and calibrated. The CSS actuation is verified as follows:

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a. A containment spray actuation' signal (CSAS) subchannel-is actuated during normal operation to  ;

start the containment spray pump. .l l

b. A separate CSAS slave relay is actuated during normal I reactor operation to ensure the opening of the i containment header valves. The CSS pump will not be i operating.- l 6.2.2.1.5 Instrumentation Requirements i

}

_,e The CSS instrumentation vas designed to facilitate automatic operation, remote control,-and continuous indication of system I

parameters.

The containment has redundant analog level channels-for sump  !

recirculation with indication and alarms in the control room.

These circuits will aid the operator in determining the

presence and rate of increase of the sump water level.

All system motor-operated valves have position indication L provided in, and are operable from, the control room. This l l allows the operator to continuously monitor system status and remotely operate valves, as necessary. Details of the design j y ..and logic of the instrumentation are discussed in Chapter 7.0.

6.2.2.1.6 Nhterials The CSS is constructed primarily of corrosion-resistant austenitic' stainless steel and contains none of the restricted materials, discussed in Section 6.1.1.1.2.

-l l Construction materials'for components in the CSS are provided

! in Table 6.2.2-2.

Further discussion of the materials associated with the CSS, l s including containment spray fluid , chemistry, is given in j Section 6.5.2.6.

6.2.2.2 Containment Coolina System I'

The containment cooling system (CtCS), in conjunction with the containment HVAC systems described in Section 9.4.6, functions Rev. OL-8 6.2.2-15 11/95

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during normal plant operation to malatain a suitable abmosphere for equipment located within the containment.

Subsequent to a DBA within the containment, the containment cooling system provides a means of cooling the containment

(

...)

atmosphere to reduce pressure and thus reduce the potential for containment leakage of airborne and gasecun radioactivity to the environment.

6.2.2.2.1 Design Bases 6.2.2.2.1.1 Safety Design Bases The CtCS, excluding the system ductwork downstream of the cooler discharge plenum, is safety related and required to function following a DBA to achieve and maintain the plant in )

a safe shutdown condition.

SAFETY DESIGN BASIS ONE - The CtCS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, or external missiles (GDC-2) .

SAFETY DESIGN BASIS TWO - The CtCS is designed to remain functional after a safe shutdown earthsacke or to perform its intended function following a postu'u* 3 nazard, such as a fire, internal missile, or pipe br.nt (G9C-3 and 4) .

SAFETY DESIGN BASIS THREE - Safety fLnctions can be performed, assuming a single active component failure coincident with the ,

loss of offsite power (GDC-38).

SAFETY DESIGN BASIS FOUR - Active components are capable of being tested during plant operation. Provisions are made to allow for inservice inspection of components at appropriate times specified in the ASME Boiler and Pressure Vessel Code,Section XI (GDC-39 and 40-) .

SAFETY DESIGN BASIS FIVE - The CtCS is designed and fabricated to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide 1.32. p .;

SAFETY DESIGN BASIS SIX - The capability of isolating components, systems, or piping is provided, if required, so that the system's safety function will not be compromised.

This includes the bypassing of the nonsafety-related ductwork portions of the system.

SAFETY DESIGN BASIS SEVEN - The CtCS, in conjunction with the CSS, is capable of removing sufficient heat energy and subsequent decay heat from the containment atmosphere following the LOCA or MSLB accident to maintain the containment pressure

)

Rev. OL-0 6.2.2-16 6/86

g, , CALLAWAY - SP aq below design values. Section 6.2.1, Containment Functional

) Design, provides the assumptions as to sources and amounts of energy considered and the analyses of the containment pressure transient following a LOCA or an MSLB accident inside the containment. Actual containment fan cooler system parameters are such that those used in the analyses are equal to or more conservative than the actual containment fan cooler system capability.

SAFETY DESIGN BASIS EIGHT - The containment coolers, including the fan / motor combination, will remain operable in the acci-dent environment.

