ML20101T963
| ML20101T963 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/16/1992 |
| From: | Starkey R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20101T966 | List: |
| References | |
| IEB-79-07, IEB-79-14, IEB-79-7, NLS-92-136, NUDOCS 9207220129 | |
| Download: ML20101T963 (21) | |
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CA&L i Carouna Power & Light Company l P O uva ital
- Ha'e ft.N C sT6 N JUL 101992 n e B M MPiY.JR Vu Presdent SERIAL: NLS 92130 N m t w .it%.v w United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Warhington. DC 20555 DRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50 325 & 50 324/ LICENSE NOS DPR-71 & DPR-02 MASONRY BLOCK WALLS
REFERENCES:
" Masonry Block Walls At Brunswick Steam Electric Plant, Units I and 2.*
- 2. Letter from Mr. R. B. Starkey, Jr. (CP&L) to Document Control Desk (USNRC) dated April 15, 1992 (Serial: NLS 92118),
- Masonry f3 lock Walls.'
" Masonry Block Walls At Brunswick Steam Eleculo Plant, Units 1 and 2."
- 4. Letter imm Mr. R. A. Watson (CP&L) to Document Control Desk (USNRC) dated May 29,1992 (Setial: NLS 9214Bh ' Corrective Action Plans,"
Gentlemen:
On April 9,1992 (Reference 1), the Nuclear Regulatory Commission requested information concerning masonry block walls at the Brunswick Steam Electric Plant, Units 1 and 2. The Company's responses were provided in a letter dated April 15,1992 (Reference 21. In a letter dated April 27,1992 (Reference 3), the NRC requested further information concerning the anchor bolt deficiencies and the Company's proposed corrective actions. Subsequently, on May 12,1992, a meeting was held between Carolina Power & Light Company and the Nuclear Regulatory Commission to discuss these structural and seisuic issues. By letter dated May 29,1992 (Reference 4), Carolina Power & Light Company documented commitments made during the May 12,1992 meeting with regard to major work items to be completed prior to start up of the Beunswick Plant, Units 1 and 2.
The purpose of this letter is to: (1) identify the commitments that were made during the May 12, 1992 meeting for implementation for each unit prior to start up from their next scheduled refueling outages, and (2) respond to the NRC Staff letter dated April 27,1992 concerning masonry block walls at the Brunswick Steam Electric Plant, Units 1 and 2. The April 27,1992 NRC letter contained a list of questions and issues to be addressed as part of a meeting between CP&L and the NRC. The NRC questions, along with Carolina Power & Light Company's responses which were discussed with the NRC on May 12,1992, are provided in Enclosure 1 of this letter. The information provided in the enclosed responses supplements the information from our May 12, I 9207220129 920716 4 PDR ADOCK 0500 0
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Document Control Desk l NLS 92130 / Page 2 )
l 1992 meeting. Where possible, references to the May 12,1992 meeting presentation slides have !
been included in the rqponses.
in the April 15,1992 i_ yonse, CP&L committed to develop by May 1,1992 a sampling plan and schedule for physical examination of raceway support anchor installation and building steel support anchor installation. The sampling plan and schedule information la provided in Enclosure 2 of this letter.
Information addressing questions concerning froren studs identified as part of CP&L's IE Dulletin 79-02 review is provided in Enclosure 3. l A summary of the commitments contained in the responses to the April 27,1992 questions 1 (Enclosure 1)is provided in Enclosure 4. If the action was also provided in our May 29,1992 .
letter, this has been noted- ;
Please refer any questions regarding this submittal to Mr. D. C. McCarthy at (919) 646 6901. l Yours very truvy, )
/h 9 R. 9. Starkey, Jr.f WRM/wrm (wallitr.wpf) l l Enclosures cc: Mr. S. D. Ebneter Mr. R. H. Lo Mr. R. L. Prevatte
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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 QUESTION l.A:
Discuss the causes of the apparent lack of timeliness of corrective actions for Emergency Diesel Generator (EDG) masonry wall bolting and service water pumps.
GP&L RESEQNH: -
Diesel Generator Walls:
In late 1986 or early 1987, a Brunswick Plant Technical Support organization engineer investigated -
a report from the maintenance instrumentation and control organization that diesel generator 1
building anchor bolts in angle frames were not penetrating into the column. The Technical Support organization addressed the concern in a site memorandum (BPE-5300) to the Brunswick Engineering Support Unit (BESU) dated February 13,1987. Due to reasons that could not be determined, a response which proposed sampling of the bolts was not made by the Brunswick site engineering unit until January 20,1988. An approval for bolt sampliig was sont to the Brunswick Engineering Support Unit via site memorandum BPE 6083 dated April 8,1988.. This site memorandum apparently was either lost or not received by the Brunswick Engineering Support Unit. Although the Brunswick Engineering Support Unit engineer involved in the response to memorandum BPE 5306 was transferred to another organization, his follow up eventually led to the discovery that the memorandum was lost. The Technical Support organization transmitted another site memorandum (BPE-6884) dated September 15,1989 to the Brunswick Engineering Support
- Unit requesting a response to the prior sampling approval. Also during the time period the memorandum was lost, re assignments within the Technical Support organization changed the engineer responsible for handling the wall deficiency.-
Ari engineer from the Nuclear Engineering Department investigated the cor,dition at the Brunswick Plant in November 1989. During the next year, the bolts in the diesel g'enerator building wall angles were as built verified using a feeler gauge. The data was used to analyze the diesel generator building walls. On November 15,1990, a short term structural integrity calculation (calculation number 015 34A 270) was completed. : Bas'ed on the calculation, the Company concluded that the walls met short term structuralintegrity criteria for the design basis loading. -
A permanent long term modification design was initially issued to the Brunswick Plant site construction organization on April 8,1991. This design was reviewed by plant organizations and was revised due to field constructability concerns; The design was re issued on February 19, 1992.-
= As discussed in our enforcement csnference on May 12,1991, the primary causes were:
- 1. Inadequate Technical Support and Engineering interface
- 2. Scope and impact of the wall deficiency was not recognized by personnel
- 3. Perception of resource limitations dominated scope of effort
- 4. No standards existed on acceotable time limits for short term conditional El 1
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- 5. Lack of a comprehensive line driven corrective action program
Reference:
See pages 1710 and page 27 of the May 12,1902 CP&L Technical Presentation package.
