ML17261B020

Design Criteria Ginna Station Steam Generator Containment Penetration.

ML17261B020
Person / Time
Site:	Ginna
Issue date:	03/14/1990
From:	Robert Davis ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17261B019	List: ML17261B018 ML17261B020
References
EWR-4998, EWR-4998-R02, EWR-4998-R2, NUDOCS 9003270212
Download: ML17261B020 (29)

Text

Design Criteria Ginna Station S/8 Containment Penetration Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 EWR 4998 Revision 2 March 14, 1990 Prepared by:

/~ 8/14/9o DATE Mechanical Engineer Reviewed by: Z-iv-'

OATE Quality Assurance Engineer Approved by:

M ager, Mech cal Engineering ATE 90032702i2 9'003i4 PDR ADOCK 05000244 PDG DC-6 page ll i

42 92

Revision Status Sheet Latest Latest Latest Page Rev. Page Rev. Page Rev.

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Design Criteria EWR 4998 Page ii Revision 03/14/90 Date 42.91

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Desi n Criteria 1.0 Summar Descri tion of the Desi n Summary 1.1.1 The Steam Generator Inspection/Maintenance (SGI/M) cabling is routed through the equipment hatch during the Annual Inspection and Outage (AI&0).

1.1.2 Generic Letter GL-88-17 addresses the potential for core uncovery to occur due to loss of Decay Heat Removal (DHR). As part of RG&E's response to GL-88-17 a commitment was made to achieve containment closure within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following initial loss of DHR. (Letter from R. Mecredy to C. Stahle, dated 2/1/89).

1.1.3 Routing the SGI/M cabling through the equipment hatch is no longer an acceptable practice for the following reason. In the event of loss of DHR, disconnecting cables and re-assembling the hatch would require considerable effort which could exceed the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time requirement.

1.1.4 The proposed design involves modification to the existing containment penetration No. 2 (spare mechanical penetration) to provide access into containment for the SGI/M cabling during AI&0's. This modification would facilitate the re-assembly of the containment hatch within the required time and allow the SGM/I cabling exclusive use of a penetration. The work scope includes:

1.1.4.1 Removing the capped ends of the existing penetration.

1.1.4.2 Installing removable flanges or other closure devices and test connections.

1.1.4.3 Evaluating the impact of the changes to the penetration.

1.1.5 The installation of the proposed modification shall be performed in two phases. Phase 1 includes the work performed outside containment and Phase 2 includes the work performed inside containment. Phase 1 could be completed prior to the AI&O and Phase 2 during the first days of the AI&0. A satisfactory leak test per Section 23.1 (to limits specified in 1.3.1) shall be performed prior to the start of Phase 1.

1.2 Functions 1.2.1 The function of the penetration closure assembly is to provide a reliable, leak-testable sealed closed barrier Design Criteria Revision 0 Page 1 EWR 4998 Date 10/19/89

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at the in and outboard ends of containment penetration No. 2.

1.3 Performance Requirements 1.3.1 The leakage for both inboard and outboard closures at penetration No. 2 shall be limited by the leakage requirements given in section 4.4 of the Ginna Technical Specifications, and section 6.2.6 of the UFSAR for Ginna.

1.4 Control 1.4.1 The penetration closure assemblies will be manually removed and reassembled during plant outages when required.

1.5 Modes of Operation 1.5.1 Normal Operation 1.5.1.1 The closure assemblies at penetration No. 2 will remain in place, until the reactor is in the cold shutdown mode per Technical Specifications 3.6.1.

1.5.2 Emergency Operation 1.5.2.1 ln the event a loss of DHR is initiated during cold shutdown the closure assemblies shall be capable of achieving containment closure as defined in the reference 2.12 letter, within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame commited in the reference 2.11 letter.

2.0 Referenced Documents 2.1 USNRC Documents 2.1.1 Regulatory Guide 8.8 "information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be as Low as is Reasonably Achievable (ALARA)"

2.1.2 Regulatory Guide 1.26 "Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste-Containing Components of Nuclear Power Plants".

