ML20101A073
| ML20101A073 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/15/1992 |
| From: | Starkey R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS-92-118, NUDOCS 9204200113 | |
| Download: ML20101A073 (25) | |
Text
Carolina Power & Light Company Nuclear Services Depadment
- I 411 Fayetteville Strett Mall-P.O. Ilm 1551 Raleigh, North Carolina 27602 April 15,1992 SERIAL: NLS 92118 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR 71 & DPR-62 MASONRY BLOCK WALLS Gentlemen:
The purpose of this letter is to respond to the NRC Staff letter dated April 9,1992 concerning masonry block walls at the Brunswick Steam Electric Plant, Units 1 and 2. The four NRC questions, along with Carolina Power & Light Company's respo ises, are provided in Enclosure 1 of this letter. Tha NRC letter asks for a response the week of Apri! 13,1992. The responses in the attachment regresent our best efforts to respond in a short time period and, in some cases, contain information that is preliminary in nature and subject to change with continuing review.
Please refer any questions regarding this submittal to Mr. D, C. McCarthy at (919) 546-6901.
T ours very truly,
(, fl h
R. B. Starkey, Jr.
WRM/wrm (mwall.wpf)
Enclosure
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Mr. S. D. Ebneter Mr.N.B.Le Mr. R. L. Prevatte Vl
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s ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR 71 & DPR 62 MASONRY BLOCK WALLS NRC OUESTION 1:
Describe your corrective actions and evaluation criteria used in determining the acceptability of the diesel generator building masonry walls.
CP&L RESPQ_NJg:
An investigation was performed to verify the integrity of the individual bolts used in the masonry block walls defining the boundaries of the diesel generator cells. A 100 percent inspection was performed on the bolts used at the perimetor of the Diesel Generator Building walls, and as-built bolt location sket':hes were generated. A 24 percent inspection sampling was performed on the through bolts supporting the missile shields. The percentage of through bolts with idemified deficiencies was not considered significant to the total number of bolts necessary to ensure the missile shield and wall act as a composite section. Therefore, the identified through bolt deficiencies are not considered significant to the seismic analysis.
The as-built perimeter botting sketches were reviewed and ranked according to the apparent safety margins. A detailed finito elemental analysis was performed on the worst case configuration and resultant bolt reactions compared to an assumed worst case anchor installation. This analysis determined that one masonry wall lacked sufficient integrity to withstand the design basis earthquake (wall 8, south wall of Diesel Generator #4 cell). This condition required declaring Diesel Generator #4 inoperable and placing the plant in a 7 day Limiting Condition for Operation (LCO).
Corrective action to restore full design basis requirements for this masonry wall involved the installation of 58 new anchor bolts. This action was completed on April 12,1992 and Diesel Generator #4 was declared operable.
The next three most critical walls were analyzed simultaneously using finite elemental techniques.
Although the installation does not conform to the licensing basis structural requirements, the walls were determined to have sufficient structural integrity to withstand the design basis earthquake and are operable. The remaining walls forming the boundary of the Diesel Generator cells all had bolting deficiencies that were bounded by the analyses performed. The schedule for permanently repairing these walls is discussed in response to Question 2.
The evaluation criteria used in all cases was seismic ctatic acceleration using floor response spectra for the 50 foot elevation. Block wall stresses and deflections were extremely small
(~ 1 ksi, <.02 inches). The wall design was limited by anchor bolt capacity only, which meets a factor of safety of three (3). The details of the evaluation criteria are contained in Nuclear Engineering Department (NED) Design Guide 11.20, " Civil / Structural Operability Reviews" and is included as Attachment 1 to this letter On April 9,1992, Dr. Ma from the NRC visited the Brunswick Plant and reviewed NED Design Guide 11.20. He indicated that his review determined that NED Design Guide 11.20 is satisf actory for evaluating the operability of structures.
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NRC QUESTION 2:
Describe the CP&L plan and schedule for evaluating the status of other walls and equipment to assure compliance with the licensing basis, if criteria different from those in the design basis are used to determine the acceptability of masonry walls or equipment for service, please provide those criteria.
C.P&L RESPONSE:
Masonry Block Walls: Carolina Power & Light Company plans to perform a 100 perceat inspection of bolts installed during original construction in all remaining masonry block walls containing anchor bolts in the Diesel Generator Building as a result of the deficiencies identified to date. Inspection of the Diesel Generator Guilding walls is planned to be complete by April 21,1992.