Service Water Pumos:
During 1979, service water pump deterioration was noted and an upgrade of the servica water pumps was initiated. in 1981, during the engineering process, an error in the origina! selsmic calculations performed for the original Peerless pumps was discovered. This calculation erroneously concluded that the pump was classified .4 a rigid structure due to r.n error in the pumps' natural frequency determination. Thl condition was reported to the Nuclear Regulatory Commistion &nd evaluated in letters dated April 10,1981 (Serlah BSEP 810791) and July 10, 1981 (Serial: BSEP 8101359). Brunswick Plant Modifications 81207 and 81208 (issued in 1982) installod new Johnson Pump components and upgraded the seismic qualification of the pumps to meet the bng term seismic requirements of the pump specification, in 1989, an additional seismic issue developed. inappropriate response spectra curves were found in an equipment specification for butterfly dampers. A non conformance report (NCR 89 8) was issued to address the problem's generic implications. In June 1989, during engineering for a service water pump upgrade to install a product lubrication enhancement, an incorrect response spectra was found in the service water pump specification. At that time, Engineering Evaluation Report 89 237 evaluated the upgraded pumps as capable of meeting their design functions. The modifications for the product lubrication upgrade were promptly changed to address the seismic response spectra error and were re-released in April 1990.
Concurrent with this effort, the original motor thrust bearings were evaluated as part of the system hydraulic review and found to require an excessive pump minimum flow, so as not to overload the pump motor bearing. Because this minimum flow requirement detracted from available system flow margin, a study was done to evaluate the best approach to resolve this issue. This study was completed in June 1990 and concluded the upgrade of service water pump motors was the best attemative. New pump motor bearings were designed and, through this process, it was determined that it would be possible to entircly climinate the service water lubrication system. Plant modifications were revised to incorporate the deletion of the lubrication water system and .
modification of the motors. These modifications were approved in 1991 and a purchaso ordor for pumps issued in September 1991.
Resolution of the seismic response spectra issue war integrated with other emerging service water system design issues relating to system hydraulics and single failure. As a result, the current service water pump design addresses seismic upgrades, eliminatici of the lubrication water system, minimum flow requirements for the pumpt, and chaiges to minim!.te camp maintenance.
Carotir.a Power & Light Company's current schedule is to begin replacement of the service water pumps starting in 1993. The service water pump upgrados are scheduled to be complead November 30,.1994. The Ccmpany has expedited vendor support for equ!pment deliveries ne6ded to achieve the schedule. The Company has deterrnined that during the Interim period, the service.
water pumps are capable of withstandirig a design basis earthquake, Carolina Power & Ught Company is conducting additional field inspections to provide assurance that thbae calculations are valid.= This inspection will be completed prior to start up of either Drunswick Plant unit. Based on the information known and the actions taken by CP&L to address service water system lasues, the Company believes that the actions.to address the servien water system seismic quaafication issue E12 4
have been timely.
Reference:
See pages 38 43 of the May 12,1992 CP&L Technical Presentation package.
M10 WESl10fLLQ:
Present the results of masonry wall bolt inspections, and provide the basa for the 25 percent sampling program for masonry wall bolts for walls other than those in the EDG Duil.iing.
CP&L RESEQfdE:
The results of the wallinspections are as follows:
- 1. Plated block shield walls (10 walls) (diesel contrator building, elevation 23 font)
- Approxirnately 60 percent of soll drilling e.<pansion anchor bolts missing l
- Approximatolv 7 percent of through bolts rnissing
- 2. Block walls e Diesel generator bui: ding, elevation 23 foot one block wall missing l 50 percent of anchor holts
- Diesel generator building, e:evation 50 foot 2 walls, sinaller diameter sleeve anchor bolts (6/8 inch in lieu of 3/4-inch) than designed
- 3. Reinforced concrete walls (diesel generator building, elevation 23 foot)
- Aporoximately 85 percent of anchor bolts missing Tho deficiencies were the result of original construction work. The walls in the diesel gt,narator building having structural angle restraints with expansion anr.horr. were 100 percent inspected by a combination of ultrasonic wmination and/or anchor / nut fornoval and re-installation. In addition, the walls with IE Buitttle + 11 modifications in both the control building and diesel generator building were reviewed by confirming quality assurance re.:ords with field reviews to ensure installations matched the drawings. No deficiencies wers found in any of the modification work prMormed in response to IE Dulletin 80-11, addition, a group of seismic walls outside 1he diesel generator building (i.e., the control building ud reactor building) were inspected. Of tho v.sismic walls,11 walls wera restrained by anchor bolted angles (6 walls by original design and constru: tion) Thess six walls were 100 percent inspected, In a few casus, anchors were discovered to be 5/84nch sleeve anchors in lieu of 3/4-inch anchors,- but all met IE Bulletin 80-11 requirements and in no cases woru fake bolts encourriered. The remaining five walls were post IE Bulletin 80-11 modified and found to be in conformance with design. No deficiencies were found.