6.2.2.2.1.2 Power Generation Design Bases POWER GENERATION DESIGN BASIS ONE - The containment cooling system, operating in conjunction with the containment heating, ventilating, and air-conditioning system described in Section 9.4.6, is designed to limit the ambient containment air tempera-ture during normal plant operation to 120 F with any three of the four containment coolers operating. During normal plant operations, the hydrogen mixing fans are designed to provide sufficient air flow through the steam generator compartments so that a suitable environment for the equipment in the steam generator compartment can be maintained.

6.2.2.2.2 System Description 6.2.2.2.2.1 General Description The containment cooling system provides cooling by recir-culation of the containment air across air-to-water heat exchangers. The bulk of this cooled air is supplied to the lower regions of the steam generator compartments. The re-maining air is supplied to the instrument tunnel and at each level.(operating floor and below) of the containment outside the secondary shield wall. The air supplied to each steam t generator compartment is drawn upwards through the compart-ments by the hydrogen mixing fans and discharged into the

upper elevations of the containment.

w 6.2.2.2.2.2 Component Description Design parameters for the major components of the containment cooling system are provided in Table 6.2.2-2.

CONTAINMENT COOLER FAN - The containment cooler fans are located vertically in the bottom of the cooler housing. Fans are vaneaxial fans with two-speed motors. The fans and motors are designed for high-speed operation during normal plant

_ operations and for low-speed operation under post-LOCA condi-fJ, t

tions.

Rev. OL-0 6.2.2-17 6/86

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CALLAWAY - SP CONTAINMENT COOLER HOUSING / DISCHARGE PLENUM - The containment ~3g cooler housing and discharge plenums are constructed of struc- -J tural steel framework and galvanized steel coverings.

The containment cooler housing, including the section of ductwork containing the fusible link plates, is designed to sustain a differential pressure of 2 psi during pressure transients associated with accident conditions. An analysis which was performed to establish the differential pressure across the cooler housing indicates the maximum differential to be less than 0.1 psi (2.8 in. w.g.) under accident con-ditions. Ductwork was not considered in the analysis since it is designed to separate from the cooler by action of the fus- '

ible link plates. The fusible link plates are steel plates '

i which are hinged to the ductwork and held in a closed position by the fusible links (typical detail is shown in Figure 6.2.2-6).

The plates will employ a release mechanism so that after fusio.n of the links the plates will release from the ductwork.

The fusible links will be designed to release at a temperature of aoproximately 160 F. The open area vacated by the plates exceeds the cross-sectional area of the fan, thus providing an unrestricted flow path.

6.2.2.2.2.3 System Operation NORMAL OPERATION - Normally, each of the four containment coolers are operating to provide containment cooling capa- ]

bilities. Although only three coolers are required to provide the proper cooling (approximately 9.5 x 10 6 Btu /hr), four coolers are operated to maintain proper air flow distribution.

The fans are normally operating at the higher speed and the cooling water flow to the coils on low (normal) flow. The coil heat removal capabilities were designed, assuming a tube fouling factor of 0.002.

Condensate from the fan cooler coils is collected and measured to detect leaks into the containment atmosphere, as discussed in Section 5.2.5.

PLANT SHUTDOWN / REFUELING - The containment coolers may be J i

operated during shutdown / refueling operations to provide s-supplemental air distribution within the containment. The containment cooler fans may be operated at low speed to reduce noise levels within the containment during this mode of opera-tion. The coolers may be operated with the service water to provide supplemental cooling or without service water for supplemental heating by utilizing the motor heat load.

CONTAINMENT INTEGRATED LEAK RATE TESTING - The containment coolers are operated during containment integrated leak rate testing (ILRT) to maintain uniform containment temperature.

• The coolers are operated with service water to provide cooling I

Rev. OL-0 6.2.2-18 6/86 e . -

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. CALLAWAY - SP l

8- and without service water to provide heating, by utilizing the J motor heat load, during the test procedure. The fans are operated at low speeds during this elevated pressure condition to prevent motor overload.

1 POSTACCIDENT OPERATION - Following an SIS, the fans are designed l to start automatically in slow speed if not already running.

If running in high (normal) speed, the fans automatically shift to slow speed. Assuming loss of offsite power, the containment cooler fans are started 45 seconds after genera- l tion of the SIS.