An inspection of 25 percent of anchor bolts in masonry block walls in other buildings containing safaty-related equipment will also be performed. This sampling will be expanded depending on the scope and nature of any identified deficiencies. Inspections of these walls are planned to be complete by May 5,1992. provides a listing of each wall remaining to be inspected along with a schedule for accomplishing the inspection.- If any walls are determined to have bolting deficiencies, they will be evaluated in accordance with NED Design Guide 11.20. Any wah failing this criteria will be declared inoperable and appropriate Technical Specif' cations LCOs for affected equipment will be entered.
Any wall requiring permanent repair to reatore the licensing basis structural requirements will be completed at the earliest opportunity consistent with operational and ALARA concerns. Every reasonable effort will be made to complete permanent repairs during the Unit 2 maintenance outage scheduled to begin April 30,1992 and the Unit 1 surveillance outage scheduled to begin no later than June 5,1992. Permanent repairs will be made no later than startup from the next refueling outage for the affected unit.
l Qther An bor Bolt Acolications: An assessment has been performed of areas other than masonry block walls where expansion anchors have bee used. Four areas have been identified as follows:
A. ' fjpino Suncorts: Anchor bolts used in piping supports were previously evaluated in response to IE BuP.otin 79 02 and supplements. Because of the deficiencios identified in masonry block wall anchors, CP&L management directed that an audit of IE Bulletin 79 02 be conducted to determine if fraudulent installation of piping support anchos bolts had also occurred that warranted additional anchor bolt inspections. The audit was conducted by three anchor bolt specialists from the Harris Plant and two specialists from the Brunswick l
Plant. The audit determined that sufficient inspection of pipin0 support anchor boits had L
been performed to reasonably identify improper installation practices. The audit concluded that no further inspections were warranted. The scope of the audit ar'd specific results are summarized in the following paragraphs:
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"Insoection and Testinn Procedure for Concrete Exoansion Anchors' Rev. 4.
Special Test Procedure 79 022 was written as a procedure for testing concrete expansion anchors in olar:e to ensure that the installed anchors employed for selected oioina systems meet or exceed the anchor's design strength.
The audit team vigorously reviewed the procedure to ensure that the technical requirements of IE Bulletin 79-02 were adequate!y addressed and that test methods used were appropriate to meet the technical requirements and identify improper bolt El-2
installations. Emphasis was placed on the inspection of the anchor's physical design characteristics such as anchor embedment la concrete, anchor diameter, anchor torquing, and bolt thread engagement in the anchor sleeve. The prncedure adequately covered all the important features of the anchor design and installation. Methodt of measuring and testing these features (i.e., measuring rod thread engagement in sleeve, hydraulic ram tension, and testing torquing of the nut) were also acceptable. Finally, the recording of this anchor feature data was accomplished via use of data sheers. These data sheett properly recorded pertinent information for later engineering review and appaval and are retained in the plant vault.
The audit team concluded that the procedure effectively conveyed appropriate instructions for the field inspectors to follow to adequately test and record anchor data.
The team noted that the sampling technique and frequency were in compliance with the requirements of IE Bulletin 79 02.
- 2. Qgigrmine if SP 79 22 was orocerly imnlementesLin the fieldjurina the IF ljulletin 79 02 nine hanner expansigr' anchor walkdown insuectioD grid if retylluyftrft cronerly cocumented End stored.
After inspection and test data from SP 79 22 was evaluated, the informPtion was cent to the piant vault for storage. The audit team reviewed, at random, mar:y data sheets of-inspection packages retained in the plant vault. The packsges were eranged by system, as mandated in IE Bulletin 79 02.
As each test anchor was inspected foi attributes specified in GP 79 22. the held inspector noted on cata sheets the anchor rod or bolt thread eagegement, the bolt diameter, type of anchor, anchor length, type of fastener, sad tett torque. The inspector then determined if the anchor passed all of the aucsptar,A criteria. If not, -
then the anchor was failed. All anchors were then reinsta!!eJ md te'arqued and attachments regrouted as needed. Attachment reinctshation was documented. Failed
-anchors were either replaced or determined to be ce'tshle es-is by analysis.
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- The audit teams review of inspection packages retain;d in the plant vault determined that field inspection and documentation wert a acordanc~ with SP 79-22.
