ML20078E578
| ML20078E578 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 11/09/1994 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20078E572 | List: |
| References | |
| SE-403037-001, SE-403037-001-R00, SE-403037-1, SE-403037-1-R, NUDOCS 9411140160 | |
| Download: ML20078E578 (30) | |
Text
l ENuolear Technical Functi:ns Safety / Environmental Determination and 50.59 Review (EP-016) 1 Unit Oyster Creek NGS Page 1 of 29 Document / Activity Title Reactor Vessel Shroud Repair SE Rev. No. O Document No. (if app // cab /e/ OC-CCD-403037-002 Doc. Rev. No. O SE No. SE-403037-001 Type of Activity (modification, procedure, test, experiment, or document /;
Modification
- 1. Does this document involve any potential non-nuclear environmental concern? O Yes M No To answer this question, review the Environmental Determination (ED) form. Any YES answer on the ED form requires an Environmental impact Assessment by Environmental Controls, per 1000-ADM-4500.03. If in doubt, consult Environmental Controls or Environmental Licensing for assistance.
If all answers are NO, further environmental review is not required. in any event, continue with Questions 2, below.
- 2. Is this activity /docum nt listed Section I or il of the matrices in Corporate Precedure QD Yes O No 1000-ADM-1291.017 If the answer to question 1 is NO, stop here. This procedure is not applicable and no documentatior; is required. (If this activity / document is listed in Section IV of 1000-ADM-1291 review on a case-by-case basis to determine applicability.) If the answer is YES, proceed to question 3.
- 3. Is this a new activity / document or a substantive revision to an activity / document? DD Yes O No
(%e Exhibit 2, paragraph 3, this procedure for exams,les of non-substantive changes.)
If the answer to question 3 is NO, stop here and complete the approval section below. This procedure is not applicable and no documentation is required. If the answer is YES, proceed to answer all remaining questions. These answers become the Safety / Environmental Determination and 50.59 Review.
- 4. Does this activitv/ document have the potential to adversely affect nuclear safety 00 Yes O No or safe plant operations?
- 5. Does this activity / document require revision of the system / component description DD Yes G No in the FSAR or otherwise require revision of the Technical Specifications or any other part of the SAR7
- 6. Does the activity / document require revision of any procedural or operating description O Yes DD No in the FSAR or otherwise require revision of the Technical Specifications Or any other part of the SAR7
- 7. Are tests or experiments conducted which are not described in the FSAR, the O Yes 00 No Technical Specifications or any part of the SAR7 IF ANY OF THE ANSWERS TO QUESTIONS 4,5,6, OR 7 ARE YES, PREPARE A WRITTEN SAFETY EVALUATION FORM.
j If the answers to 4,5,6, and 7 are NO, this precludes the occurrence of an Unreviewed Safety Question or Technical Specifications change. Provide a written statement in the space provided below (use back of sheet if necessary) to support the determination, and list the documents you checked.
NO,because: see attached documentation Documents checked: see attached documentation
- 8. Are the design curria as outlined in TMI-1 SDD-T1-000 Div.1 or SDD-OC-000 Div.1 O Yes 00 No Plant Level Critr a affected by, or do they affect the activity / document?
If YES, indicat how resolved: N/A APPROVALS (print name and sigt)
D - r/GrioNtor M%M Ph L'Ja.,( g (adit.) {, , Data i bat-M
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f r w sible Technscot Revower Date Odei Reviewer (s) Date
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9411140160 941109 PDR ADOCK 05000219 P PDR
l duclear Technical Functions Sofety Evaluation (EP-016)
UNIT Ovster Creek NGS PAGE 2 of 29 ACTIVITY / DOCUMENT TITLE Reactor Vessel Shroud Repair _ SE No sE-403037-oot Rev. No. O DOCUMENT NO. (if applicable) OC-CCD-403037-002 Rev. No. O Type of Activity / Document Modification (Modification, prucedure, test, experiment, or document)
This Safety Evaluation provides the basis for determining whether this activity / document involves ,
an Unreviewed Safety Question or impacts on nuclear safety.
Answer the following questions and provide reason (s) for each answer per Exhibit 7. A simple statement of conclusion in itself is not sufficient. The scope and depth of each reason should be commensurate with the safety significance and complexity of the proposed change.
- 1. Willimplementation of the activity / document adversely affect nuclear '
safety or safe plant operations? O Yes DDNo The following questions comprise the 50.59 considerations and evaluation to determine if an Unreviewed Safety Question exists:
- 2. Is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report increased? O Yes 00 No
- 3. Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report created? O Yes 00 No
- 4. Is the margin of safety as defined in the basis for any Technical Spc.::ification reduced? O Yes DDNo if any answer is "yes" an impact on nuclear safety of an Unreviewed Safety Question exists. If an adverse impact on nuclear safety exists revise or redesign. If an unreviewed safety question with i no adverse impact on nuclear safety exists forward to Licensing with any additional documentation to support a request for NRC approval prior to implementing approval.
- 5. Specify whether or not any of the following are required, and if "yes" indicated how it was resolved Yes TFAAl/PFU/Other No
- a. Does the activity / document require X an update Explain: The FSAR should be revised to describe the ohysical addition of tie-rods to reactor intemals
- b. Does the activity / document require X a Technical Specification Amendment?