1

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To compensate for the reduced air flow over the coils and to maximize heat removal, the cooling water flow through the

.;)

'~

cooling coils for each unit is automatically increased from 1,100 gpm to 2,000 gpm upon receipt of a SIS. The fusible link plates open to allow unrestricted flow through the air coolers. Each containment cooler is capable of removing at least 100 x 108 Btu /hr under design post-LOCA conditions.

The coil heat removal capabilities were designed, assuming a ,

tube fouling factor of 0.002.

)

l The fan can be operated from the control room at any time, but I cannot be manually operated at high speed if a containment I high pressure signal is in effect in order to prevent motor overload.

s The postaccident air-distribution system is designed to dis-charge the air from each unit through the opening left by the fusible link plate. The fusible link plates are steel plates which are hinged to the ductwork and held in a closed position by the fusible links. The plates will employ a release mecha-nism, using counterbalance weights to ensure that after fusion of the links the plates will release from the ductwork without l the aid of the fan head and against the pressure differential l established during the pressure transient. The fusible links '

will be designed to release at a temperature of approximately 160 F. The open area vacated by the plates approximately I

() equals the cross-sectional area of the fan, thus providing an unrestricted flow path.

Under design conditions, it is assumed that the existing ductwork is restricted so that all the air is discharged through this opening. Under these conditions, the throw is approximately 100 feet. Thus, the discharge from the units is well beyond their intake regions, preventing any short cir-cuiting. The air streams drop off toward the end of the throw and tend to settle toward the bottom of the containment due to the slightly lower temperatures and the air flow patterns established by the operation of the hydrogen mixing fans.

These expected air flow patterns are shown in Figure 6.2.2-7.

/ The volume of air recirculated in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by the combined air

(,_ flows of one train of the containment coolers and one train of Rev. OL-0 6.2.2-19 6/86

CALLAWAY - SP the hydrogen mixing fans will be approximately four times the 'T containment free volume. These air flow patterns and recir- 'f

'I culation volumes provide adequate circulation and, therefore, sufficient postaccident mixing of the containment atmosphere.

6.2.2.2.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases in Section 6.2.2.2.1.

SAFETY EVALUATION ONE - The safety-related portions of the containment cooling system are located in the reactor building.

This building is designed to withstand the effects of earth-quakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena. Sections 3.3, 3.4, 3.5, n,

3.7(B), and 3.8 provide the bases for the adequacy of the "

structural design of these buildings.

SAFETY EVALUATION TWO - The safety-related portions of the containment cooling system are designed to remain functional after a SSE. Sections 3.7(B).2 and 3.9(B) provide the design loading conditions that were considered. Section 3.5, 3.6, and 9.5.1 provide the hazards analyses to assure that a safe shutdown, as outlined in Section 7.4, can be achieved and maintained.

SAFETY EVALUATION THREE - The system description for the containment cooling system shows that complete redundancy is )

provided and, as indicated by Table 6.2.2-8, no single failure will compromise the system's safety functions. All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0.

SAFETY EVALUATION FOUR - The containment cooling system is initially tested with the program given in Chapter 14.0.

Periodic inservice functional testing is done in accordance with Section 6.2.2.2.4.

Section 6.6 provides the ASME Boiler and Pressure Vessel Code, Section X1 requirements that are appropriate for the contain-ment cooling system. \- )

SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to the safety-re]eted portion of this system and supporting system.

All the power supplies and control functions necessary for safe function of the containment cooling system are Class IE, as described in Chapters 7.0 and 8.0.

SAFETY EVALUATION SIX - Section 6.2.2.2.2.3 describes pro-visions made to allow the bypassing of the nonsafety-related ductwork portions of the system.

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Rev. OL-0 l 6.2.2-20 6/86

! o; CALLAWAY - SP SAFETY EVALUATION SEVEN - As shown by the containment analysis and the description of the analytical methods and models given in Section 6.2.1, the containment cooling system, in conjunc-l tion with the containment spray system, is capable of removing sufficient energy and subsequent decay heat from the contain-l ment atmosphere following the hypothesized LOCA or MSLB acci-I dent inside the containment to maintain the containment below the design tressure. Both analyses assume the single failure which resu';s in the minimum containment cooling capability.

