ML20106D925
| ML20106D925 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/31/1992 |
| From: | Balbraith S, Beth Brown, Porter A EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20106D920 | List: |
| References | |
| EGG-MS-10141, NUDOCS 9210160106 | |
| Download: ML20106D925 (24) | |
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. .j Al l %tiMtNI 4 EGG MS-10141
. . Marct 1992 TECHNICAL EVALUATION REPORT TECHNICAL EVALUATION REPORT ON THE FIRST
/daho 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
TEXAS UTILITIES (TU) ELECTRIC COMDMY, Nat/onal - COMANCHE PEAK STEAM ELECTRIC STATI'H, UNIT 1, Eng/neer/ng DOCKET NUMBER 50-445 Laboratory ,
Managed ' 'r by the U.S.
Department ofEnergy B. W. Brown S. G. Galbraith A. M. Porter i
0 Prepamd for the 9m EGcG-U.S. NUCLEAR REGULATORY COMMISSION wasw g g<
No. DE AC07-?6lDC1bR 9210160106 921001 PDR ADOCK 05000445 P PDR
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ABSTRACT 1
This report presents the results of the evaluation of the Comanche Peak Steam Electric Station (CPSES), Ilnit 1, first 10-Year interval Inservice inspection (ISI) Program Plan, Revision 0, submitted October 15, 1990, inciuding the requests for relief from the American Society of Mechanical Engineers (ASME) l Boiler and Pressure Vessel Code Section XI requirements that the Licensee has determined to be impractical. The Comanche Peak Steam Electric Station, Unit 1, first 10 Year interval inservice inspection Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for ;
(a) compliance with the appropriate edition / addenda of Section XI, ;
(b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission ,
(NRC) review before granting an operating license. The requests for relief are evaluated-in Section 3 of this report.
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l This work was funded under:
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').S. Nuclear Regulatory Commission FIN No. D6022, Project 5
-Operating Reactor Licensing Issues Program, Review of ISI for ASME Code Class 1, 2, and 3 Components -
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SUMMARY
The Licensee, Texas Utilities (TV) Electric Company, has prepared the Comanche Peak Steam Electric Station, Unit 1, First 10-Year interval inservice inspection (ISI) Program Plan, Revision 0, to meet the requirements of the 1986 Edition of the ASME Code Section XI. The first 10-year interval began August 13, 1990 and ends Avi st 17., 2000.
The inh rmation in the Comanche Peak Steam Electric Station, Unit 1, First 10 Year interval inservice inspection Program Plan, Revision 0, submitted October 15, 1990, was reviewed. Included in the review were the requests for relief from the ASME Code Section XI requirements that the Licensee has determined to be impractical. As a result of this review, a request for additional information (RAl) was prepared describing the information and/or clarification required from the Licensee in order to complete the review. The Licensee provided the requested information in the submittal dated September 13, 1991.
Based on the review of the Comanche Peak Steam Electric Station, Unit 1, First 10-Year interval Inservice inspection Program Plan, Revision 0, the Licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations for granting relief from the ISI excminations that cannot be performed to the extent required by Section XI of the ASME Code, it is concluded that the Comanche Peak Steam Electric Station, Unit 1, First 10-Year Interval inservice Inspection Program Plan, Revision 0, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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CONTENTS ABSTRACT . ............................... 11
SUMMARY
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- 1. INTRODUCTION ............................ I EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . 4 l 2.
m 2.1 Documents Evaluated . ...................... 4 1
2.2 Compliance with Code Requirements ................ 5 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . 5 2.2.2 Acceptability of the Examination Sainple . . . . . . . . . . . 5 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . 5 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . 5 2.3 Conclusions ........................... 6
- 3. EVALUATION OF RELIEF REQUESTS . . . . . . . . . . . . . . . . . . . . 7 3.1 Class 1 Components (No relief requests) 3.2 Class 2 Components . . . . . . . . . . . . . . . . . . . . . . . . 7 3.2.1 Pressure Vessels (No relief requests) 3.2.2 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.2.2.1 Request for Relief No. C-1, Examination Category C-C, item C3.20, Integrally Welded Attachments to Containment Spray Piping . . . . . . . . . . . . . . . . .. 7 3.2.3 Pumps (No relief requests) 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests)
- 3.3 Class 3 Components (No relief requests) 3.4 Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . -9 3.4.1 Class 1 System Pressure Tests ................ 9 3.4.1.1 Request for' Relief-No. B-1, Examination Category B P. Item 15.50, Reactor Coolant Piping System Leakage lest . . . . . . . . . . . . . . . . . . . . . . . 9 iv em.., p,-w,.n,,- ,e.n,--.,,e---y e,y, n n .w . . - - . . , , y- ,,np ,w.- .-p.mg ,,e em., .,e g.e ,y y -m,.,.-n-,.r- -g +,-r,-m, --
4 EGG.HS.jo34y TECHNICAL EVALUATION REPORT ON THE FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
TEXAS LITILITIES (TU) ELECTRIC COMPANY.