Explain: No. Chances will not affect operations and coeratino limits of the olant.
SE-4 03 03 '7
- M I Re e 0 1
", PAGE 3 OF 29 Yes TRTTFWR/Other No
- c. Does the activity / document require X a Quality Classification List (OCL) Amendment?
Explain: Details of reactor vesselinternals are not cart of OC list
- d. Other: (If none, use NA) N/A This form with the reasons for the answers, together with all applicable continuation sheets constitutes a written Safety Evaluation.
List of Effective Paaes Pace No. Rev. No. Estae No Rev. No. Pace No. Rev. No.
1 0 11 0 21 0 2 0 12 0 22 0 3 0 13 0 23 0 4 0 14 0 24 0 5 0 15 0 25 0 6 0 16 0 26 0 7 0 17 0 27 0 8 0 18 0 28 0 9 0 19 0 29 0 10 0 20 0 Appendix A 0 Approvals (Print Name and Slgn) Date Engineer / Originator @AO g MMfggt@pg,j g _
1( *l-94 Sectian Manager h d_Y O _j jh M l t-%q Responsible Technical Reviewer q C. j)Av/d C.9/d#) //"9 #)/f-I Independent Safety Reviewer h [ [g g,M, y 4 } ///9/q9 Other Reviewer (s) $ pe h [ [ [ g g (ej
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tDJ $9-1994 17:43 FROM NOB J.J. IERTOt4-OctG3 TO 39507444 P.01
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SE-403 03 9 ~O0 I Lo PAGESOF 29 Yes TR/fFWRJDther Na
- c. Does the scisvityttiocument respara X a Questy cussiflaselon ust IOcu Ansodment7 Emple'en _Dagais.1d.h _ "' a-- -l; are not man af Oc ramt
- d. Other: W none, use MA) M This iofm with 94 sessens for the answert, together with e5 appEcable continustion shams caruftetutes a wntten Safety livahasiH,n.
List c1 Effecteve Pants Page N L kv.. NA Phast NsL Ret No. Paat Nn. May. No.
1 0 11 0 21 0 2 0 12 0 22 0 3 0 13 0 23 0 4 0 14 0 24 0 l 5 0 15 0 26 0 '
S O 16 0 26 0 ;
7 0 17 0 27 0 8 0 18 0 28 0 9 0 19 0 29 0 10 0 20 0 Appen&c A 0 n, eum <w as.,
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a l PAGE 4 OF 29
> Nesoleer ENVIRONMENTAL DETERMINATION Division / Unit Document No. Revision No.
Technical Functions Oy:ter Creek, NGS SE-403037-001 0 )
l Document Title '
Reactor Vessel Shroud Repair
- 1. Wii Document implementation result in an increased potential to release O Ye- DD No haza edous chemicals (gas, liquid, solid, or semi-solid) to the environment?
o
- 2. Will Docunant implementation compromise existing capability to control, O Yes 00 No treat, or monitor releases to the environment? l
- 3. Will Docum'ent implementation cause a physical or chemical change in the O Yes DONo characteristics of facility discharges, effluents, or withdrawals?
- 4. Will Document implementation result in the permanent or temporary 0 Yes DD No ,
storage (for use, disposal, or transfer) of any hazardous or other regulated waste, or hazardous chemical (s), outside of established ;
handling facilities or procedures where the margin of control or containment will increase the potential of a release to the environment? *
- 5. Will the implementation of the Document result in an increase in the O Yes 00 No amount or a change in the type of hazardous waste (s) typically generated, and/or previously evaluated for the type of activity?
- 6. Does Document implementation result in land disturbance O Yes DDNo (e.g., excavation work, graoing), or modification or alteration of storm water drainage systems that would change site storm water runoff or increase sediment loading of storm water runoff 7
- 7. Will Document implementation result in a physical alteration to a O Yes DD No ,
wastewater treatment facility or other facility system (s) or component regulated by environmental permits (e.g., discharge to groundwater permit, discharge to surface water permit, etc.17
~ SIGNATURES Preparer Date '
H Cg -W Cognizant Supervisor / Manager, or Designee Date '
Mh) D;Q ,
\ N-H ,
Env onmental T Qa eviewer l Date II
- 9' k m
Other/eviewer(s) Date l l
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P:ge 5 of 29
- No. SE-403037 001 Revision 0 CONTENTS .
Section Eagn 1.0 WIRODUCTION 7 1.1 Purpose
1.2 Background
1.3 Description of Modification 2.0 SYSTEMS AND DOCUMENTS AFFECTED BY MODIFICATION 8 2.1 Systems Affected 2.2 Documents Describing Affected Systems 2.2.1 Design Documents 2.2.2 Modification Design Documents 2.23 Licensing Documents 3.0 EFFECTS ON SAFETY 11 3.1 Documents Defining System Safety Functions 3.2 System Safety Function 3.3 Effects of Modification on the System Safety Function 3.3.1 System Performance 3.3.2 Quality Standards 3.33 Natural Phenomena Protection 3.3.4 Fire Protection 3.3.5 Environmental Qualification 3.3.6 Misn'e Protection 3.3.7 Internal Flooding /High Energy Line Breaks 3.3.8 Electrical Separation / Isolation / Loading 3.3.9 Single Failure Criteria 3.3.10 Separation Criteria 3.3.11 Cnntninment Isolation 3.3.12 Materials Compaubility 3.3.13 Water Impingement Due to Water Type Fire Suppression Systems
Page 6 of 29 No. SE.403037 001 Revision 0 3.4 10CFR50.59 Determination 3.5 FSAR Changes
4.0 CONCLUSION
23 APPENDIX A Figures 24
Page 7 of 29 No. SE-4m037-001 Revision 0 LO INTRODUCIION L1 Purpose The purpose of this safety evaluation is to document that modifications to permanently repair cracking of any and all circumferential welds in the Oyster Creek core shroud do not result in an unreviewed safety question per USNRC 10CFR50.59. This safety evaluation addresses all activities / operations during/after restart.