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- 3. Investiaate the inordinatelv larae_ number 6.fjig!Edr!llina exoansion anchors which were not tested under orocedure SP 79 2Lduq1g.g " frozen" levelina nut or due to a " frozen" I
stud and determingif these anchoraces renresent an unsatisfactory condition or l _.
compromise the desian _ integrity.
l' As the test anchors specifLd in SP 79-22 were being examined, inspectors soon found-that a large number of anchors could not be completely examined because of a " frozen" In effect, either the bolt stud or leveling nut was seized, stud or " frozen" levelir 1 ne a.
corroded, or otherwise bemg held tightly such that they could not be rotated.
Note: The majorits af these installations involved grout attachments.
Most attach'1ents. employed the use of self-drilling anchors and had a threaded rod to clamn the attachment to the concrete surface. If the concrete was uneven, that l-attac.iment was leveled by use of leveling nuts under the attachment. The area under the attachment was then grouted. Other attachments did not require leveling so in those cases the attachment was fastened r*ectly to the concrete surface and no leveling nuts or grout was needed.
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i 19 9-During the irispection process under SP 79-22, the inspector was required to unscrew the threaded rod (or hex head bolt) from the anchor and determine the amount of thread engagement of the rod into the anchor, if the attachment was surface or flush mounted, then it was simply a matter of unscrewing the rod or bolt. If leveling nuts wem empleryod, then the grout was chipped from under the attachment to break loose the leveling not and bolt or rod from the grout. The levelinD nut was then backed off from the unduside of the attachment and the rod was unscrewed from the self-drilled anchor in the concrete.
Out of a total 433 anchors that were esamined,156 anchors could not be fully evaluated because the stud (rod, bolt) or leveling nut was, for unknown reason, "freron" or seized. The audit team went through plant vault records and examined each inspection report for all 156 anchers. The following represents a breakdown _of
' the v-arbus causes of the frozen leveling nuts and studs:
EELQZEN STHQ3 A total of 115 frozen studs were reported:
1176 These frozen studs were grouted placements that had the grout chipped away and the. leveling nut mji backed off. There was no indication that anything other than grout in the anchor or corrosion was the causo of seizure. No indication of fraudulent installation was evident.
- 2) 15. These frozen studs _were grouted placements that had the muut chipped away but the inspector made no mention in his comment section that the leveling nuts had been backed off. By procedure these nuts were required to be backed off and there was no necessity to reiterate in the comment section that the nuts
- had indeed been backed off. The only time a comment would be mandated was l
when the leveling nut could not be backed off, in this case, the absence of comment holds the presupposition that the leveling nuts were backed off. With l
this being the case, there is no reason to believe that anything other than grout in the anchor or corrosion was the cause of the seizure. No indication of fraudulent installation was evident.
l 3114 These frozen studs (10 threaded rods and 4 hex head bolts) were surface or flush mounted attachments, No grout was placed beneath these attachments and no leveling nuts were used. The assumption in these cases is indeterminate as there was no way to be sure what the cause of the seizure was.
- 4) 6 Theso frozen studs fell under the miscellaneous category (i,e., studs were bent, interferences existing such that the studs did not have clearance to be pulled out, or no threads available above the r:ut for double nutting in order to turn the l
studs). The assumption in these cases is indeterminate as there was no opportunity to actually examine under the attachment.
- 5) 4 These four were actually cast-in-place anchor bolts and were incorrectly included in the test sample. These are not expansion anchors and by design will be "irozen" in place as they were placed in concrete.
FROZEN LEVELING NUTS A total of 41 leveling nuts were reported to be frozen in place. The audit teau investigated each one of these occurrences snd the results are summarized below:
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1)26 These leveling nuts were in the Service Water System and were frozen due to corrosion. They were unable to be tested and were subsequently replaced under Plant Modification 79124. There was no evidence of any other cause of the seizure other than corrosion.
2)15 After the grout was chipped away, these !aveling outs were found to be frozen in place. The assumption in these cases is indeterminate as the inspector did not mr.ntien if the nuts were corroded.
CONCLUSIOft
'The audit team concluded that out of a total of 156 frozen leveling nuts and frozen studs, only 35 entis could not be rationally and indisputably explairied. These 35 ca.es must be bbelled as indeterminate in cause. There was no indication of inappropriate installation it should be further noted that of all the records the audit team examinN, there was not a single instance of inspector comment regarding tack welds or fa4ification of the installation, in the absence of other implicating evidence, the audit team must conclude that these 35 unexplained cases of seizure are most likely the result of grout in the anchor sleeve or corrosion.
The results of the audit are consistent with the overallinsp]ction results associated with IE Bulletin 79-02; only 2.5 percent of the anchor tests for Unit i and 3.6 percent of the anchor tests for Unit 2 failed. These results indicate that anchors bolts for piping systems were properly installed with few exceptions and are not indicative of the problems encountered with masonry block wall anchor bolts.