CONANCHE PEAK STEAM ELECTRIC S'\ TION, UNIT 1 DOCKET NUMBER 50-445
- 8. W. Brown S. G. Gtibraith A. M. Porter Published March 1992 EG&G Idaho, Inc.-
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Division of Engineering Tcchnology Office of Nuclear Reactor Regubtion Under DOE Field Office, Idaho Contract DE-AC07-761001570 FIN No. D6022 (Project 5)
3.4.2 Class 2 System Pressure Tests (No relief requests) l 3.4.3 Class 3 System Pressure Tests .'. .............. 12 3.4.3.1 Request for Relief No. D-1, Paragraph IWD 5223, i Hydrostatic Pressure Testing of Reactor Coolant Pump (RCP) Thermal barrior Heat Exchangers .......... 12 i
3.4.4 General (No relief requests) +
3.5 General (No relief requests) !
- 4. CONCLUSION ............................. 14 ;
- 5. REFERENCES ............................. 16 t
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TECHNICAL EVALUATI')N REPORT ON THE FIRST 10 VEM INTERVAL INSERVICE INSPECTION PROGRAN PLAN:
TEXAS UTILITIES (TV) ELECTRIC COMPANY, C0r*JCHC PEAK STEAN ELECTRIC STATION, UNIT 1 DOCKET NUMBER 50-445
- 1. INTRODUCTION Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including
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supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1 Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice 9xamination requirements, set forth in the ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components" (Reference 2), to the extent practical within the limitations of design, geometry, and materials of 4 construction of the components. This section of the regulations siso requires that inservice examina', ions of components and system pressure tests conducted during the initial 120-month inspection interval comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the date of issuance of the operating license, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth [
in s'Jbsequent editions and addenda of this Code th4t are incorporated by '
reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. The Licensee, TU Electric Company, has prepared the Comanche Peak Stean Electric Station, Unit 1, First 10-Year Interval inservice inspection (ISI) Program Plan, Revision 0 (Reference 3), to meet the requirements of the 1986 Edition of the ASME Code Section XI. The first 10-year interval began August 13, 1990 and ends August 12, 2000.
As required by 10 CFR 50.55a(g)(5), if the licenee determines that certain Code examination.requiremt.nts are impractical and requests relief from them, the licensee shall submit information and justifications to the Nuclear Regulatory Commission (NRC) to support that determination.
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Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determination that Code requirements are impractical to implement. The NRC l t
may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the comon defense and security, and are otherwise in the public interest, giving due i consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. t Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the Licensee's determination that either (i) the proposed alternatives provide an acceptable level of quality and safety or (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety.
Proposed alternatives may be used when authorized by the NRC.
The information in the Comanche Peak Steam Electric Station, Unit 1, First 10-Year Interval 151 Program Plan, Revision 0, submitted October 15, 1990, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the Licensee has determined to be impractical. The review of the ISI Program Plan was performed using the Standard Raview Plans of '
NUREG 0800 (Reference 4), Section 5.2.4, " Reactor Coolant Boundary Inservice Inspections and Testing," ;ad Section 6.6, " Inservice Inspect ion of Class 2 and 3 Components."
In a ietter dated August 12, 1991 (Reference 5), the NRC requested additional information that was required in order to complete the review of the ISI Program Plan. The requested information was provided by the Licensee in the
" Response to Request for Information Related to the Inservit.e inspection Program Plan" dated September 13, 1991 (Reference 6).