Six walls in tne control building (elevation 40 f oot) were determined tu be required to be in place post earthquake for control room habitability requirements, but were deemed to be non-safety by
,. the 1980 IE Outletin 80-11 reviews. These walls were declared inoperable. As committed in our
! letter dated May 29,1992 (Reference 4), Carolina Power & Light Company will complete repairs l
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upgrading to seismic classification walls in the control building (elevation 49 foot) that have been determined to be required post earthquake for control foom habitability requirements. Also as indicated in our letter date:1 May 29,1992 (Refertince 4), Carolina Power & Light Company will perform a design review . sod a field inspr:ction review of non-safety masonry walls to verify the walls are appropriateiy clas lfied as non safety.
Referenns: See p1ges 13,22, and 244S of the May 12,1992 CPLL Technical Presentation packuge.
NBC.MELIMtRC:
Explain syhy you are inspecting less than 100 percent of through wall bolts. Are non functioning bolts t,, +,e removed?
CML.MSMNSE:
i One hundred percent of the through bolts in the diesel generator buil ding block walls have been ultrasonically exam lned with a total cf approximately seven (7) percent determined to be img roperiy installed. As indicated in our letter dated May 29,1992 (Reference 4), accessible non-functional through bolts will be removed and cover plates instrilled over the holes prior to start up of the two Brunswick Plant units.
Referenco: See page 21 of the May 12,1992 CP&L Technical Proson13 tion packar,0.
4 NBG.1111MDDBLLD..
Des;:nbs you. program for hspection and enalysis of reinforced concrete walls.
CP&l. RESPONSE:
Five reinforced concrete non load bearing wall penels were poured in the diesel generator building.
The Company has reviewed the balance of the plant and determined that the diesel generator building was the only location in the plant where this type of wall was utilized. Thess five walls are currently undergoing repair to restore them to their design configuration. As indicated in our May 29,1992 letter (Reference 4), these repairs will be completed prior to start up of the two Brunswick Plant units.
Reference:
See page 23 of the May 12,1992 CP&L Technical Presentation package.
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NRC QUESTION l.E:
With regard to pipe supports, you stated that the sampling technique and frequency of expansion anchor bolt inspections were in accordance with the requiremonts of IE Dulletin 70-02. You also stated that 'out of a total 433 anchors that were examined,156 anchors could not be fully evaluated because the stud (rod, bolti or leveling nut was, for unknown reason, ' frozen' or seized.'
If the bolts were frozen, that means those bolts could not have been backed out for measurements of bolt thread engagement length, anchor sleeves embedment lengths, and the anchor torquing could not have been verified. Explain how the samrling technique and inspection frequency used could have met the requirements of IE Dulletin 79 0., as stated above.
CP&L RESPONSE:
During the IE Bulletin 79 02 effo.t completed in the early 1980's, the sampling program cl.osen involved testing of one anchor per base plate as stated in our response to the NRC dated July 12, 1979 (3erial: GD 791739). The data showed, however, essentially all anchors were tested where possible. When the April 1992 CP&L audit team reviewed this data, it was clear the original IE Bulletin 79 02 sampling exceeded the one anchor per base plate requirement of the Dulletin, This fact formed the basis for the Company's conclusion that the sampling technique met lE Bulletin 79 02 requirements.
With respect to the number of frozen anchors, Carolina Power & Light Company's letter to the Nuclear Regu'atory Commission dated July 26,1982 (Serial: BSEP/821616) provides the summary data cited in our April 15,1992 response (Reference 2). A copy of the July 27,1982 letter is enclosed (Attachment 1). On page 5 of the letter, Qe total number of anchors tested is 433, and the number of anchors not tested due to frozen studi (117) or frozen leveling nuts (39) is a total of 150. The frozen anchors were not considered test failures or passes. For purposes of IE Bulletin 79 02, these anchors were considered as anchors not completely tested. As noted on page 4 of the July 27,1982 letter, those anchors with frozen studs were load tested, but could not be checked for thread engagement or embedment. Most of the 156 frozen anchors ht) groumd base plates. The grout was removed, providing access to the underside of the base plate.
Any fraudulent installation practices, such as tack welds or cut off anchors, would generally have been visible at the timt; the base plate grout was removed.
Thus, the large number of anchors tested and the detailed data sheets documenting the generally good test results lead Carolina Power & Light Company to the conclusion that the Brown & Root installed expansion t.achors on safety related pipe supports did not involve deficient installation practices.
The CP&L audit team leader discussed these issut:s with a member of the NRC Staff on April 30, 1992 at the Brunswick Plant site. The Company's understanding was that all NRC questions concerning this issue were addressed satisfactorily.