L2 Backrrou nd The core shroud is a stainless steel cylinder within the reactor pressure vessel that provides lateral support to limit deflections of reactor vessel internals including fuel assemblics, core support plates, control rod guides, etc. The shroud also serves as the partition between the feedwater flowing in the reactor vessel annulus region and the flow through the core.
The heat affected zones of axial and circumferential welds in the shroud are susceptible to intergranular stress corrosion cracking and inadiation-assisted stress corrosion cracking. Cracking of circumferential shroud welds has occuned at several U.S. and foreign BWRs as discussed in GE SIL 572 (Reference 2.2.1.1.1), USNRC Information Notice 93-79 (Reference 2.2.3.3) and USNRC Generic Letter 94-03 (Reference 2.23.4). Recent inspections of the shroud at Oyster Creek have shown intermittent, part through thickness cracking in the H4 weld.
13 Deserlotion of Modification The design of the Oyster Creek shroud repair consists of 10 stainless steel tie-rod / radial restraint assemblies which are installed in the shroud / reactor vessel anmilus, between attachment points near the top of the shroud and the lower shroud support cone. These tie rod assemblies provide vertical restraint to the shroud. A spacer ring is provided which raises the separator flange sufEcient to permit engagement of the upper bracket onto the shroud upper flange. In addition, radial restraints are provided to impart lateral stability to the shroud assembly, as well as limit lateral displacement of the shroud so as to ensure control rod insertion. Together, the tie rods and radial restraints resist both vertical and lateral loads resulting from normal operation and design accident loads, including seismic loads and postulated pipe ruptures.
This design protects against potential through-wall cracking in any and all of the circumferential welds from the H1 weld at the top of the shroud to the H 6s weld near the bottom of the shroud (see Figure 1 for identification
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Page 8 of 29 No. SE-403037-001 Revision 0 of the shroud welds). Welds H7and Hg are already protected by a previous repair discussed in Reference 2.2.3.2. Currently, only weld H4 is known to contain significant circtunferential cracking. Installation of this repair with an appropriate preload minimizes the need for future shroud inspections to ensure that the remaining cir'unferential welds are free of cracks.
Figures 2 through 4 show the configuration and features of the proposed shroud repair for Oyster Creek. The design complies with all requirements specified by the BWR Owners Group in their shroud repair criteria in Reference 2.2.1.4.1.
2.0 SYSTEMS AND DOCUhENTS AFFECTED BY MODIFICATION 2.1 Systems Affected The systems physically affected by the modification are the Reactor Vessel Internals System (222) and the Reactor Vessel (221). j 2.2 Documents Describing Affected Systems 1
2.2.1 Design Documents 2.2.1.1 General Electric Documents 2.2.1.1.1 GE-SIL 572, Rev.1, Core Shroud Cracks.
2.2.1.1.2 GE Spec 21 A1105AF, Rev. O, Reactor !
Pressure Vessel Specification.
2.2.1.1.3 GENE-771-44-0894, Justification of Allowable !
Deflections of the Core Plate and Top Guide - I Shroud Repair. !
2.2.1.1.4 GE Spec 21A-5369, Rev. O, Specifications for the Core Structure.
2.2.1.1.5 GE Report NEDC-32405, Final Test Report, "CRD Performance Evaluation Testing With !
Driveline Misalignment," September 1994.
2.2.1.1.6 GE letter DRF-B11-00604 dated October 25, 1994 to GPUN confirming the applicability of Reference 2.2.1.1.5 to Oyster Creek.
Page 9 of 29 No. SE-403037 001 Revision 0 t-2.2.1.1.7 GE Paper, Kass, J.N., et. al, Stress Corrosion Resistance of XM-19 Stainless Steel, Corrosion. Volume 35, No. 6, June 1979.
2.2.1.2 Combustion Engineering Documents 2.2.1.2.1 CENC-1143, Analytical Report for Jersey -
Central Reactor Vessel 2.2.1.3 Drawings (see Reference 2.2.2.5 for a table of drawings)-
2.2.1.4 BWROG Documents 2.2.1.4.1 BWRVIP letter (J. T. Beckham) .to USNRC dated September 13,1994 forwarding BWR Core Shroud Repair Design Criteria, Revision 1, dated September 12,1994.
2.2.2 Modification Design Documents 2.2.2.1 Letter from A. Collado (GPUN) to S. J. Weems (MPR) dated June 7,1994, forwarding " Design Criteria for OC Core Shroud Repair Clamps."