As a result of telephone conversations with the Licensee on October 4, 1991, and January 23, 1992, the Licensee submitted further information in a letter dated January 24, 1992 (Reference 7), regarding the Comanche Peak Steam Electric Station, Unit 1, First 10-Year Interval Ibi Program Plan.
The Comanche Peak Steam Electric Station, Unit 1, First 10-Year interval ISI Program Plan is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of 2
Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with IS!-related commitments identified during the NRC's previous reviews. ,
i The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI, i 1986 Edition. Specific inservice test (IST) programs for pumps and valves tre being evaluated in other reports.
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- 2. EVALVATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to 151 activities. This section describes the submittals reviewed and the results of the review.
2.1 Documents Evaluated Review has been completed on the fol'0 wing submittals from the Licensee:
(a) Ccmancho Peak Steam Ele..ric Station (CPSES), Unit 1, First 10-Yuar Interval inservice inspection Program Plan, Revision 0, submitted October 15, 1990 (Reference 3).
(b) Letter, dated August 21, 1991 (Reference 8), containing interim change requests for the ISI Program Plan, a clarification regarding the weld marking system, and correction of several errors in the plan.
(c) Letter, dated September 13, 1991 (Reference 6), response to the NRC request for additional information dated August 12, 1991.
(d) Letter, dated October 16, 1991 (Reference 9), containing Request for Relief B-1.
(e) Letter, dated October 30,1991 (Reference 10), containing additional information regarding Request for Relief B-1.
(f) Letter, dated January 24, 1992 (Reference 7), containing CPSES Augmented ISI Plan, CPSES/FSAR commitment to comply with NRC Regulatory Guide 1.150, Westinghouse Technical Bulletin NSD-TB-75-1 regarding the reactor coolant pump tFarmal barrier, and ISI boundary diagrams for the comptnent cooling w.lar system.
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22 Comoliance with Code Reauirements y, 2.?.1 (omoliance witt,juolicable Code Editions The Inservice inspection Program Plan r .i be based v4 the Code L editions Nfined in 10 CFR 50.55a(g)(4) ar.d 10 CFR 50.55a(b). Based on the sis.-ting date of August 13; 1990, '.he Code applicable to the s
first 15! interval is the 1986 Editior., As stated in Section 1 of V.is report, the Licensee has prepared the Comanche Peak Steam S' F7cetric Station, Unit 1, First 10-Year ISI Program to meet the rcquirements of 1985 Edition of the Code.
. 2.E.- $_q.qqatability of the E> amination Samole t
inservice volumetric, surface, and visual examinations shall be performed on ASME Cor' ' lass 1, 2, and 3 components and their )-
supports usir>3 samply chedules describ+a 'Section XI of the ASME [
Code and 10 CFR 50.55alb). Sample ri;.c ar9 m M selection have been implemeMed in accordance w'.th the Code art 5 vfR 50.55a(b) and
. app w c correct.
2.4.3 Ey.gsg+ ion Criteria The criteria used to exempt co?ponents from ev. amination shall De consistent with Par. graphs IWB-1220, IWC-1220, IWC-1230, IWD-:220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the Licensee in accordance with the Code, as discussed in the ISI Program Plan, and appear to be correct. .
2.2.4 Auamented Examination Commitmenti In addition to the requirements as specified in Section XI of the ASME Code, the Licensee has committed to perform augmented examinations according to the following documents:
1 (a) NRC Regulatory Guide 1.14 Reactor Coolant Pump Notor flywheel Integrity, Revision 1 (Reference 11).
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(b) NUREG-0797, Saf~ety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2, Supp1emental Safety Evaluation Report 12, regarding safety injection pump shrouds (Reference 12).
(c) NRC Bu11etin 88-C?, Thimble Tube Thinning in Westinghouse Recctors, (Reference 13) and CPSES-9006199.
(d) MEB 3-1, " Postulated Pupture locations in Fluid System Piping Inside and Outside Containment", (Reference 14) and FSAR 6.C.8.
(e) NRC Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations, c, (Reference 15).
3 (y' 2.3 Conclusigni Based on the review of the documents listed above, it is concluded that the Comanche Peak Steam Electric Station, Unit 1, First 10-fea;' Interval .