2.1.3 Regulatory Guide 1.29 "Seismic Design Classification".

2.1.4 Standard Review Plan NUREG-0800, Rev. 2, dated July 1981.

2.1.5 Generic Letter No. 88-17, dated October 17, 1988.

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2.2 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 1986 Edition.

2.2.1 Section II Material Specification.

2.2.2 Section III Nuclear Power Plant Components.

2.2.3 Section IX Welding and Brazing Qualification.

2.3 American National Standard Institute (ANSI).

2.3.1 ANSI/ASME NQA-2, 1986 Edition "Quality Assurance Requirements for Nuclear Power Plants".

2.3.2 ANSI B16.5-1988 "Pipe Flanges and Flanged Fittings".

2.3.3 ANSI N45.4-1972 "Leakage-Rate Testing of Containment Structures for Nuclear Reactors".

2.3.4 ANSI/AWS D1.1-1988 "Structural Welding Code".

2.3.5 ANSI/ANS 56.2-1984 "Containment Isolation Provisions for Fluid Systems" 2.3.6 ANSI/ANS 56.8-1987 "Containment System Leakage Testing Requirements" .

2.3.7 ANSI/ASME NQA-1, 1986 Edition, "Quality Assurance Program Requirements for Nuclear Facilities".

2.4 R. E. Ginna Nuclear Power Plant Technical Specifications.

2.5 Updated Final Safety Analysis Report (UFSAR) for R.E.

Ginna Nuclear Power Plant.

2.6 American Institute of Steel Construction AISC "Steel Construction Manual" 8th Ed.

2.7 American Concrete Institute "Code Requirements for Nuclear Safety Related Structures" ACI 349-85.

2.8 American Society for Nondestructive Testing, Recommended Practice SNT-TC-1A 1980 Ed.

2.9 Institute of Electrical and Electronics Engineers, IEEE Std. 383-1980 "Standard for Type Test of Class 1E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations".

2.10 Code of Federal Regulations 10 CFR 50 1988 Ed.

2.11 RG&E letter from R.C. Mecredy to C. Stahle (NRC), dated February 1, 1989.

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2. 12 Westinghouse transmittal "WOG-88-156", dated 11/7/88.

"Transmittal of Mid-Loop Operations Interim Guidance and Workshop Attendance List".

2.13 Ginna Station, Quality Assurance Manual.

3.0 Seismic Cate or 3.1 Consistent with USNRC Regulatory Guide 1.29 this modification is seismic category I.

4.0 ualit Grou /Code Class 4.1 Regulatory Guide 1.26 does not consider nor classify containment-penetration assemblies therefore, no code class is required, however, ASME Code Class 2 will be used as guideline for this modification.

5.0 Electrical S stem Safet Classification N/A 6.0 A Pro ram A licabilit 6.1 The QA program requirements of the Ginna QA Manual shall be applicable to this modification.

7.0 Codes Standards and Re ulator Re uirements The following requirements shall apply to the design and installation of the modification of containment penetration No. 2:

7.1 The design, materials, fabrication, installation, examination and testing of the penetration closure assembly shall be in accordance with the requirements of ASME Section III, Subsection NC.

7.2 The design, fabrication and erection of any closure assembly support shall be in accordance with the requirements of the AISC Specification.

8.0 Desi n Conditions 8.1 Inside Containment Design pressure 60 psi (UFSAR Table 3.11-1)

Design temperature 286 F (UFSAR Table 3.11-1) 8.2 Outside Containment Design pressure = atmospheric Design temperature = 2 F min. + 91 F max.

(UFSAR sect. 2.3.2.1)

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8.3 The proposed penetration modification shall be designed for the most severe (inside containment) condition, however the outside containment min. temperature shall also be considered.

9.0 Load Conditions 9.1 The penetration shall be evaluated to withstand, pressure, deadweight, thermal and seismic loading conditions.