B.
Raceway suonorts Fraudulent expansion anchor installation in raceway components is not considered an issue based on past experience with work associated with raceways. Portions of raceway have
- been disassembled / removed / replaced due. corrosion, interference removal, and plant modifications. Fraudulent anchor ir:stallation lias not been observed as a problem.
To validate these observations, a sampling plan and a schedule for physically examining raceway supports will be developed by May 1,1992 and communicated to the NRC.
C. Buildina Steel Fraudulent installation of bolt anchors for buildity, steelis not considered an issue based on previous evaluations of building steel bolting due to activities such as plant modifications, corrosion replacements, interference removal / reinstallation and other bolt evaluations.
Instances of fraudulent installations have not been reported.
To validate these observations, a sampling plan and a schedule for physically examining building steel supports will be developed by May 1,1992 and communicated to thu NRC.
D. gauinment Foundations Fraudulent installation of bolt anchors for equipment foundations is not considered an issue for the following reasons:
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- 1. The normalinstallation practice associated with foundations for large' pieces of equipmont employs embedded anchors.
- 2. For small equipment, random simples of installations due to equipment decommissioning, corrosion refurbishment, and plant modification work has not resulted in any fraudulent installations being reported.
A review of representative OC records for safety related foundation supports will be conducted to verify that inspection activities reviewed the adequacy of anchors for foundation supports. This review will be completed by May 15,1992, if OC records cannot validate the proper installation of foundation anchors, a plan and a schedule for physically examining foundation anchors will be prepared by June 1,-1992 and communicated to the NRC NRC OUESTIOR_3:
Describe your jun..fication for ccntinued operation while you conduct your evaluation of the remainin0 masonry walls.
CP&L RESPONSE:
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Nonconformino Co.n_djhon During NRC Inspection 92-10, CP&L determined that the inspection technique used to ascertain which bolts were installed properly in the Diesel Generator BuildMg masonry walls was not effectiv9. The technique of using a shim to determine the effect.seness of the installation led to non conservative conclusions on the adequacy of several bolts. As a result, the engineeritig analyses previously performed assessing the operability of the walls was invalid, end the ability of the walls to maintain their functional capability was questioned.
The non-conforming walls had been evaluated for des;gn adequacy in response to IE Bulletin 80-11, " Masonry Wall Design", A total of 85 walls were required to be reviewed under IE.
Bulletin 80-11. Sixty-five of the 85 walls were evaluated as acceptable as-is, and were not modified as a result of the Bulletin review.
- 11. Justification for Ccalinuei Operation Adequate design margin is generally demonstrated in recent evaluations of the deficient botting installation in the Diesel Generator Building walls. Of the 12 walls which have completed
- inspection to date, all but one (Diesel Generator Building wall 8) were ablo tc be qualifi9d for structural integrity. Diesel Generator Building wall 8, which had a significant nurnber of improperly installed bolts, coud not be shert-term qualified and was permanen*1y repaired by installing proper anchor bolts. Therefore, the evidence to date would indicate that, although construction was not in accordance with design and did not rneet the licensing bases, sufficient design margin did exist to provide structural integrity with one exception.
While no absolute statement can be made that no other inadequate walls exist the likelihood of a wall having to sustain a design basis earthquake until remaining walls can be examined is acceptably low. All remaining Diesel Generator Building masonry block walls containing anchor bolts will be examined by April 21,1992. Based on current EPRI seismic curves, the probability of a desien basis earthquake in the interirn period until April 21,1992 is 4.6E-6. All other masonsy block watts containing anchor bolts will be examined as described in response to question 2 by May 5,1992. The probability of a desian basis earthquake in tha interim period -
l until May 5,1992 is 1.4E-5.
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Therefore, CP&!, believes that sufficient justification exists to demonstrate that there is negligible increated risk to the health and safety of the public associated with continued J
operation of the Brunswick Plant, Urits I and ? while evaluations continue on safety-related j
masonry walls.
1 NRC QUESIlQN.,4.:
Describe the root causes of the identified deficiencies, e.g., weaknesses in contractor oversight and weaknesses in quality assurance and quality control. Also, explain why CP&L failed to identify and correct these deficiencies when the corrective actions for Bu:letin 80-11 were implemented.