Curves s1 wtag sump temperature, heat generation rates, heat removal 16 e. of the containment heat removal systems, and containment .Stal pressure, vapor pressure, and temperature as a function of time for minimum engineered safety features

.,' performance are givon in Section 6.2.1. The containment cooler heat removal rates as a function of containment tempera-ture and pressure are given in Figure 6.2.1-15. This data l has been furnished by American Air Filter and is supported by their topical report (Ref. 1). A constant essential service water temperature of 95 F at the coil inlet has been assumed.

l This is the maximum conservatively calculated temperature that would exist at any time during the accident. The assumptions l used in calculating this temperature are discussed in Section l 9.2.5.

l l

SAFETY EVALUATION EIGHT - The containment cooler fan / motor combination is qualified to operate during the DBA, in accor-

/ dance with IEEE 334, 1974. Section 6.2.2.2.2.2 provides the basis for the assumption of structural integrity of the cooler i housing and discharge plenum during a DBA. American Air Filter (Ref. 1) demonstrates the compatibility of the housing l and plenum materials with the DBA environment.

i 6.2.2.2.4 Tests and Inspections Preoperational testing is described in Chapter 14.0. One containment cooler fan is tested in accordance with AMCA Standard Test Code 211, " Certified Rating for Air-Moving

^ Devices."

-*~~ The analytical data used to predict coil performance for both normal and DBA conditions are based upon the tests and data in Reference 1.

I Major components are accessible during normal plant operation ,

l for inspection, maintenance, and periodic testing. I

6.2.2.2.5 Instrumentation Applications Each containment cooler is monitored for leaving air temper-ature and fan vibration via the plant computer. In addition,

_5 containment air temperature will also be monitored in the area

%)

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CALLAWAY - SP O^'

of each containment cooler intake. Direct control room indi- .]'

cation is provided for the inlet air temperatures. The leaving ~/ '

. air temperature can be_ displayed in the control room vi a the plant computer.

Each containment cooler fan is operable from the control room.

'6.2.

2.3 REFERENCES

1. Topical Report AAF-TR-7101, " Design and Testing of Fan cooler-Filter Systems for Nuclear Applications"; February 20, 1972; American Air Filter Co., Inc.; Louisville, KY.

2. Topical Report OCF-1, " Nuclear Containment Insulation ')

t System,"-August 1977, Owens-Corning Fiberglas Corpora- , ",._

tion, Lenexa, KS.

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)

Rev. OL-0 6.2.2-22 6/86 e

L .,:,...

L- O v.-i- CALLAWAY - SP 6.2.3 SECONDARY CONTAINMENT FUNCTIONAL DESIGN i

Based on the fission product removal and control systems dis-l cussed in Section 6.5 and the radiological consequences analyzed ')

l il in Chapter 15.0 following a LOCA, no secondary containment is required for SNUPPS.

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Rev. OL-0 6.2.3-1 6/86 s

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CALLAWAY - SP ,

TABLE 6.2.2-l' COMPARISON OF THE RECIRCULATION SUMP DESIGN WITH EACH OF THE POSITIONS OF REGULATORY GUIDE.1.82 Regulatorv Guide 1.82 Position Recirculation Sumo Design 4

m

1. A minimum of two sumps should be provided, Two sumps are provided, and each has suf*icient each with sufficient capacity to serve one capacity to serve one of the redundant halves of the redundant halves of the ECCS and CS of the ECCS and CS systems.

, systems.

2. The redundant sumps should be physically The redundant sumps are physically separated separated from each other and from high from each other and from high energy piping.

energy piping systems by structural barriers, to the extent practical, to  !

preclude damage to the sump intake filters  !

by whipping pipes or high-velocity jets of {

water or steam. i

3. The sumps should be located on the lowest The sumps are located in El. +2,000, which is floor elevation in the containment exclusive the lowest floor elevation in the reactor of the reactor vessel cavity. As a minimum, building, exclusive of the reactor cavity. The i the sump intake should be protected by two sump intake is protected by-three screens: (1) screens: (1) an outer trash rack and (2) a an outer trash rack which is grating, (2) fine inner screen. The sump screens should coarse industrial wire mesh with 1/2-inch ,

not be depressed below the floor elevation. openings, and (3) a fine inner screen of fine industrial wire mesh with 1/8-inch openings. i i