MlE CLE11103.LE:
With respect to Design Guide 11.20, ' Design Guide for Civil / Structural Operability Reviews" (DG),
for piping and p1 ping supports, the staff finds that the DG doss not address or inadequately addresses the following attributes in the op9rability determination criteria:
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(1) How the comprehensive loading combinations for both normal and faulted conditions are .
considered in the criteria. ,
(2) What damping values and response spectra are to be used, as well as a comprehensive j methodology and analysis procedure similar to Ft. Calhoun's and Dresden/Ouad Cities',
which have been accepted by the staff.
(3) How other occasionalloads, including water hammer or steam hammer, as well as secondary loads, are to be used.
(4) The appropriateness of using the " Structural Review Panel," in lieu of a comprehensive i evaluation provided in the DG.
Explain how these issues are oddressed in your operability evaluations for piping and piping suppore.
CfikflESPONSE:
eda t:
Additional documents applicable to the Brunswick Plant provide the specific load case combination information. Study Report M 020, ' Criteria for Evaluating and Performing Computerized Piping Analyses of Existing Systems with Minor Modification," Appendix A, page 3 of 5, provides load combinations and stress limits for piping. Study Report M 021, ' Evaluation Criteria for Existing Pipe Supports Associated with NRC Bulletins IE79-02,79-07, and 7914," provides load com'inations c for pipe supports. Coples of the two study reports referenced are enclosed (Attachments 2 and 3). Additional structuralload combinations are provided in the applicable sections of the Updated Final Safety Analysis Report.
EEL 2:
The response spectra critical damping ratio used for Brunswick Plant short term evaluations is PVRC N 411 damping. The use of N 411 damping for the Brunswick Plant was approved in an August 28,1985 NRC letter to Carolina Power & Light Company, in some cases, a higher damped l curve wi:1 be invoked to provide a "first cut" at the operability assessment, Final assessment would be based on time history or gap evaluation.
EEL 3:
The specihc Brunswick Plant short term structural integrity (STSI) load case combinations specified in Study Reports M 20 (Piping) and M 21 (Supports) require that any applicable system transients be considered in an operability evaluation. Secondary (i.e., self limiting) loads are not considered in operability pipe stress evaluation; however, secondary loads are cons!dered in the pipe support load combinations.
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PJut4:
The use of a ' Structural Review Panol" concept was based on the following:
- The use of experienced structural engineers to make field assessments of nonconforming con ("tions.
- Use of previously performed, similar structural evaluations.
- Encouragement of ' hands on* engineering in lieu of drawing only review.
- Use of calculations as backup to reinforce judgements where appropriate.
- lssue plant instructions to repair conditions, i Experienced engineers use this process to provide hand calculations, strdy analysis, or document judgement to accept existing items that require modification for full coinpliance. This allows the engineer to quickly move to the redesign phase and focus efforts on providing a rigorous long term calculation and applicable modification drawings. These criteria are not used for long term documentation of structure acceptability. The long term documentation calculations currently performed provide a complete, cornprehensive, and detailed evaluation, it should be noted that the Brunswick Plant piping offort now underway is a design basis reconstitution effort to address as built errors discovered in 1980. Between 1979 and 1980, a complete IE Bulletin 7914 program was performed, resulting in the analysis of identified safety lines and the modification of over 2000 supports (both units). The reconstitution effort now has re-inspected approximately 3500 supports since 1980. Since original calculations and modifications were previously performed, the ' Structural Review Panel
- operability review would consist of a documented review of the as built differences identified by the re-inspection against the existing IE Dulletin 7914 analysis to ensure operability. The piping would then be re-analyzed in the updated, as built condition and any "long term" fixes identified and issued. This method of review of as built impact on existing analyses has been used for operability determinations at the Brunswick Plant.
In additi on, DG 11.20 is being revised to incorr.nate Brunswick specific piping and support criteria, as well as update requirements for short term operability in accordance with Generic Letter 91 18.
Reference:
See pages 46-48 and page 50 of the May 12,1992 CP&L Technical Presentation package.
NRC OVES 110Mlfi:
Your root causs 'se provided a discussion of the paper work that you had reviewed but raached no cd br, ion. Discuss the progress you have made with respect to determining the root cause.
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See the response to Question I.A above.
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l NRQ_QUESTIONJJ:
Discuss the progress you have made on your ple 1.: to inspect and correct identified deficiencies.
CP&L RESPQNSE:
All masonry wr.Ils that were considered seismic in the origin ,1 IE Bulletin 8011 walkdown with stro tural angle restraints attachod with expansion anchors that are located outside of the diesel generator building have been reviev ed. The results of these inspections are summarized in the response to question I.B.
NRC QUESTION 11.A:
Address how you determined there was an irsue, why it was overlooked in your earlier response to our letter, and what actions you took to evaluate and correct the deficiencies.
Cfs.L[lESPONSE: ;
Cosmetically applied anchor bolt heads in structural angle restraints attached to reinturceu oncrete walls in the diesel generator building were found to be un issue when engineers supporting repairs of the masonry wall missile barriers questioned the installation of similar structural angle restraints on local reinforced concrete walls, even though their design basis in a reinforced concrete wall was not immediately apparent. These walls were not addressed in CP&L's April 15,1992 response
, because they were found afterward, Please note that the responses provided in the Company's April 15,1992 letter (Reference 2) were preliminary and based on the best available inforrnation.
n fLRC QUESTION ll.B:
in light of yw recent idantification that .ie anchor bolt deficiencies were more widespread in the DG building than originally anticipated, discuss your plans for validating the original conclusions -
resulting from your IE Bulictin 80-11 program reviews.