ISI Program Plan, Revision 0, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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- 3. EVALUATION OF RELIEF REQUESTS The requests for relief from tha ASME Code requirements that the Licensee has determined to be impractical for_ tne first 10-year inspection interval are evaluated in the following sections.
3.1 Class 1 Comoonents (No relief requests) 3.2 Class 2 Comoonents-3.2.1 P;-essure Vessels (No relief requests) 3.2.2 Pioina 3.2.2.1 Reauest for Relief No. C-1. Examination Cateaory C-C.11gg C3.?0.
Intearally Welded Attachments to Containment Soray Pioina Code Reauirement: Section XI, Table IWC-2500-1, Examination Category C-C, Item C3.20, requires a surface examination as defined by Figure IWC-2500-5.
Licensee's Codq_,ltelief Recuest: Relief is requested from performing 100% of the Code-required surface exaT.ination on the eight welded lugs of support number CT-1-024-003-522R on containment t.pra) line number 16-CT-024-301R-2.
Licensee's Basis for reauestina relief: The Licensee states that the-eight welded lugs were examined with liquid penetrant to the maximum extent practical without. removing an adjacent welded clamp. The clamp prohibits examination of the 1/2 inch examination zone on one side of each of the lugs, licensee's Procosed Alternative Examination: None. The Code--
required surface examination was performed to the maximum extent l
L practical.
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fvaluation: Table IWC-2500-1, Examination Category C-C, item C3.20 requires a surface examination per Figure IWC-2500-5. The Licensee states that a best effort surface examination was performed on the welded lugs, but that a 1/2 inch wide portion of each lug was obstructed by an adjacent welded clamp. As shown in the drawing attached to the Licensee's relief request, the clamp obstructs a portion of the required examination trea, making the surface examination impractical to perform to the extent required by the Code. In order to perforrr the Code-required examination, the lugs would have to be redesigned and replaced. Imposition of the requirement on the Licenses would cause a burden that would not be corr.pensated by an increaw in safety above that provided by the limited examination.
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Conclusion:
Based on the abosa, it is concluded that the Code required surface examination is impractical to perfona to the full extent required by the Code at Comanche Peak, Unit 1, and that public health and safety will not be endangered by allowing the limited examination in lieu of the Code requirement.
Therefore, pu.'suant tc 10 CFR 50.55a(g)(6)(i), it is recommended that relief be granted.
3.2.3 Pumos (No relief requests) 3.2.4 -Valves (No relief requests) 3.2.5 General (No relief requcsts) 3.3 C ss 3 Comoonom (No re 'ief requests) l
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3.4 ff_essere Testi 3.4.1 Class 1 System Pressure Tests 3.4.1.1 Reouest for Relief No. B-1. Examination Cateoory B-P. Item 15.50.
Reactor Coolant Ploino System leakaoe Test Code Reouirement: Section XI, Table'IWB-2500-1,' Examination Category B-P, Item B15.50, requires a VT-2 visual examination during the System Leakage Test. As required by IWB-5221, the System Leakage Test is performed at'or above the nomir.al operating pressure associated with 100% rated reactor pressure.
Licensee's Code Relief Reouest: Relief is requested from performing the Code-required VT-2 visual examination during the System Leakage Test for reactor coolant piping between the reactcr pressure vessel (RPV) and concrete harrier wall, and within the concrete wall penetration.
Licensee's Basis for reouestino relief: The Licensee states that the reactor vessel at Comanche Peak, Unit 1, is surrounded by a concrete barrier wall. The length of the sleeve in the wall where the hotleg (reactor vessel outlet) of the main coolant loop penetrates, measures approximately 74 inches, and approximately 132 ire .es where the < 19 (raactor vessel inlet) penetrates.
The orea-where the nozzles penetrate the wall is inaccessible for direct visual examination due to the limited separation between the coolant pipe and the penetration sleeve. The separation between the insulation and the sleeve is 3 inches on the hotleg, and 2 inches on the coldleg.
The reactor nozzle areas are accessible fra the reactor cavity through access ports over each of the eighs nozzles. Entry into the nozzle areas renders approximately 30 inches of main coolant piping accessible for examination.