10.0 Environmental Conditions 10.1 The following environmental conditions shall be considered for this modification inside containment:

Normal Accident Temperature 60'F-120'F 286'F Pressure 0 psig 60 psig Humidity 50: (nominal) 100%

Radiation lrad/hr see tables 3.11-2 and 3.11-3 ref. 2.5 Chemical Spray Solution of boric acid (2000-3000 ppm boron) solution ph 8 10 Flooding 7 ft., maximum submergence elevation is 242'-8" 10.2 The following environmental conditions shall be considered for this modification outside containment:

Normal Temperature 2 Fg +91 F Pressure atmospheric Humidity 10'00 o 10.3 Severe Weather Conditions The effect of severe weather conditions shall be considered in the design. This shall include consideration of environmental conditions associated with routing or locating hardware (equipment, components, piping, tubing, valves, instruments, panels, cabinets, etc.) near exterior walls, doors louvers or other openings.

10.4 Proximity of Existing Systems or Equipment The potential for localized environmental conditions (contagious or intermittment) due to proximity of Design Criteria Revision 1 Page 5 EWR 4998 Date 03/02/90

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adjacent hardware (equipment, components, pipings, tubing, valves, instruments, panels, cabinets, conduit, etc.) shall be considerd in the design. This shall include the effects of routing (or locating) new (or modified) hardware near existing hardware which operates at high temperatures.

11.0 Interface Re uirements 11.1 All modifications to existing penetration shall not degrade existing structure nor compromise containment.

integrity.

12.0 Material Re uirements 12.1 Materials used for modifications to the existing penetration shall be compatible with existing materials.

13.0 Mechanical Re uirements 13.1 All mechanical devices shall be installed such that containment integrity is not compromised.

14.0 Structural Re uirements 14.1 The need for missile protection for the penetration shall be evaluated.

14.2 The structural integrity of the proposed modified containment penetration shall be evaluated.

14.3 Any modifications made to existing concrete shall comply with the applicable requirements of the ACI 349 code.

15.0 H draulic Re irements None 16.0 Chemistr Re irements Design of items located inside containment shall include consideration that they may be subject to a chemical spray as specified in section 10.

17.0 Electrical Re uirements None 18.0 0 erational Re irements 18.1 During cold shutdown this modification shall permit this penetration to achieve containment closure (as Design Criteria Revision 1 Page 6 EWR 4998 Date 03/02/90

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defined in the reference 2.1.5 letter) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of initial loss of DHR as committed in the reference 2.11 letter.

18.2 The modification to the penetration shall not degrade existing containment integrity requirements.

19.0 Instrumentation and Control Re irements 19.1 Any instrument added to the penetration shall not degrade the function of the penetration.

20.0 Access and Administrative Control Re uirements 20.1 The need for anti-tampering features in order to maintain containment integrity shall be reviewed.

20.2 All pertinent plant/maintenance procedures shall be revised, that may if necessary, to incorporate any requirements result as part of this modification.

21.0 Redundanc Diversit and Se aration Re irements 21.1 Modification to the penetration shall maintain existing redundancy, diversity and separation requirements.

22.0 Failure Effects Re uirements 22.1 The penetration shall remain functional following a safe shutdown earthquake.

23.0 Test Re uirements 23.1 The penetration shall undergo Appendix J Testing in accordance with the provisions of 10CFR50.

23.2 Upon'completion of phase 1, a simulated and unannounced test shall be performed to verify that SGI/M cabling can be disconnected and penetration closed within the required 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> using work crews who will normally be assigned to perform the function. A satisfactory containment closure test shall be performed after penetration has been assembled in this manner.

24.0 Accessibilit Maintenance Re air and Inservice Ins ection Re irements 24.1 This modification shall not degrade existing penetration accessibility, maintenance, repair, and inservice inspection requirements. New welds and supports shall be located, in a manner so they are accessible for in-service inspection.

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25.0 Personnel Re irements 25.1 Welding associated with the safety-related mechanical portion of this modification (such as piping, pipe supports, equipment supports, tubing supports, etc.)

shall be performed by personnel who are qualified in accordance with the requirements of ASME Section IX.