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CP&L RESPONS.E:
An overall review nf the IE Butletin 80-11 program is underway for the Brunswick Plant. The review will address existing masonry wall functions, including missile barrieri tornado barrier, ventilation barrier, or other functions for which it is not analved.
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- NRC QUESTION 1114
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Characterize the type, number and safety significance of tho backlog of items Qualified under your ,
short term structural integrity program.
l Cf&LRESfQN$li:
The current short term st'uctural integrity list is tracked by a total of 48 octstanding items. The items can be characterized as follows:
} ITEMS CONTENTC IMPACT COMPLETION
! SCHEDULE 41 217 Pipe Support Minor Repairs Reload 8 (Unit 1)*
Design Turnovers Short Term Oualified Reload 10 (Unit 2)*
Identified 1 Service Water Pumps Short Term Qualified 11/30/94 1 Diesel Generator Fixed Prior Building Walls (Long Term Qualified) to Start-up 1 Air Tubing Supports Short Term Qualified Reload 8 (Unit l}
Reload 10 (Unit 2) i 1 RWCU Supports Short Term Qualified Reload 8 IUnit 1)
Reload 10 (Unit 21 1 Fuel Oil Small Bore Short Term Qualified 1995
( 1 Main Steam Radiation Short Term QualiEed Reload 8 (Unit 1) j .
Monitor - Reload 10 (Unit 2)
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l 1 Diesel Generator Short Term Qualified' Prior Exhaust to-
.-Start up
' Exceptions: Diesel Generator Service Water Supply and Return - Unit 2 Reload 11 -
l - Service Water Lubrication Water Supports - Unit 1 Reload 10 and Unit 2 Reload 11 l
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These items are tied to existing modification schedules.
The Brunswick Unit 1 Reload G outage is currently anticipated to begin Maren 1993. The Brunswick Unit 2 Reload 10 outage is currently scheduled to begin September 1993.
Of the total of 48 items listed, engineerir.g for long term fixes is complete on thirty-four, in general, the repairs are typically the addition of gussets, beam stiffeners, adjusting U-bolts, additional welding, component replacement, and the removal of snubbers. Because CP&L is reviewing, reconstituting, or redoing calculations done by our architect-engineer in the i980 to 1986 time frame for piping and supports, the type of repairs are not as significant as would be found in a first time IE Bulletin 7914 review. The Company hLs adopted more conservative criteria in accc,7 dance with up-to-date analysis methods as tfn work has progressed. Efforts are undorway to ensure a complete short term structural integrity program is outlined and documented, noting both items under review as well as those with fixes issued. The Company has initiated a third-party review of the short term structural integrity program which may result in the identification and scheduling of additional re p The short term structuralintegri.y list abcve will be reviewed for completeness and any additional items will be added, if necessary.
The above items were characte<ized for safety significance by evaluating them in a limited scope seismic probabilistic risk assessment of the Brunswick Plant and then by verifying the results and assumptions with plant walkdowns. From the list of short term structura! integrity items, only the service water pumps and the diesel generator building walls were found to be significant contributors to core damage frequency. The pipe supports were evaluated as not significant and the remaining items were screened out based on lack cf applicability to safe shutdown as
- determined in the probabilistic risk assessment modol.
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Seismic interactions from the service water pumps and the diesel generator building walls were -
incorporated into a limited scope seismic probabilistic risk assessment to examine the safety significance of the as-built condition versus the as designed condition of these two short term structural integrity items. Seismic fragilities for the pumps and the walls were deve!oped for the as-built and as designed conditions. The core damage frequency in the as built condition increased by a factor of approximately 2.3 over the as-desig".ad condition. Howevei, in absolute terms, the core damage frequencies were small(i.e., on the order of 1E 5 per year for both cases).
The pipe supports were found not to be a significant safety concern . Plant isometric drawings showing the pipe supports in question were reviewed and evaluated using the probabilistic risk assessment model of the plant to determins if they were concentratr.d near safety significant ,
components such that multiple safe shutdown paths would become unavailable during a seis.nic event. Effects such as " unzipping" of a series of supports and seismic-induced flooding were also considered. The plant walkdowns confirmed these findings and also verified that other potential pipe or valve interactions not identified in the drawing review wou!d not significantly affect the estimated core damage frequency. Overall, the supports were found to be of high quality material and good construction.
It is concluded that these as found conditions resulted in an increased estimate of cora damage frequency from seismic events. However, the value of this increased estimate is well within the range of core damage frequencies from other seismic probabilistic risk assessments and is not considered to be a significant risk.
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NRC OQESTION ElJ:
Discuss the schedule for correcting those items and the reason more timely corrective action was not taken.
CP&L RfiSEQNSE:
The schedule for correcting the backlog of short term structuralintegrity program items is provided in the response to Question Ill.A above. The primary reason more timely corrective action was not taken with regard to these items was the lack of standards establishing acceptable time limits for short term conditions to exist.