2.2.2.2 MPR Design Specification 083-9403-001 for OC Core Shroud Repair, Rev. O, September, 28,1994.
2.2.2.3 GPUN Installation Specification for OC Nuclear Generating Station Core Shroud Repair, OC-CCD-403037 002, Rev. O.
2.2.2.4 GPUN Fabrication Requirements for Oyster Creek Core Shroud Tle-Rod Assemblies, SP-1302-12 276, Rev. O, dated July 11,1994.
2.2.2.5 ' MPR Report: Oyster Creek NOS Core Shroud Repair -
Design Report, Revision 1, dated October 1994 (two volumes).
2.2.2.6 GPUN Safety Evaluation for Removal of RPV Internals !
Vibration Brackets, SE-328354-002.
2.2.2.7 GPUN Calculation C-1302-222-5450-007, Assessment of Tie-Rod Modifications on Reactor Pressure Vessel Thermal Hydraulics, dated November 9,1994.
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Page 10 of 29 No. SB-403037 001 Revision 0 2.2.2.8 GPUN Letter C321-94-2133, NRC Generic Letter 9&O3 Response, August 24,1994.
2.23 Licensing Documents 2.2.3.1 Oyster Creek Nuclear Generating Station Updated Final Safety Analyses Report.
2.2.3.2 Oyster Creek Facility Description and Safety Analysis Report, Amendments 3,12,37, and 40.
2.23.3 NRC Information Notice 93-79: Core Shroud Cracking at Beltline Region Welds in Boiling Water Reactors, September 30,1993.
2.23.4 NRC Generic Letter 94-03: Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors, July 25,1994.
2.23.5 GPUN Technical Data Report No.1138.
2.23.6 GPUN submittal C321-91-2345 dated December 30,1991 to NRC on new Site Specific Ground Motion Spectra.
2.2.3.7 GPUN submittal C321-93-2321 dated December 23,1993 to NRC of new In-Structure Response Spectra developed by E Q E.
2.23.8 NRC SER transmitted by NRC letter dated March 18, 1992 which approved new O.C. Ground Motion Spectra.
2.23.9 GPUN submittal C321-94-2133 dated August 24,1994 to NRC in response to Generic Letter 94-03.
2.23.10 GE Report, Justification of Allowable Displacements of the Core Plate and Top Guide Shroud Repair, GENE-771-44-0894, Rev.1, September 2,1994.
2.23.11 GE Letter DRF B11-00604, Applicability of CRD Testing to Oyster Creek, dated October 25,1994.
2.23.12 GPUN Operational Quality Assurance Program.
- .-- - - - -. _ . . . . .- .- . ~ ~ ~ - - . . _-
i Page 11 of 29 No. SFA03037-001 '
Revision 0 2.2.3.13 NRC Report, Integrated Plant Safety Assessment Systematie Evaluation Program - Oyster Creek NGS, NUREG-0822, January 1983.
3.0 EFFECTS ON SAFEIY ,
3.1 Documents Dennine System Safety Functions 3.1.1 " Updated Safety Analysis Report" - Description of the shroud is located in UFSAR Section 3.9.5.1.1.
3.L2 Plant Technical Specifications Bases - 3.3 Reactor Coolant - nis i section discusses the bases for the reactor vessel and providing adequate flow path in the annular space (between RPV and core shroud).
3.2 System Safety Function The safety funcdon of the reactor shroud at Oyster Creek is to provide -
reactor coolant flow boundaries and to limit deflections and deformations to ensure that control rods and the emergency core cooling systems can perform their safety functions during anticipated operational occurrences and transients, so that safe shutdown and decay heat removal are not ,
impaired per the requirements of Reference 2.2.1.4.1. A detailed discussion of these issues is included in Section 3.4 of this report. 1 The reactor vessel system includes a conical shaped support skirt inside the vessel, whose safety function is to provide structural support and reactor coolant flow boundaries at the bottom of the shroud. The shroud repair attachments in the area of this conical support and of oths areas at radial seismic restraints result in stresses within FSAR allowables. Dus, there is no impairment of the reactor vessel system due to the repair. Reactor vessel stresses are covered in Reference 2.2.2.5.
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33 Effects of Modification on the System Safety Function 33.1 System Performance Increased Vessel Downcomer Annulus Pressure Dron - The .
installation of the tie. rod assembly top will decrease the available
. flow area near the top of the downcomer annulus. Analysis (Reference 2.2.2.5, Appendir M) indicates that the pressure drop resulting from this reduction in downcomer flow area is less than 0.04 psi which is considered insignificant. Because the downcomer annulus increases from 12" to 17" below the top bracket and only
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Page 12 of 29 No. SE.403037-001 Revision 0 the tie rod shaft section is included in this section, the effect on
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recirculation system flow and pressure drop is insignificant.
Increased Heat Storare Capacity - The mass of an individual assembly of the tie-rod restraint system weighs approximately 3
1,800 lbs (for ten tie-rods, approximately 37 ft of metal volume which displaces about 275 gallons of water). The total mass added by this shroud repair system and resulting displacement of water is negligible compared to the 230,000 lbs of reactor internals weight and therefore has no significant effect on the heat storage capab2ity of the existing reactor vessel and its internals results.