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Since the System Leakage! T ests are performed'at' elevated temperatures,7 access to the nozzle penetrations represents _a.
personnel hazard due to the' extreme heat within the confined y
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represent significant radiation exposure areas with an estimated-dose accumulation ~of approximately 12 man-REM.-
4 The efforts associated with the removal of.the nozzle-access covers, the personnel hazards, and the radiation dose accumulations associated with performing the direct visual examination of this limited area do not result in a corresponding increase in-quality and safety.
Licensee's Prooosed Alternative Examination: None. The' area ini which the main coolant piping exits the coi; crete barrier wall:
will be examined.for steam or other signs of leakage. "The
'aurce for'any steam present shall be examined for evidence of '
leakage,. including any residue or discoloration."- Additionally, existing plant monitoring systems provide further assurance that any significant leakaga will be detected.
Evaluation: Relief is requested from the Coda-required VT-21 g visual examination during the System Leakage Vest for'the reactor coolant pipine between the reactor vessel and the outside of the concrete barrier that surrounds the ves.;el. ,
Paragraph IWA-5241 addresses visual examinations for noninsulated components as-follows:
(a)_'The visual examination VT-2 shall be conducted by -
examining the accessible external exposed. surfaces of pressure retainir.g components for evidenceof leakage.
-(b). For components whose external surfaces are inaccessible for direct visual ' examination _VT-2, only
-the examination of ;urrounding area (including floor -
_ areas- or equipment surf aces located underneath the -
components) for evidence of leakage shall be required.
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The portion of piping that penetrates the concrete wall is in' accessible due to limited clearance (2-3 inches)-between the piping and the concrete. As stated in paragraph IWA-5241 for components that are not accessible for direct visual examination, only examination of surrounding areas for evidence of leakage is required. Since direct visual examination is not possible,-
relief is not required for the reactor coolant piping located within the concrete wall penetration, provided that the surrounding areas are examined for lenkage as required _by the Code. -
For the portion of piping between the RPV and the concrete wall, the Code required visual examination is impractical to perform due to the high radiation levels and the extreme heat generated during the System Leakage Test. In order to perform the examination to the extent required by the Code, the reactor coolant system would require extensive design modifications.
Impositfra af this Code requirement on the t.icensee would cause a burden that .cyld not be compensated by an increase in safety above that prcvided by the limited examination.
Conclusion:
It is concluded that the VT-2 visual examination is impractical to perform at Comanche Peak, Unit 1, to the extent required by the Code. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), it-is recommended that relief be granted for the portions of piping between the reactor vessel and the concrete barrier wall. For the portions of piping within the sleeve penetrating the concrete Larrier wall, it is concluded that relief is not required, provided that the surrounding areas are e 1 mined for evidence of leakage.
3.4.2 Class 2 System Pressure Testl (No relief requests) i 11
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3.4.3 Class 3 Svitem Pressure Tasis _
3.4.3.1 Recuest for Relief No. 0-1. Paraaraoh IWD-5223. Hydrostatic Pressure Testina of Reactor Coolant Pumo (RCP)- Themal Barrier Heat Exchanaers Code Reaui ument: Section XI, paragraph IWD-5223(a), requires a system Fydrustatic test pressure of at least 1.25 times the system pressure (P,,) for systeias with design temperature above 200'F.
Licensee's Code Relief Reauest: Relief is requisted from the Code-required hydrostatic test pressures of paragraph IWD-5223(a) for reactor coolant pump (RCP) thermal barrier heat exchangers, licensee's Basis for reauestina relief: The Licensee states that the Pc , for the portion of the component cooling water system containing the RCP thermal barrier heat exchangers is 2485 psi with a design temperature of 650*F. This yields a test pressure of 3106.25 psi. To attempt to pressurize the tube' side of the RCP thermal barrier heat exchangers would potentially damage these componants. The manufacturer (Westinghouse) has issued a technical bulletin advising that the maximuu allowable field _
hydrostatic pressure is 225 psi for the component cooling water side oir these RCP thermal barrier heat exchangers. These heat exchangers are desig'ned for high differential pressures in the direction opposite from that imposed by a component cooling water hydrostatic test (i.e. from the reactor coolant system). To rneet the conditions that could exist in the event of a heat exchanger leak inside of the RCP, the external cont.ections and adjacent piping are designed for 2500 psi internal pr$isure. 3 Licensee's Prooosed Alternative Examination: The portion of the component cooling water system that constitutes the tube side of the RCP thermal barrier heat exchangers shall be hydrostatically tested along with the portions other than those sections designed 12 l
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for 2500 psi. The design pressure of these portions is 150 psi and the test pressure shall be the required 1.25 times P,,.