.QG1JiESPONSE:
ROOT CAUSE:
Information recently gathered from Diesel Generator Building Civil drawings (in particular F-1929) and recent telephone conversations with previous employees of Brown &
Root and United Engineers & Constructors indicate that missile protection plate and winforcing a >gles were installed on the east / west walls on the 23 foot elevation in 1973 as original construction. Consistent with construction practices of the early 1970's, construction of these masonry walls was considered seismic but non safety related and, as such, would not have required inspection installation documentation. As a result, no turnover documentation, Iristallation recorde, or audit / inspection reports were fcund during approximately 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of
. document searches conducted. Turnover information was found for work completed in the Diesel Generator Buildings by the Mechanical Services, Electrical, and Instrumentation organizations within Brown & Hoot's work force. Documentation was also available from Civil groups on concrete pours, but again, nothing concerning the instaliation of masonry walls was l _'
found. It is apparent from our raview that OC documentation does not exist for masonry block walls.
The following were reviewed and form the basis of the above statements:
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- 1. Computer Key Word Searches of Plant Records on the Following:
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Missile, Missile Protection. DG, CGB, Diesel Generator, Diesel Generater Building, Masonry l'
Walls, Walls, Engineering Change Packages /DG/DGS, Field Engivoering Change Packagec/DG/DGB, Brown & Root to United Engineers & Constructors Field Reports.
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- 2. Manual Sea.ches of Document Inuicas:
Engineering Change Package, Field Engineering Chango Packages, Brown & Root to United q
Eng,aeers & Constructors Field Reports, Turnovers by system, Quality Assurance inspection
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- 3. Brown and Root Quality Arsurance Procedures
- 4. Diessi Generator Buiding Turnover Packagds
- 5. As-built Drawing Turnover Letters from United Engineers & Constructors El-7 I ".
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- 6. Brunswick Civi! Drawings.
F 1926 F 1927 -
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F-1928 -
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F 1301 D-1645, sheets 1, 2, & 3 F-1643 F-1663 F '1604 F 1665 FSC 533
- 7. Final Safety Analysis Report
- 8. Specification for Masonry and Caulking 9527 01-29 1
- 9. Safety Evaluation Report Brunswick 1 and 2, November 1973 ILBULLETIN 8011:
1he focus of the IE Bulletin 80-11 was the desian adequacy of masonry block walls to support safety related attached components and to remain intact for postulated design basis loads. An engineering revinw determined that the walls would have structuralintegrity assuming the installation was as designed. Some walls did not meet code allowable stresses and required modification. A field walkdown was performed to visually verify that desian features, such as
. cover brackets, throuDh bolts, and steel plates, existed. CP&L did not question the adequacy of installation of design features, only whether the appropriate design features were present. Further, IE Bulletin 8011 did not require anchor bolt testing or verification as was required by IE Bulletin 79 02 for pipe supports. Therefore, the scope of IE Bulletin 80-11 would not have resulted in the identification of deficient or fraudulent botting installation. Carolina Power & Light Company inspections, evaluations, and repairs of masonry block walls were in accordance with the scope of IE Bulletin 8011, as indicated by our letters dsted July 7,1980, November 5,1980, November 25,1980, December 9,1980, July 29,1983 and April 27,1984.
AQplTIONAL NRC REQUIST I
. On April 13,1992, the NRC verbally requested that this response include a discussion of events occurring snce the initial identification of bolting issues associated with Diesel Generator Building masonry block walls.. Due to the lateness of this request, there was insufficient time to preparo and validate this information, This item will be discussed in response to a potential violation resulting from NRC inspection 9210.
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ATTACHMENT 1 DESIGN GUIDE NUMBER DG-il.20 DESIGN GUIDE FOR CIVIL / STRUCTURAL OPERABILITY REVIEWS i
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CAhDLINA POWER & LIC!lT COMi'ANY NUCLEAR ENGINEERING PEPARTMENT DESIGN CUIDE FOR CIVIL / STRUCTURAL OPERABILITY REVIEWS 1
DESIGN CUIDE NUMBER DG II.20 i
M @k PPR0iED REVISION SURMITTED
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Civil /Strue. Oper, Review a
I t.IST OF EFFECTIVE PACES i
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11.20.
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11.20*3 2
11.20*4 1
11.20 5 1
11.20 6-0 II.20 7 0
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Desi n Guide 11.20 Civi f/St rue. Ope r. Reviews TABLE OF CONTENTS EAgt EL.
I.
INTRODUCTION A.
furpose 1
B.
Applicability 1
II.
CENE RAL A.
References 1
B.
Responsibilities 1
C.
General Design criteria 2
III.
PRACTICE A.
Acceptability 2
B.