Reference:
See page 27 of the May 12,1992 CP&L Technical Presentation package-MC OUISJION ll? C:
- Provide the basis for assumed validity of existing analyses for short term structural integrity in view ,
of defic;encias found recently in analysis for CBEAF (Control Buildirig Emergency Air Filters) supports and masonry wall bolting.
CfALBESPONSE:
As discussed in the preceding questions, the overall short term structural integrity list is dorninated by pipe support repairs resulting from the piping design tumover efforts. Most repairs are somewhat minor in nature. However, to ensure the proper evaluation of the items addressed, Carolina Power & Light Company has initiated a third party review of the short term structural integrity program. The review will addrsss evaluation techniques, field validation of critical assumptions, as well as a review of other communications from the Technical Support organization to the Engineering organization. This revi2w began the week of May 11,1992 and is emected to l be complete by July 31,1992.
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Reference:
See page 52 of the May 12,1D92 CP&L Technical Presentation package.
. MQ_ QUESTION 111.0:
Provide the basis foi design va!ues assumed in masonry wall analyses (i.e. bolt, mortar, rebar and -
grout stiength).
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f CP&L RESPONSE:
Bolu:
Manufacturers' allowables for Phillips Redneads were used in the original construction. These allowables use a safety factor of five (5). Subsequent research by ths Seismic Qualification Utility Group !SOUG) for USI A-46 confirms these allowables. The allowablos for the Phillips bolts used in the short term evaluations were below the SOUG or vendor allowables.
Mortar. Block, Grout Strenath:
Allowable stresses for both flexible members and shear walhi have been established based on tests for shear walls.
Two major test progra es have evaluatad the sheer strength of concrete block masonry walls. The first test was performed by Schneider; his test results were used as the basis for developing the USC, NCMA, and ACI code allowable stresses for reinforced masonry. A subsequent test program was performed at the University of California - Berkeley, These test results, were used as a u comparison with the code allowables. Therefore, code allowables aregenerally accepted and
.- specifically used at the Brunswick plant.
NCMA Code Method 2 was used to establish concrete masonry strength. This method requires testing of concrete masonry units. The allowable compressive stress (f'm) is determined from the test results for various me: tar types. Specific test results for concrete masonry units at the Brunswick P! ant are attached (Attachment 5).
No tests were performed on the mortar; however. Specification 9527-01-29-1 required that mortar adhere to the following ASTM Standards: ASTM C91, ASTM C144, ASTM C270, ASTM C476, and ASTM C780. }
FAE Vertical reinforcement bars are deformed bars ASTM 615-68 Grade 60 for sizes Number 6 to l Number 11 and Grade 40 for smaller sizes.' Horizental joint reinforcing is standard Dur O Wat galvanized collar joint reinforcing (verticall is welded wire tabric. Presence of rebar in both masonry and concrete walls were determined by magnet:c rebar scanner.
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Reference:
Seu page 33 of the May 12,1992 CF&L Technical Presentation package.
NHCLQQfftIlON lli.E:
Describe the quality controls applied to verify bolt torque values during recent masonry wall work.
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, GP.ALfl.ESPRNSE:
Permanent construction work at the Brunswick Plant is performed under approved Quality Control (OC) procedures. W 'l repair work underway is be!ng ronducted with the appropriate OC !
verification of work and materials. )
The verification of anchor existence during the week of April 6,1992 was performed to obtain
- information to determine the operability status of the plated missile shield walls. The 3/4-inch Ehillips lied Head self dri!!ing anchors were backed out, checked for length, and re installed to a
' snug tight" condition. This work was directly observed, supervised, and documented by Nuclear Engineering Department site engineering staff psrsonnel, j l
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I fillG.DMESTION llLE:
Explain the non-uniformity in the use of steel angles on masonry walls in the switchgear rooms in )
the EDG building, and in the use of steel bracings for the stairwell enclosures in the same rooms, GEAL.BE.SPONSE:
A floor plan for the diesel generator building showing the interior wall designations for elevation 23 foot is provided in Attachment 4.
In the switchgear rooms, three types of block walls exist:
- Plated block walls which are part of the standard missile shield wall typically between the diasel generatort.
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- Unplated block walls between the switchgear.
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- Unplated block walls which form an enclosure around the stairwell and serve as a fire barrier between floors.
The typical detail for plated block walls required steel angles on the top, bottom, and sides of the wall due to sr,ismic and miss!!e loading. These angles were installed except where penetrations or other obstructions exist.
l- The typical detail.for unplated walls between switchgear requires steel angl63 only on the top of the wall; The sides art restrained by mortar joints with dove-tail anchors (as required), and the bottom is restrained by a fully bedded first course. The reinforced concrete itself serves as the missile barrier.
The typical detail (original construction) for the stairwells uses no steel angles.-
l Steel bracing for the stairwell enclosures was installed as an IE Bulletin 8011 fix for Wall 9a. This j wall was classified as safety-islated due to proximity to safety-related equipment. The mirror-image of this wall on the north end of the building is Wall 9a. Wall 9a is classified as nonsafety-related because the potential targst (safety-related equipmenti does not exist in this end of the
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building. Wall 9a was downgraded to non-safety in later IE Bulletin 80-11 submittals and, therefore, stool bracing was not installed. However, recent reviews have determined that wall 9a
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is required to support wall Sc, which is safety related. Therefore, additional analysis is in progress to determine if steel bracing is required for wall Sa.