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1 Rev. OL-0 6/86

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i; CALLAWAY - SP TABLE 6.2.2-1 (Sheet 2) l Regulatory Guide 1.82 Po'sition Recirculation Sump Design i

.4. The floor level in the vicinity The floor is level in the vicinity of of the coolant sump location the sump. However, a 6-inch concrete  :

should_ slope gradually down curb is provided on which the screen i away from the sump. is supported to prevent high density particles from entering the sump. 1

5. All drains from the upper regions All drains in the upper regions of i of the reactor building should the reactor building are terminated terminate in such a manner that in such a manner that direct streams  ;

direct streams of water, which of water whica may contain entrained may contain entrained debris, debris will not impinge on the filter will not impinge on the filter assemblies.  ;

assemblies. ,

6. A vertically mounted outer trash A vertically mounted outer trash rack rack should be provided to prevent is provided to prevent large debris i large debris from reaching the fine from reaching the fine inner screen.

inner screen. The strength of the This trash rack is designed to with-trash rack should be considered in stand the differential pressure which protecting the inner screen from would exist if it were 100. percent missiles and large debris. clogged.

7. A vertically mounted fine inner A vertically mounted fine inner screen '

screen should be provided. The is provided. Table 6.2.2-9 provides design coolant velocity at the the coolant velocities for a large LOCA inner screen should be approximate- and an MSLB at several locations and 6 cm/sec (0.2 ft/sec). The avail- at several operational times. The able surface area used in deter- intent of item 7 is met.

mining the design coolant velocity should be based on one-half of the free surface area of the fine inner '

screen to conservatively account for partial blockage. Only the vertical screens should be considered in

' determining available surface area.

Rev. OL-0 6/86 J

O CALLAWAY - SP .o TABLE 6.2.2-1 (Sheet 3) **

Regulatory Guide 1.82 Position Recirculation Sump Design

8. A solid top deck is preferable, The top deck is solid. The present and the top deck should be designed expected water elevation in the reactor to be fully submerged after a LOCA building following a LOCA does not and completion of the safety in- reach the top deck of the containment jection. recirculation sump; however, the design is for full submergence.

9. The trash rack and screens should The trash rack and screens are designed be designed to withstand the vi- to be seismic Category I.

bratory motion of seismic events without loss of structural integrity.

10. The size of openings in the fine The size of the particle which could screen should be based on the pass through the inner screen is less minimum restrictions found in than 1/8 inch. The containment spray systems served by the sump. The pump is designed to pass particles minimum restriction should take less than 1/4 inch in size, and the into account the overall opera- minimum restriction in the spray bility of the system served. system is the 7/16-inch orifice in the spray nozzle.

11. Pump intake locations in the The pump intake location in the sump sump should be carefully con- is horizontal to limit any degrading sidered to prevent degrading effects due to vortexing.

e ffects , such as vortexing on the pump performance.

Rev. OL-0 6/86

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CALLAWAY - SP TABLE 6.2.2-1 (Sheet 4)

Regulatory Guide 1.82 Position Recirculation Sump Design

12. Materials for trash racks and The materials used for the trash racks screens should be selected to and screens are selected to withstand avoid degradation during periods 40 years of no operation, and will of inactivity and operation and withstand the environment of the should have a low sensitivity to reactor building following a LOCA.

adverse effects, such as stress-assisted corrosion, that may be induced by the chemically reactive spray during LOCA conditions.

13. The trash rack and screen structure An access opening is provided to facil-should include access openings to itate inspection of the sump.

facilitate inspection of the struc-ture and pump suction intake.

14. Inservice inspection requirements Inservice inspection requirements for coolant sump components (trash consist of visual examination during racks, screens, and pump suction every scheduled refueling downtime.

inlets) should include the fol-lowing:

a. Coolant sump components should be inspected during every refueling period downtime, and

b. The inspection should be a visual examination of the com-ponents for evidence of struc-tural distress or corrosion.

Rev. OL-0 6/86

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