Evaluation: Paragraph IWD-5223(a) requires test pressures of
- 1. 2 5 - P,y for systems with de:ign temperatures above 200*F. The P,, for the portion containing the RCP thermal barrier heat exchangers is 2485 psi with a design temperature.of 650'F, yielding a test pressare of 3106.25 psi. Westinghouse Technical' Bulletin NSD-TB-75-1 (included in Reference 7) advises that the maximum internal hydrostatic test pressure for the thermal barrier heat exchanger is 225 psi The Liccasve's proposed alternative is to perform the hydrostatic pressure tests of the subject heat exchangers at 1.25 times P,, (which has not- been speci fied) .
Pressurizing the tube side of the RCP thermal barrier heat exchanger above the 225 psi maximum could potentially damage these components. Therefore, the Code requirement is impractical. In order to perform the examination to the-extent required by the Code, the RCP would require extensive' design modifications. Imposition of this Code requirement on the Licensee would cause a burden that would not be compensated by an increase in safety above that provided by the Licensee's proposed alternative.
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Conclusion:
It is concluded that hydrostatic pressure test is impractical to perform at Comanche Peak, Unit 1, at the test pressure required by the Code. Therefore, pursuant to 10 CFR 50.55a(g)(6)(1), it is recommended that relief be granted.
3.4.4 General (No relief-requests) 3.5 . General (No relief requests) l 13
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- 4. CONCLUSION ,
Pursuant to 10 CFR 50.55a(g)(6), it has been determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code. For the Relief Requests C-1 and 0-1, the Licensee has demonstrated that specific Section XI requirements are impractical. For Request for Relief B-1, it is concluded that relief may be granted in part, and that relief is not required for the remainder of the request.
This technical evaluation has not identified any practical method by which the Licensee can meet all the specific inservice inspection requirements of Section XI of the ASME Code for the existing Comanche Peak Steam Electric Station, Unit 1, facility. Compliance with all the exact Section XI required inspections would necessitate redesign of a significant number of plant systems, sufficient replacement components to be obtained, installation of the new components, and a baseline examination of these amponents. Even after the redesign efforts, complete compliance with the Section XI examination equirements probably :ould not be achieved. Therefore, it is concluded that la oublic interest is not served by imposing certain provisions of Section XI te '1e ASME Code that have been determ:ned to be impractical. Pursuant to 10 CFR 50.55a(g)(5), relief is allowed from the requirements that are impractical to implemen!.. Relief may be granted only if the relief will not endanger life, property, or the common defense and security and is otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The Licensee should continue to monitor the development of new or improved examination techniques. As improvements in-these areas are achieved, the
- . Licensee should incorporate these techniques into the ISI program plan
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examination requirements.
l Based on the reviev cf the Comanche Peak Steam Electric Station, Unit 1, First 10-Year Interval in::crvice Inspection Program Plan, Revision 0, the Licensee's ,
response to the NRC's Request for Additional Information, and the recommendations for granting relief from the ISI examination requirements that have been determined to be impractical, it is concluded that the Comanche Peak :
Steam Electric Station, Un!t 1, First 10-Year Irterval inservice Inspection 14 i
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Program Plan, Revision 0, is acceptable and in compliance with 10 CFO 50.55a(g)(4).
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- 5. REFERENCES
- 1. Code of Federal Regulations, Title 10, Part 50.
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1:
1935 Edition :
- 3. Comanche Peak Steam Electric 3tation, Unit 1, First 10-Year Interval Inservice Inspection Program Plan, Revision 0, dated Octi ar 15, 1990.
- 4. HURE6 C800, Standard Retrist Olan for the Revist of Safety Analysis Reports for Nuclear Power Plants, Section 5.2.4, " Reactor Coolant Boundary Inservice Inspection and Testing," and Section 6.6, " Inservice Inspecticn of Class 2 and 3 Components," July 1981.