El-14
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ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS. 50-325 & 50-324 OPERATING LICENSE NOS. DPR 71 & DPRI 62 MASONRY BLOCK WALLS .
PLAN AND SCHEDULE FOR PHYSICAL EXAMINATIONS i Et090 Carolina Power & Light Company plans to select, by means of walkdowns and design document research, a substantial raceway run, a substantial building steel sample, a substantial HVAC duct run , and a sample of various equipment foundations with accessible Red Head self drilling snap off anchor bolts that are original plant construction in each of the following buildings for each Brunswick Plant unit: the control building, the reactor building, the diesel generator building, and o
MdinfThe~d.iweilis excluded from the ini'lai sample in order to reintain -' --
personnel exposure as low as reai:onably achievable. The drywell may be included in an expanded sample if results of the above sampling demonstrate the need.
Insoettion:
For the anchors selected, the inspections will primarily be conducted by ultrasonic testing with the uption of removing anchors for inspection in order to identify deficiencies such as missing, cosmetically applied, or faked anchors. During the walkdowns and inspections, personnel will also observe for any bolt heads that appear to be welded to their plato that may or may not be part of the sample. The need for sample expansion will be dotermined by CP&L management based on the results of the inspections.
Schedule:
The walkdowns and design reviews are currently in progress to identify the structures to be?
inspected. Completion of the inspections is expected to be completed by July 31,1992, pasumentation:
. Inspectica 9sults will bo documented, including the defkiencies identified. Documentation wilt be adequate to permit the inspection results to be verified later. Deficiencies identified will tilso be
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doccmented in accordance 'with applicable plant procedures and Nuclear Engineering Department guidelines. The need for sample expansion will be determined by CP&L management based on the
. results of the inspections.
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ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND e-NRC DOCKET NOS. 50 325 & 50 324
-- OPERATING LICENSE NOS. DPR 71'& DPR 02--
MASONRY BLOCK WALLS' IE BULLETIN 79-02 (SP 79 22) REVIEW' The purpose of this information is to address questions concerning the fourteen (141 frozen studs in surface / flush mounted attachments reported in Carolina Power & Light Company's letter dated -
April 15,1992 (Sedal: NLS-92118)/ This information was discussed in CP&L's response to NRC -
Question 2, Response item A.3, " Frozen Studs,"_ on page El 4.,
W141)-oL'hess-were thraedad4tudtandlour_(41were hex head boltsc Although frozen, the J threaded studs were still properly and successfully tested idr~dtsign load capa@y; maatinn tha a requirements of Special Test Procedure (SP) 79-22 (" Proof Load Test"). The hex head bolts could '
not be propeny tested since it was impossible to provide a gap behind the baseplate for insertion of 1/4 inch shims as required for the " Proof Load Test." Test results were inconclusive for.these four hex head boltsJThe Company's review of the installations using these hex _ head bolts _(drawing - _
numbers L 022601372 and L 02200-5230, Sht.1) resulted in the following conclusions:
- 1. L-033601372, Mark No. PS 1372 (Iso. D-02840, Sht.178A D.P.3)L This support is installed on line number 2 RNA 222-1-170 which is Quality Class D (non-safety related). The' support anchors on this support will be reworked prior to start-up of either Brunswick unit.
- 2. ' L-02260-5230. Sht: 1, Mark No. PS 5230 (Iso. D-02846i Sht 166Ai D.P 560)
Based on the Special Procedure 79 22 documentation, only one bolt of this two bolt-installation was frozen. The other bolt passed inspection, but could not be properly tested '
ouc to the frozen bolt. This support was re-inspected under the IE Bulletin 7914 closecut--
program. Both bolts were reworked under trouble ticket 88-ARUU1 (the bolts tightened and torques to 110 foot pounds). As a result of lino re analysis, this support is now also non-O. Therefore, the Company believes that no further actirm is needed.
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ENCLOSURE 4
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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 4
SUMMARY
OF COMMITMENTS As discussed in the information provided in Enclosure 1 of this letter, Carolina Power & Light Company commits to complete the following activities:
- 1. Carolina Power & Light Company's current schedule is to begin replacement of the service water pumps starting in 1993. The service water pump upgrades will be -
completed by November 30,1994. The service water lubrication water system will
__beJerr.avedyart of this replacement.
- 2. Carolina Power & Light Company will, prior to start-up of either Brunswick Plant unit, complete additional field inspections to provide assurance that calculations supporting interim seismic operability of the service water pumps are valid,
- 3. Carolina Power & Light Company will complete repalts upgrading seismic -
classification walls in the control building (elevation 49 foot) that have been determined to be required post earthquake for control room habitability requirements. (This action was included in CP&L's May 29,1992 letter, Reference 4).
- 4. Carolina Power & Light Company will perform a design review and a field inspection review of non-safety masonry walls to verify the walls are appropriately classified as non-safety. (This action was included in CP&L's May 29,1992 letter, Reference 4).
- 5. Carolina Power & Light Company will remove accessible non functional through-bolts and install cover plates over the holes. :(This action was included in CP&L's May 29,1992 letter, Reference 4).
- 6. Carolina Power & Light Company will complete repair of five reinforced non-load
- bearing wall panels in the emergency diesel generator building to restore the walls to their design configuration. These repairs will be completed prior to start up of the two Brunswick Plant units. (This action was included in CP&L's May 29,1992 letter, Reference 4).