The principal effects and issues of operating the plant with circumferential shroud welds cracked and tie-rod / radial restraint systems that structuraDy replace these welds are as follows:
. Tie-rod system induced leakage
. Shroud weld xack leakage
. Lateral displacement of shroid
. Vertical separation of shronJ
. Normal operation with fue circumferential shroud wcld having 360' thrn well c ach
. Core spray piping between shroud and reactor vessel
. Tle-rod features to address flow induced vibration and shroud vibration
. Tie-rod features to address the cold feedwater transient
. Materials of construction of the tie rod / radial restraint system
. Potentialimpact of loose parts
. Seismic analysis
. Effect of raising separator / dryer assembly
. Reactor vessel mechanical effects i
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9 Page 13 of 29 No. SB.403037 001 Revision 0 Reactor Resoonse to Shroud Bypass Leakane - The leakage evaluation for various conditions and shroud crack extents are listed in the table below:
Condition Leakage / Evaluation Normal Operation 0-60 gpm depending on extent Transient Conditions of cracking. No significant effect on normal or transient Operating Basis EQ conditions.
Design Basis EO Gaps can open up to 0.6 inches for a short duration.
Main Steam Line Break No specific FSAR limits on Main Steam Line Break w/ EQ 1eaknge. See Reference 2.2.2.8 Recirculation Line Break f r a specific evaluation which concludes ECCS operation is Recirculation Line Break w/ EQ acceptable.
Each of the above issues, as they relate to the safety operation of the plant, are discussed below.
3.3.1.1 Tie-Rod System Induced Leakage !
The installation of the tie-rod shroud repair does agi require the cutting of any access or attachment holes in l the shroud or the steam separator for installation of the tie-rod system. Thus the configuration of the tie-rod I system and the techniques by which it is attached to the l shroud does not cause new icatage paths between the !
areas inside the shroud and the downcomer (i.e., shroud annulus). Accordingly, the performance of the steam separator, fuel thermal margin, ECCS performance, and fuel cycle length are not effected by the physical installation of this tie-rod system.
3.3.1.2 Shroud Weld Crack Leakage The preload on the tie-rod system has been set so that there is no separation during plant operation with any or l all circumferential shroud welds (i.e., welds 3H thru Hg) l with 360* thru wall cracks (Reference 2.2.2.5, l Appendix K). While the preload has been set so there will be no separation during plant operation, some small l amount ofleakage can occur between the area inside the I shroud to the downcomer annulus via assumed thru wall cracks resulting from corrosion (not separation). If the
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. t Pege 14 cf 29 No. SPA 03037 001 Revision 0 l crack is conservatively estimated to provide a 0.001" l
leakage path per weld, then the leakage is approximately 4 gpm for welds above the lower core plate and 18 gpm for welds below the lower core plate. If all seven shroud welds (H3thru Hgg) have 360' thru wall cracks, the leakage is estimated to be approximately 60 gpm or less than 0.05% of total flow. The leakage from welds below the core plate will be water of approximately the same temperature as the downcomer water. Leakage from welds above the core plate to the upper grid plate.will also be similar temperature water. The leakage between welds H3 and H 2which are above the upper grid plate will be a mixture of water and some steam. The '
differential temperature between the water within the shroud and the downcomer annulus during plant - ,
operation is approximately 30*F. Therefore, the effect on recirculation temperature is insigni6 cant.
Also, the effect of such leakage on ECCS performance is sufficiently small to be considered insignificant, t Accordingly, the peak fuct cladding temperature heensmg bases for Oyster Creek for normal operation with no shroud crack leakage is considered applicable. Similarly, ,
no changes are needed to alert operators to the presence of leakage.
3.3.1.3 Lateral Displacement The maximum laterni displacement, of the shroud at both ,
core support plate and upper guide plate of the shroud under normal conditions and for load combinations of DBE, main steam line, and recirculation pipe LOCA,-is
- c. limited to about 3/8 of an inch. This is the total i
clearance between the shroud and vessel wall for the radial restraints on the tie-rods plus a small amount of deflection of the shroud shell a: the restraints. The t bumper clearance is set based on etual field measurements at each restraint locatbn. Thus the radial restraints serve as mechanicallimit stops to control lateral motion of the shroud relative to the reactor vessel-(Reference 2.2.2.3).
The maximum lateral displacement is well within that GE - l test data for the Hatch plant to ensure control rod insertion. The applicability of the Hatch analysis is !
documented in Reference 2.2.3.11. Specifically, l l
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Page 15 of 29 No. SP 403037-001 Revision 0 i
Reference 2.2.3.10 indicates satisfactory scram per technical specification with core support plate permanent misahgnment as much as 0.75 in. Limiting shroud permanent displacements to about 3/8 in. is considered satisfactory on the basis that internals of the shroud are structurally adequate for existing design basis seismic loads and based on the fact that internals loads, including fuel, are not increased due to this repair. See l Section 3.4.L11 for a discussion of seismie loads. With )
regard to the applicability of the Hatch control rod insertion rod data to Oyster Creek, GE letter dated October 25,1994 (Reference 2.2.1.L6) to GPUN i confi-med the applicability of that data to Oyster Creek control rod drives.