Operability 2
C.
Reportability 5
IV.
Attachments A.
Flow Chart Procedure to Evaluate
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Opershility of Systems Other than Piping B.
Operabality Review Cover Sheet 8
C.
Industry Criteria Comparison 9, 10 Rev. 1 Pace 11.20 i1 (10330cS2.ctw/cro
l Desirn Guide 11.20 C4vil/Struc. Oper. 1: e v i e ws I.
INTi.0Dl!CTION A.
l'u t po s e The purpcse of this
,;u i d e t r.
to tr.tablish technical critcria to te fol' C4vil Discipline persennel when pe r f e rn.i n g ope r a bi..
- _ po rt a bi li t y i evi ews f o r 11NP, LNP, and KNP.
'lhese teviews/;aalvses relate st rictiv to the st ruc t ural re' lated to post Scismic aspects of structure /corponen'/ptptng as operability inclusive of all other loading conditions, 11.
A;plicability This guidance is applicatle to all Cavtl Discipline personnel (direct and contract) involved in nuclear plant structur</ccmponent,' piping cperability revtews.
Deviations from tt.is design guide shall l'e with the approval of the Principal Engineer anly.
It is the responsibility of the Responsible Engineer to inform the Project and Principal Engineers of deviations e x i t, t i n g, in any sutmitted calculation.
Deviation approval shall be deeme.d the 4.pproval of the calculaticin by the Principal Engineer.
In adOtion, any deviatier. wh;ch occurs consistently shall te t>rour,ht to the attention of the Discipline Manager ior resolutton.
II.
CLNERAL A.
References 1.
NED Guideline E 25 2.
10CTR50.72 3.
ENP 01 0'., O! 4.1 4
RNF Memerandum PNPD/F9 3551. 10/25/59 (Contained in E 28)
B.
Responsibilities Lead Engineer Perforn appropriate evaluations as di ected by the Principal Engineer.
Principal Engineer (Lead Sectioni
'nture evaluation is performed in accordance with plant guidance and this document.
Discioline Manager (Civil)
Canarrence with operability /reportability
><aluations.
Assists in informing plant af conditions per E 28..
Rev 0 f ro I I 20 1 (HESS-1Qebi
Design Guide II.20 Civil /St ruc. Oper. Reviews e
C.
General Design Criteria Issues which are identified either by plant personnel or internally through the design process may require operability review if the condition is considered to deviate f rom the analysed design basis. A determination whether operability review is required per 10CFR$0.72 will be the joint decision of plant management and NED and should consider suen f actors as:
Plant condition at the time the issue 16 found.
Whether the issue le covered by other Tech Spec contingencies.
Required condition of the system in question.
When notified that an operability review is required, the time frame in which the review must be done and the notification pv4'ess shall be per E 28 ( i.. e., administrative activities).
This desig.. gaide establishes technical criteria to be followed in the course of the evaluation consistent with requirements specified in NRC Generic Letter 91 18.
III.
PRACTICE The evaluation to de t e rmine the status of an existing civil field condition which does not comply with design will consist of three stages:
acceptability, operability, and reportability.
A.
Acceptability Once the issue is defined and the review has begun, the first cut to be reviewed is shether or not the condition is acceptable "as is' with no modification. This is defined as meeting:
All applicabic code allowable stresses (AISC, ACI, AWS, Fiping Codes).
l Expansion anchor Factor of Saf ety of 4 or $ as required.
j Use of committed dampint ratio.
r FSAR and technical specification connitments.
Utilizing accepted industry practice for _salysis.
If the cendation meets the acceptability criteria, the analysis may be stopped, documented by standard calcula tioi, and the plant verbally notified.
If the condition does not meet acceptability criceria, based on the issue and upon management concurrence, generate a Desi n Deficiency E
Report per 3.18 and proceed to the operability review.
B.
Operability Civil / structural operability is defined as the ability of a structure / component to remain elastic and pe rf o rm its design f unc t it.n without permanent de f o rmation or detrimental effect on adjacent safety related components / structures, i
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Design Guide 11.20 Civil /St rue. Ope r. Reviews This evaluation may be accomplished in either of two ways:
i Specific analysis or testing using acceptance limits established in this guideline. This method may be used in all cases.
A review and a
an established ' Senior Structural Panel.' pproval of Review This method may only be used with the concurrence of the Civil Discipline Hanager or designee.
1.