- 5. Letter, dated August 12, 1991, T. A. Bergman (NRC) to W. J. Cahill Jr.
(TV Electric Company), containing request for additional information on
-the First 10-Year Interval ISI Program Plan.
- 6. Letter, dated September 13, 1991, W. H. Cahill (TV Electric Company) to Document Control Desk (NRC), centaining response to NRC request for additional information.
- 7. Letter, dated January 24, 1992, W. H. Cahill (TV Electric Company) to Document Control Desk (NRC), containing additional information regarding the First 10-Year Interval ISI Program Plan.
- 8. Submittal, dated August 21, 1991, Tu Electric to Document Control Desk (NRC), containing interim change requests for the ISI Program Plan, e clarification regarding the weld marking system, and correction of several-errors in the plan.
- 9. Letter, dated October 16, 1991, W. J. Cahill (TV Electric Company) to Document Control Desk (NRC), containing Relief Request B-1.
- 10. Letter dated October 30, 1991, W. J. Cahill (TV Electric Company) to Document Cantrol Desk (NRC), containing additional information regarding Relief Request B-1.
- 11. NRC Ragulatory Guide 1.14, Reactor Coolant Pump Motor Flywheel Integrity, Revision 1, August 1975.
i 12. NUkEG-0797, Safety Evaluation Report R11ated to the 00eration of Comanche Peak Steam Electric Station, Units 1 and 2, Suppiementa1 Safety Evaluation Report 12, July 31, 1981.
- 13. NRC Bp ,etin 88-09, Thimble Tube Thinning in Westinghouse Reactors, l
July h 1988.
i
- 14. Generic Letter 87-11, dated June 19, 1987, containing Branch Technical Position MEB 3-1, " Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment," Revision 2, June 1987.
16 1
,- - , m - -
1S. NRC' Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preser':fce and Inservice Examinations, Revision 1, Februarj 1983.
s e
17
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. nras u.o s.ar:rLa EGG-MS-10141' Technical Evaluation Report on the First 10-Year can unc=T ev ius-t:
Interval Inservice Inspection Program Plan: a-
-- *m Texas Utilities E~iectric Company l _
.omanche Peak Steam Electric Station, Unit 1 . March 1992- -
- *m oa caar vsta Docket Number 50-945 FIN-06022 iProj. 5) 6 TYPE oF REPC AT
- 5. Au recaf 51 .
Technical L eta ic h a u o "-- a-B.W. Brown, S.G. Galbraith, A.M. Porter 1 -
o m or, .,o u.2 ~.. ,, c .
agegaggApe4rioN - Niue aNo acoa ass .,, ~,c.
EG&G Idaho, Inc.
P.O. Box 1625 Idaho Falls, 10 83415-2209
.r.w anc o.- o,,er nw. y 1 m- a r. c,,
R WG QMGMQATION . NAME ANo AcoR E55 n,*ac. syme w m.m '. re. .
- 9. sPC.NSO9 W *.sh gr 44 Materials and Cher. :a1 Engineering Branch Offica of Nuclear .segulatory Commission U.S. Nu. clear Regulatory Conission
- o Washington, D.C. 20555
- 10. $UPPLEuiENT AR Y NQYES ,
- 11. AasTR ACT (Joo u. .r ==,
Thir teport presents the results of the evaluation of the Comanche Peak Steam-Electric Station (CPSES), Unit 1, First 10-Year Interval Inservice Inspection (ISI)
Program Pian. Revision 0, submitted October 15, 1990, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure ,
Vessel Code Section XI requirements that the Licensee has determined to be impractical. The Comanct.e Peak Steam Electric Station, Unit 1, First 10-Year Interval Inservice Inspection Pr mram Plan is evaluated iu Section 2 of this report.
The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptabil,ity of examination sample, (c) ccrrectness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-re ated l commitments 'dentified during the Nuclaar Regulatory Commission (NRC) review before granting an operating license.
The reauests for relief are evaluated in Section 3 of this report.
a u v woaos,oesca.Prcas s ,- , C ,, ~ ,, ,. ,u a - ~ .~ n u u ..~r Unlimited
,aa .. n c-.. .w. -
aw Unclassified
, no . ,
Unclassified is.uvuesa c" acts
- 18. P a lCV mhcpCa43M 449
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