The manmum lateral displacement of any cylindrical section of the shroud with all shroud circumferential welds cracked is about 3/8 of an inch. If this unlikely ciremnstance should occur, this displacemem equals about a quarter of the 1.5" shroud wall thickness and such displacement will not significantly affect cooling of l the core since the cylindrical sections still overlap one j another by 1-1/8 inches. l
\ l 3.3.L4 Vertical Shroud Senaration The tie rods preload restrains vertical separation of any or all circumferential shroud welds that have 360* thru l wall cracks. The tie-rod preload (see Section 3.4.1.5) will l prevent separation of any or all circumferential welds !
during normal plant operation. The manmum separation of any weld above or below the core support plate during a main steam line break or recirculation line LOCA plus a DBE is 0.6 inches (Reference 2.? ? 1, Appendix L). -
This is rcuch less than the maximum vertical separation of 13.3 inches determined in Oyster Creek's GL 94-03 submittal (Reference 2.2.3.9) for H 3 through Ha above l
the core support plate and the displacement of 0.7 inches for welds below the core support plate for plant operation of the shroud without tic-rods (Reference 2.2.3.2, Amendment 37).
3.3.L5 Normal Operation Tne tie-rods are installed with a cold preload such that the tie-rods remain tight when the steam separator is placed on the shroud. In addition, this preload will l
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- Page 16 of 29 No. SE 403037-001 Revision 0 ;
assure no separation of any or all cracked welds from H2 through H6B during normal plant operation. Further, if the design basis cold feedwater transient should occur,
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the tie-rod system, tie-rod / shroud attachment points, and the shroud plate between the H 2 and H3 welds (regardless of whether H 2 and H3welds have both failed) remain within their code allowable limits. Also, this preload will ensure that if shroud welds H2 and H 3, as well as H68 , should failin subsequent operation, there is sufficient preload to prevent weld separation due to the change in rigidity of the shroud structure. By providing a preload strategy which address all crack locations and configurations of locations, the need for weld inspection requirements at future plant refueling outages can be reduced.
3.3.1.6 Core Sprav Picing Between Shroud and Reactor Vessel
'Ibe core spray piping is anchored to both the reactor vessel and the shroud. If a main steam line break were to occur and 360' through wall cracking of a circum-ferential shroud weld were present, upward displacement of the shroud relative to the vessel could result. Analyses were performed to determine the resulting stresses in the core spray piping (
Reference:
2.2.2.5, Appendix L). The analyses concluded that the mmrimum stress due to l shroud displacement caused by main steam line break, including seismic loads due to safe shutdown earthquake and differential thermal expansion, is less than the ASME code allowable stress. Accordingly, no pinching or other Dow path restrictions need to be accounted for.
Special steps were taken in the development and checkout of the installation procedures to assure that the tie-rod assemblies would not damage the core spray piping in the annulus area during installation. The l
installation tooling and procedures were checked out in a l
full scale mockup.
3.3.1.7 Tie-Rod Features to Address the Vibration Issue The tie-rods were analyzed and tested to ensure that reactor coolant flow would not induce unacceptable vibration. Results of these analyses and tests are presented in the MPR Design Report (Reference 2.2.2.5, Appendix D). In brief, stresses resulting from Dow induced vibration are small and pose no fatigue concern.
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3.3.1.8 Tie-Rod Features to Address the Cold Feedwater Transient GPUN Technical Data Report 1138 (Reference 2.2.3.5) describes a thermal transient invoMng cold feedwater injection into the shroud annulus. This could result in a maximum temperature difference of 130"F between the shroud and the (cooler) tie rod components, which would cause a tensile load in the tie rods. Special features are provided in the tie rod design to accommodate this temperature difference (specifically, a variable spring design that minimim rod stiffness for thermal tran ients, when stiffness is undesirable, and increases stiffness and strength for large pressure drop transients when such strength is needed). Appendix C to the MPR design report for the core shroud repair (Reference 2.2.2.5) shows that the stresses in the tie rod components meet ASME code Izvel A stresses for this transient.
3.3.1.9 Meterials of Construction The tic-rod shroud repair components are fabricated from solution annealed Type 304 or 316L stainless steel with the following exceptions:
. The tie-rod itself is fabricated from ASTM A479 hot-rolled XM-19 stainless steci bar with a specified minimum yield strength of 105 ksi.
. Upper radial restraint disk springs are fabricated from Alloy X-750 and heat treated to the overaged condition specified in EPRI NP-7032.
The raw stock for the Type 304 and 316L stainless steel parts was subjected to intergranular corrosion tests per ASTM A262, Practice E. Since no welding or heat treating is performed during manufacturing, all Type 304 and 316L parts are considered to be highly resistant to stress corrosion cracking.
The hot rolled XM-19 stainless steel was selected for the tie rod due to its high strength, thermal expansion coefficient, and expected excellent stress corrosion cracking resistance in BWR service. Solution annealed XM-19 material was among the most corrosion resistant of all alloys tested by General Electric when developing nuclear grade stainless steels for BWR service TOTAL P.04
. P ge 18 of 29 No. SE-403037 001 Revision 0 (Reference 2.2.1.1.7), The expected good stress corrosion cracking resistance of the hot rolled form has been confirmed by ASTM A262, Practice E tests on the starting barstock. Further, slow strain rate tests (also known as CERT tests) in simulated BWR coolant were performed on the actual material going into service.