Speelfic Analvnig:
The component / structure in question may be considered operable based on specific computer or hand calculations provided the following conditions are satisfied:
The following material stress limits are not exceeded Tenslie Stress -. 9 F,*
- Special consideration required for pin cornected menbers and threaded parts.
i Shear Stress -. 6F
- Bending Stress l'.5F
.9F Compresuive Stress I.$5F, s,. 9F, Weld Stress
.45F, Factor of saf ety for expansion anchors is greater than 2.
For embedded plate Nelson studs, factor of safety
> 1.4 against conctete nitimate cepacity.
Damping ratios increased based on increased stress levels or test data.
No visibic signs of pe rmanent component / structure deformation are introduced resulting in questionable comporent/ structure performance.
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No excessive deflections are introduced resulting in spatial interaction with adjacent safety +related components / structures.
The piping in question may be considered operable provided the following conditions are satisfied:
The primary material' stresses do not exceed the higher value of.95,, 2.4S, or 3S, (Appendix F criteria).
g Secondary load,s have been considered in the support design.
No excessive deflections are introduced resulting in spatial interaction with adjacent safety related components / structures.
1 Rev. 2 Fare II.20 3 no3kc97.;tutete)
I
Design Guide 11.20 Civil /Strue. Oper. Reviews NOTE: Other evaluation conditions may be in. posed at the discretion of the Civil Discipline Manager.
In lieu of analysis, a test (static or dynamic) may be used and the component deemed operable if the test shows it can meet its intended function following a seismic event.
2.
ELLuw.trALEtyltiLhnal:
g This method of evaluation of operability concerns is I
to be used only with the concurrence of the Civil I
Discipline Manager or his designee.
Considerations to be addressed when using this method include:
{
Complexity of problem being evaluated.
I l
Similarity of the problem with other designs or e va l ua t i ori.
Availability of industry data directly relating to the issue.
Experience of engineers involved with related j
tasues.
I The purpose of the strur~ ural review method of i
operability determination is to utilize engineering l
judgement, experience, and evaluation of only those quality attributes crttical to the ability of the structure to function to its design requirements post-earthquake.
It will be used as an interim measure only, not as method to determine long term acceptability.
The method consists of two primary parts:
a)
System Walkdown:
The system whose cperability is in question vill be walked down and reviewed bv two experienced structural engineers These ' engineers shall have a minimum five years of nuclear structural experience.
The walkdown shall review and identify critical areas of potential f ailure and gather enough field data for an evaluation.
Examples of critical attributes include seismic anchor movements or spatial interactions. The appropriate Project or Principal Engineer vill outline to the Walkdown Team critical attributes which must be considered but will not limit the Team's judgement.
b)
Evaluation and Approval:
The evaluntion of system structural operability will censist of enough information to convey the logic used to determint that the system will function post-earthquake.
This could be a key. 1 Page II.20. 4 (HES$-1/geb)
i Design Guide II.20 Civil /Strue. Oper. Reviews simple write +up of the conditions considered with simplified calculations on critical attributes. This evaluation will be signed by both Walkdown Yeam members. The minimum approval of the evaluation will be three Civil Discipline supervisory personnel to include the Civil Discipline Manager.
Upon approval of structural operability using either of these methods, it is the Project Engineer's responsibility to ensure steps'are taken to document the evaluation in accordance with NED Guideline E 4 utilizing the Operability Review Approval Sheet (Attachment B).
Also, the Project Engineer shall be responsible for scheduling plant activities to restore the condition to long term acceptable status as soon as possible. This time frame is normally within one refueling outage.
If the condition does not meet the operability
- criteria,-the following steps should be taken:
The Civil Discipline Manager should be notified for concurrence with the evaluation.
Provisions of E 28 should be invoked to notify the plant to determine responsibility for' performing JC0 (Justification for Continued Operation).
-Work with the plant to determine if fixes can be made within system LCO (Limited Condition of Operation) window per Tech Specs.
Work with other NED discipline personnel to c
determine if component.is necessary for safe shutdown (i.e.; Mechanical, Electrical personnel may determine the component need.not operate post earthquake).
Document operability calculation in accordar.ce with NED Guideline E 6 utilizing the Operabi14ty Ra. view Approval Sheet.(Attachment B).
HQII: The criteria contained herein is for general
-conditions.
Specific criteria cited for L
specific conditions will supersede this
- i l
document.
C.
Reportability Reportability calls-to the NRC per Tech Spec guidance and 10CFR50.72 is the responsibility of the plant.
For Civil / Stress / Structural items,- the plant will l-request assistance in determining reportability once c
an item is dctermined to be inoperable.