25.2 Welding associated with the structural, and non-safety related mechanical portion of this modification shall be performed by personnel who are qualified in accordance with the requirements of either ASME Section IX or AWS Dl.l.

25.3 All NDE personnel, where required in performing the modifications, shall be qualified in accordance with the requirements of ANSI NQA-1 and/or SNT-TC-1A.

26.0 Trans ortabilit Re uirements None 27.0 Fire Protection Re uirements 27.1 Electric cable used in the modification shall meet the flame test requirements of IEEE 383-1974.

27.2 The modifications shall not degrade existing fire barriers or affect the performance of any existing fire protection equipment.

27.3 All new penetrations through fire barriers shall be sealed with appropriate fire seals.

27. 4 Where cable installation requires cable insertion through a silicone foam fire stop or seal, care must be taken to prevent loss of seal integrity. Since the seal is established by the pressure of the silicone foam on the cable, rather than by adhesion, a new cable may be inserted if there is sufficient space available.

After the cable has been fully inserted, the seal shall be visually inspected on both sides to assure that the seal is still fully intact. If visual inspection reveals any damage to the seal, removed and replaced.

it shall be repaired or 27.5 All new penetrations through walls, ceilings, or floors which are not fire barriers but provide a Halon fire Suppression System Boundary shall be sealed with appropriate air seals.

27.6 Material used in this modification shall not increase the probability or consequence of a fire.

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27.7 A review shall be performed to ensure the modified system continue to comply with 10CFR50 Appendix R requirements.

28.0 Handlin Re uirements Material required for this modification shall be handled in accordance with the requirements of ANSI/ASME NQA-2-1986 Edition.

29.0 Public Safet Re irements None 30.0 ~1'1 None Personnel Safet Re uirements 31.1 The design and installation of all modifications shall comply with the applicable requirements of USNRC Regulatory Guide 8.8. ALARA shall be addressed for any portion of the work; in particular, for any work required in a high radiation area. Applicable typical survey maps of the appropriate areas are attached and are considered to be part of this criteria.

32.0 Environmental uglification Re uirement None 33.0 Uni ue Re uirements 33.1 Operational Experience Relevant experience from Ginna and other facilities should be considered during the design process.

Applicable NRC bulletins, notices, and INPO SOERs, as a minimum, should be requested from the Operational Assessment Group at Ginna for review during the development of the design.

33.2 Provisions shall be taken so that the closure assembly is re-assembled within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time requirement.

Design Criteria Revision 1 Page 9 EWR 4998 Date 03/02/90

Safety Analysis Ginna Station S/G Containment Penetration Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 EWR 4998 Revision 2 Narch 14, 1990 Prepared by:

3 /4/9o DATE Nechanical Engineer Reviewed by: H I+So DATE Nuclear Engi eer Approved by:

~'f- 'F o DATE Nanager, Nuclear Engineering Page i SA3 42.92

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Revision Status Sheet Latest Latest Latest Page Rev. Page Rev. Page Rev.

Safety Analysis

'ERR 4998 Page ii Rev<s<on 03/14/90 Date 42 9'I

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Safet Anal sis 1.0 Description of Scope Summary 1.1.1 The Steam Generator Inspection/Maintenance (SGI/M) cabling is routed through the equipment hatch during the Annual Inspection and Outage (AI&0).

1.1.2 Generic Letter GL-88-17 addresses the potential for core uncovery to occur due to loss of Decay Heat Removal (DHR). As part of RG&E response to GL-88-17 a commitment was made to achieve containment closure within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following initial loss of DHR. (Letter from R. Mecredy to C. Stahle, dated 2/1/89).

1.1.3 Routing the SGI/M cabling through the equipment hatch is no longer an acceptable practice for the following reason. In the event of loss of DHR, disconnecting cables and re-assembling the hatch would require considerable effort which could exceed the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time requirement.