These tests confirmed the materials' high resistance to stress corrosion cracking in BWR service. Finally, steady state operating stresses in the tic rods have been limited to about one third yield strength, which will further reduce chances for stress corrosion cracking in the tie rod material.
Alloy X-750 was selected for the radial restraint disk springs due to its high strength and excellent corrosion resistance in BWR coolant. The spring stock was procured to ASTM B637 and heat treated to the overaged condition speci6ed in EPRI NP-7032. This heat treat condition is considered by the industry to be highly resistant to intergranular stress corrosion cracking in BWR service.
The materials as described above are equal to or better than the original materials of construction.
3.3.1.10 Potential Impact of loose Parts The various pieces that make up the tie-rod / radial restraint system are captured and restrained by appropriate locking devices. The locking device designs have been used successfully for many years in reactor internals. I.oose pieces cannot occur witho it failure.
Such locking devices and the stresses in the pieces which make up the tie-rod / radial restraint system are well within allowable limits for normal plant operation. Further, the design includes suitable features to prevent detachment of the tie-rods even if preload were lost. However, in the ,
unlikely event that a tic rod becomes detached from Ls attachment point during normal plant operation, there are no nuclear safety consequences to the shroud or to the other tie-rods. Shroud crack leakage may increase slightly if one tie-rod should fail, but such leakage would be too small to be detectable. If individual components should somehow break off the tie-rod assembly, they will l fall into the "V" shaped section at the bottom conical l
support plate or if small enough, could be transported L
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. Page 19 of 29 No. SB-403037@l Revision 0 into the recirculation loop and its pump. The i consequences of a loose tie-rod component are no different than those from other loose parts from the reactor internals within the recirculation system. Also, the NRC has determined that there is no safety concern requiring monitoring for loose parts in the reactor system j for Oyster Creek (Reference 2.2.3.13). l 3.3.1.11 Seismic Analysis The purpose of the seismic design analyses is to demonstrate that,in the event of a design basis earthquake, the shroud repair modification will 1) assure that the original shroud functional requirements are met for assumed 360", through-wall cracks in any or all of the cylindrical welds in the stainless steel portion of the shroud,2) assure that loads imparted to the reactor vessel, fuel and other affected internals are within
. acceptance criteria, and 3) demonstrate that the addition of the shroud repair components does not sigmricantly affect the seismic response of the intact (i.e., uncracked) shroud assembly, including the fuel The seismic design of the shroud repair (Reference 2.2.2.5) utilizes the pre loaded tie-rod assemblies to react vertical seismic (and other) loads and lateral seismic restraints (bumpers) to react horizontal and overturning loads, and to limit the displacement of the core at the core top guide and core support plate elevations to acceptable values. Some resistance to overturning is provided by the tie-rod assemblies. Lateral seismic restraints are also included at intermediate locations to prevent unacceptcble lateral displacement of shroud shell sections bety;een the top guide and core support plate in the event of multipfe, assumed circumferential weld failures. Other features of the seismic design include the following:
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. The lateral seismic supports transmit seismic reactions directly from the shroud to the reactor vessel wall at the elevations where the core loads are transminmt (at the top guide and core support plate) and at si i sections.
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. The lateral supports are installed in multiple circumferentiallocations to assure acceptable loads on the vessel and shroud.
. The supports are relatively rigid and provide positive restraint to limit the m L. um possible displacement to values which meet the requirements of References 1
- 2.2.1.1.5 and 2.2.1.1.6.
. Radial gaps are provided between the lateral supports and the vessel wall.' These gaps are set based on as-built measurements taken at installation and are small (about 3/8 inch) relative to the abroud diameter (15 feet). The gaps limit the seismic response of the shroud and core assembly.
The Oyster Creek shroud repair has used the seismic criteria design bases for the Oyster Creek plant.
Specifically, the sizing of shroud repair components was based on the Oyster Creek seismic accelerations specified in Reference 2.2.3.2. The licensing basis seismic accelerations are:
Description Vertical Horizontal OBE (2/3) (.24)g 0.24g SSE (2/3) (.48)g 0.48g Thus, the shroud repair is consistent with the seismic criteria used for the existing reactor vessel and its internal structures and meets the above criteria.
3.3.1.12 Meet of Raising Separator /Drver Assembly The shroud repair includes the installation of a ring which ,
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raises the separator / dryer assembly 1.5 inches. This change has been evaluated and no affects on safety were ,
identified (Reference 2.2.2.7).
3.3.1.13 Reactor Vessel Mechanical Effects ,
The tie-rod assemblies connect or contact the reactor pressure vessel at the top and bottom. The hook assembly at the bottom attachment is designed to -
maintain contact with the vessel at all times. The seismic .
bumpers will contact the vessel only under a significant ,
seismic lateralload. Evaluations were made to determine the affects on vessel stress level These evaluations are
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., .P g3 21 of 29 No. SE 403037-001 Revision 0 documented in Reference 2.2.2.5 and show that original code stress limits are met.
33.2 Quality Standards The design, manufacture and installation of the repair are classi5ed as " Nuclear Safety Related." Accordingly, Reference 2.23.12 will be applied to this activity. Quality assurance requirements are in accordance with the Design and Fabrication specifications (References 2.2.2.2 and 2.2.2.4).