Various plant l
L procedures are involved, however, a typical situation l
Rev. 1 Page II.20.-5 i
(HESS 1/geb)
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Design Guide 11,20 Civil /Struc. Oper. Reviews put the plant in a condition it could potentially not shut down sa f ely post ea rthquake.
Criteria to perform reportability evaluations for Civil are as follows:
Advanced evaluation techniques, such as plastic analysis may be used to determine the actual made of failure of the component.
Testing may be used on the component as a whole or critical parts.
Additienal criteria may be imposed by the Civil Principal Engineer or Civil Discipline Manager as conditions warrant, Results of the evaluation should be able to determine if actual gross structural failure of the component is expected and if that failure would put the plant in a condition in which it could not safely shut down.
The reportability evaluation documentation should include cause, corrective actions required, and address any similar plant conditions, g
Eev. O Page 11.20. 6 (HESS 1/geb)
4 Design Guide 11.20 ATTACIB1ENT A Civil /Strue. Oper. Reviews Pto2Iutt 70 firALUATE OrtitAtltiff of 8TSToes cinta thu P!Plive ICrat FICAfton 0F 1
St FltLD STAAf Ikyts11tatlow ut Docuss eiAtIcm REVIEW
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I Design Cuide II.20 Civil /Strue. Oper. Reviews ATTACHMENT B CAROLINA POWER & LICHT COMPANY OPETAMLITY REVIEW FOR
( PTsiil)
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~
EVALUATION ID NUMBER:
SAFETY Cl.ASSITICATION:
SEISMIC CLASSIFICATION:
METHOD OF EVALUATION UT.ILIZED: METHOD 1: SPECIFIC ANALYSIS / TESTING METHOD 2: SENIOR STRUCTURAL REVIEW PANEL VALEDOWN TEAM:
APPROVALS:
(Method 2 Only)
Project Approved Discipline h
Ey Checked Engineer Erinciral Eng.
Manager Rev. 1 Page 11.20 8 (HESS 1/geb)
..___m.___.,___.___-_._...__..~._..._..________________.
l Design Guide II.20 Civil /St rue. Ope r.14eva ews J
4 APPENDIX C l
l ASMF. APPEND 11 F EVALUATION i
l The following evaluation compares the NED Design Guide II.20 SD07, a
- H2O to two editions of ASME Appendix F.
Both pipe and pipe support i t.. r i a are compared. Note that ASME Appendix F is for pressure boundar" only and does not assure operability of components. This is due to large de(Iections allowed by the analysis. This should be considered if operabic valves are in the vicinity of high stress areas, The following is a comparison of che major stress limits.
tut is not in:1usive for all restrir.tions.
Refer to the particular document being used as the acceptance criteria prior to using stress limits shown below i
to deter:nine applicability. Also note that the stress limits for piping are given in terms of stress intensity for Class 1 analysis and not stress categories for Class 2/3 pipe.
This results in si.gnificantly different allouable limits and dif ferent stress roultipliers (stress indices vs. stress intensifiers) when using the rules of Appendix F.
i In summary, the stress limits that have been used for normal STSI evalutions for ENF are more conservative than those provided in Appendix F.
For example.
straight pipe stress limit i s 54ksi in A rendix F vs 36ksi NED II.20 allowable.
P Structural pipe support in tension is 1.2 Fy in Appendix F vs.9 Fy in NED II.20.
The length of time that a component found to be qualified only to some Interim criteria would reinforce the continued une of existing interim allowables au given in 11.20. My recommendation is that the Appendix F all?vables would more closely fit our current practice as criteria for reportability determinations.
i i
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s Civil /Struc. Oper.' Reviews l
t APPENDIX C NED II.20 SDC-7/M-20 APP. F 1977 APP. F 1986 Stress CateRory 1
Primary Higher of 2.4 Sh 3.0 Sm.
Lesser of I
Stress Pipe; 2.4 Sh or P c 2P design 3.0 Se or j
see Note 1
.9 Sy 2 Sy 4
Allowable ft <.9 Fy ft < Fy NF App XVII NF 3320
- Stress, is <.6 Fy fs <.625 Fy limits
- limits
- Component or NF faulted lesser of lesser of 2 Stande-d limits 1.2* Sy/Ft or or
]
Suppon
.7 Su/Ft 1.167*Su/Sy Approx 1.9 Approx 1.9 for A36 for A36 Note 1:
Only pressure boundary integrity is evaluated; component operability is not assured.
l 4
i Rev. 2 Page II.20-10 (1033DG92.CEw/ctie)
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