1.1.4 The proposed design involves modification to the existing containment penetration No. 2 (spare mechanical penetration) to provide access into containment for the SGI/M cabling during AI&0's. This modification would facilitate the re-assembly of the containment hatch within the required time and allow the SGI/M cabling exclusive use of a penetration.

1.1.5 The work scope associated with the proposed modification includes:

1.1.5.1 Removing the capped ends of the existing penetration.

1.1.5.2 Installing removable flanges or other closure devices.

1.1.5.3 Evaluating the impact of the changes to the penetration.

2.0 References 2.1 Design Criteria for EWR 4998, Rev. 2, dated 03/14/90 "S/G Containment Penetration".

2.2 Ginna Station Updated Final Safety Analysis Report.

2.2.1 Section 3.1.2.1, "Overall Requirements".

2.2.2 Section 3.1.2.5, "Reactor Containment".

2.2.3 Section 3.8.1.5, "Penetrations".

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2.2.4 Section 6.2.4, "Containment Isolation System".

2.2.5 Section 6.2.6, "Containment Leakage Testing".

2.3 R. E. Ginna Nuclear Power Plant Technical Specifications.

2.3.1 Section 3.6, "Containment System".

2.3.2 Section 4.4, "Containment Tests".

2.3.3 Section 5.2.2, "Penet'rations".

2.4 USNRC Generic Letter No. 88-17, dated October 17, 1988.

2.5 ANSI/ANS 56.2-1984, "Containment Isolation Provisions for Fluid Systems after a LOCA".

3.0 Safet Anal sis 3.1 A review has been made of the design basis events to determine those related to the proposed modification.

The events associated with this work are:

3.1.1 a) Fires b) Seismic Events c) Natural Phenomena d) Missile Hazards e) Containment Integrity 3.2 The following assessment is made:

3.2.1 The probability and consequences of a fire have been addressed in Section 27.0 of the modification design criteria. As described in the criteria, fire barriers will not be degraded and material used. will meet criteria equal to or greater than those presently installed. This modification will be reviewed against the assumptions of 10CFR50 Appendix R to assure continued compliance. Therefore, the modifications will not alter the area fire loading, the sources of fire initiation, nor the acceptability of the consequences of a fire.

3.2.2 Section 3.1 of the 'Design Criteria classifies the penetration as Seismic Category I. The penetration modification will be evaluated, as addressed in sections 9.0, 11.0 and 14.0 of the modification design criteria, to withstand seismic loading and to maintain the structural integrity of the existing penetration.

3.2.3 Section 10.3 of the Design Criteria requires that the effects of severe weather conditions be considered in the design of the penetration modification. It is also Safety Analysis Revision 2 Page 2 EWR 4998 Date 03/14/90

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required, in section 11.1, that the existing structure not be degraded by the proposed modification.

3.2.4 The need to evaluate for missile protection for this proposed penetration modification is addressed in section 14.1. of the Design Criteria.

3.2.5 Sections 11.1 and. 13.1 of the modification Design Criteria require the proposed modification not to compromise containment integrity.

3.2.6 Phase 1 of the proposed penetration modification can be completed prior to cold shutdown since the welded. cap assembly inside containment is not affected. This welded cap assembly ensure that containment integrity is maintained and can be considered an extension of the containment liner (see Section 3,2 of reference 2.5).

The penetration is also required to be tested under 10 CFR 50 Appendix J testing criteria prior to initiation of work to verify the integrity of the welded cap assembly inside containment (section 1.1.5 of the Design Criteria) .

Preliminar Saf et Evaluation 4.1 The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, previously evaluated in the Safety Analysis Report are not increased.

4.2 The possibility of an accident or malfunction of a type different from any previously evaluated in Safety Analysis Report has not been created.

4.3 The margins of safety as defined in the basis for any technical specification is not reduced.

4.4 Therefore, the proposed modification does not involve an unreviewed safety question. No changes to the Technical Specifications are required as the result of the proposed modif ication to maintain the present margins of safety.

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