333 Natural Phenomena Prutection The modiScation is confined to the inside of the reactor vessel; accordingry, the modification will not affect current provisions for tornado, hurricane or flood protection.
The modification would be loaded by a seismic event. Analyses were performed to determine the permitted seismic loads for the shroud assuming various combinations of weld cracks with the tie rod assemblies and radial supports installed. See Section 33.1.11 below for a discussion of the results of the seismic analyses.
33.4 Fire Protection The modification is confined to the inside of the reactor vessel; accordingly,10CFR50, Appendix R provisions are not affected.
3.3.5 Environmental Qualification The effects of radiation and temperature are addressed in the design. Specifically, the effects of thermally and radiation induced relaxation of the tie-rod preload stresses have been considered and taken into account. These effects are calculated to reduce the tie-rod preload by less than 5% at end oflife (Reference 2.2.2.5).
33.6 Missile Protection The modification is confined to the inside of the reactor vessel and is therefore protected from missile impingement. The modification consists of passive components and does not increase the potential for missile generation.
33.7 Internal Flooding /High Energy Line Breaks The modification is confined to the inside of the reactor vessel and is not subject to the effects ofinternal flooding. Structural /
hydraulic loading effects of main steam line break and recirculation line break were considered as design basis events in the analyses
. completed for this modification as discussed in Section 33.1 of this evaluation.
Pzge 22 of 29 No. SE-403037-001 Revision 0 33.8 Electrical Separation / Isolation /Ioadig No electrical equipment is added, suatracted or moved as a result of the moddication. Electricalloadini si in no way affected.
Accordingly, compliance with plant electrical separation, isolation, and loading design criteria are not affected by the modification.
33.9 Single Failure Criteria No active components are involved with the modification.
Accordingly, single failure design criteria are not affected.
3 3.10 Separation Criteria Requirements for physical separation of redundant equipment trains are not affected by the modification.
3 3.11 Containment Isolation Containment isolation provisions are not affected by the modification.
3 3.12 Materials Compatibility Materials of construction used in the design (Section 33.1.9) are suitable for the reactor envinnmental conditions. No materials, consumables, or lubricants are used which could have a detrimental effect on the corrosion resistance of other materials in i use in the reactor coolant environment.
3 3.13 Water Impingement Due to Water Type Fire Suppression Systems The modification is confined to within the reactor vessel and is therefore not affected by water impingement from fire suppression systems.
3 3.14 Inservice Inspection The tie-rod assemblies are designed to permit required inservice inspection of reactor vessel and its internals. A baseline visual inspection (VT-1) will be donc prior to restart.
3.4 10CFR50.59 Determination 3.4.1 De modiilcation does not significantly increase the probability of occurrence or the consequences of an accident or equipment malfunction previously evaluated in the SAR since the basic physical configuration and function of the core shroud are maintained per existing licensing basis requirements.
Page 23 of 29 No. SE 403037-001 Revision 0 3A.2 The modification does not create the possibility of an accident of a different type than previously identified in the SAR since the basic physical arrangement and function of the core shroud are maintained. Potential for loose parts was evaluated (33.1.10), flow induced vibration was evaluated (3.3.1.7), materials are as good or better than original construction (3.3.1.9), and all design loads have been evaluated and design criteria have been satisfied (3.3.1.11, i 33.1.4-6,333-3.3.13). '
3A3 The modi 5 cation does not affect the plant margins of safety as defined in the basis for plant technical specifications since no change is made to the plant operation or operating limits. The l
reactor response to normal operation, transients, and accidents has i been shown to be acceptable and within the design basis (33.1, 33.1.1,33.1.8,3.1.5,3.3.1.11).
3.4.4 The modification will not create any new potential radiation release paths; therefore, radiological safety is not affected by the modification. No changes have been made to containment-isolation function (33.11).
3A.5 Implementation of this modification will not adversely affect i nuclear safety or safe plant operations since it: 1) does not create the possibility of an accident different than previously considered,
- 2) does not affect plant margins of safety, and 3) does not create any new potential radiation release paths. (See Sections 3.4.1 through 3.4.4.)
3.5 FSAR Ch===es Section 3.9.5.1.1 of the UFSAR will be modified to add a description section similar to Section 13 of this safety evaluation and a figure similar to i Figure 2. These changes are being implemented through the appropriate GPUN UFSAR change process.
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P ge 24 of 29 No. SE-403037-001 Revision 0
4.0 CONCLUSION
This safety evaluation has determined that the addition of tic-rod assemblics to the Oyster Creek reactor vessel shroud does not significantly increase the probability of occurrence or consequences of an accident or equipment malfunction previously evaluated in the UFSAR. Further, this change will not create the possibility of an accident of a different type and will not affect plant margins of safety. No new potential radiological release paths are created.
Accordingly, it is concluded that the installation of the tie rod / radial support assemblies to permanently repair cracked circumferential welds in the Oyster Creek reactor vessel shroud does not constitute an v.aeviewed safety question.
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. Page 25 of 29 No. SE-403037-001 i Revision 0 APPENDIX A 1
FIGURES Figure 1 Oyster Creek Shroud Horizontal Welds Figure 2 Oyster Creek Shroud Repair Figure 3 Shroud Restraint Assembly Upper Bracket Figure 4- Shroud Restraint Assembly Lower Hook 1
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