- 7. Carolina Power & Light Company is revising Design Guide 11.20 to incorporate -
Brunswick specific piping and support criteria and to update requirements for short term operability in accordance with NRC Generic letter 91-18.
- 8. Carolina Power & Light Company will perform a review of IE Bulletin 8011 program -
for the Brunswick Plant. The review will address existing masonry wall functions including missile barrier, tomado barrier, ventilation barrier, or other functions for
. which it is not analyzed.
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- 9. Carolina Power & Light Company will complete long term qualification of the 217 '
identified pipe support items p;ior to start-up following the Unit 1 Coload 8 outage and the Unit 2 Reload 10 outape.-
- 10. Carolina Power & Light Company will complete long term qualification of the instrument air system tubing supports prior to start up following the Unit 1 Reload 8 outage and the Unit 2 Reload 10 outago.
e
- 11. Carolina Power & Light Company will complete long term qualification of reactor water cleanup system supports prior to start up following the Unit 1 Reload 8 outage and the Unit 2 Reload 10 outage.
- 12. Carolina Power & Light Company will complete long term qualification of diesel fuel oil smMI bore lines prior to start up following the Unit 1 Reload 8 outage and the-Unit 2 Reload 10 outage.
- 13. Carolina Power & Light Company will complete long term qualification of the mein -
steam line radiation monitor supports prior to start up following the Unit 1 Reload 8 outage and the Unit 2 Reload 10 outage.
- 14. Carolina Power & Light Company will complete long term qualification of the emergency diesel generator exhaust supports prior to start up from the current outage.
- 15. Carolina Power & Light Company will complate long term qualification of the diesel generator service water supply and return line supports prior to start-up following '
the Unit 2 Reload 11 outage.
- 16. Carolina Power & Light Company will, followmg the installation of new service water pumps in 1994 which do not require an external lubrication water system, decommission the remaining portions of the existing service water lubrication water system and its supports. This work will be completed by the end of the Unit 1 -
Reload 10 and Unit 2 Rcload 11 outages.
- 17. Carolina Power & Light Company will ensure a complete short term structural integrity program is outlined, including items under review and items with fixes issued.
- 18. The short term structural integrity list will be reviewed for completeness and any additionalitems wil be added,if necessary. .
- 19. Carolina Power & Light Company will perform a third party review of the short term structural integrity program to address evaluation technioues, field velidation of '
critical assumptions, and a review of commu_nications from the Technical Support j organization to the Engineering organization. The third-party review is expected to be completed by July 31,1992.
A E4-2 J
t ATTACHMENT 1 CP&L LETTER DATED JULY 20,1982 (SERIAL: BSEP/821610)
-SUPPLEMENTAL RESPONSE TO lE UULLETINS 79 02,70-07/ AND 7014 T
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1 G ea$ i ia >U Cartlina Psw:r & Ught Comp;ny W(w W ctfI , ,
)
gr F Brunswick Steam Electric Plant P. O. Box 10429-Southport, NC 28461-0429 July 26, 1982 FILE: 509-13510C SERIAL: BSEP/82-1616 Mr. James P. _0'Reilly, Director U. S. Nuclear Regulatory Ccemission Region II, Suite 3100 .
101 ?!aristra Street N.V. '
Atlanta, GA 30303 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 2
, DOCKET NO. 50-324 LICENSE NO. DPR-62 SUPPIE"hTAL RISPONSE TO IE BULLETINS _79-02, 79-07, AND 79-14 4
Dear Mr. O'Reilly:
In our letter (BSEP/81-0440) dated February 25, 1981, we committed to completo the Phase I and Phase II portion of the seismic reanalysis of the plant to satisfy the requiremects of IE Bulletins 79-02, 79-07 and 79-14 by July 31, 1981, and March 31, 1982, respectively. This letter is to report the Phase I, or generic analysis, and Phase II, the individual analysis, have been completed in accordance with these dates, s The Phase I and Phase II-programr. did not includa as-built evaluation of inaccessible isometrics as that work required a unit outage for access to complete. These isometrics on Unit No. 2 have been as-built and reanalyzed during the current outage. Two inaccessible isometrics in Unit No I remain.
These will be as-built and evaluated during this year's outage.
In the February 25, 1981, letter, we provided a list of potantial problem areas and inconsistencies that were discovered during cur review program, together with intended resolution and schedules. We will address the s,tatus of these areas in the same order as listed previously. -
hf b:$ b '
Mr. O'Reilly .
July 26,1982 A.
Lines Originally Seismically Analyzed. But Not Bulletin 79-07 Efforts Included in IE . .-
Upon the completion of the seismic lina review, 38 isometrics remained to be analyzed.
February 25, 1981 Twenty-onc of these isometrics were addressed under the letter.
under the Phase I program. The remaining 17 (isometrics) were analyzed on this analysis. There were no short-term fixes required based have been issued and are in the process of being installed.All long-ter Vs plan to end of the next Unit No. complete the fixes associated with-these isometrics on both scheduled to stare in September 1 refueling
_1982. outage. This outage is presently For Unit No. 2, approximately 35 the current outage, however, due to insufficient _
If outage time. in necessary, these few remaining supports will be completed during the next available nart Unit No.
outage 2 refueling of tufficient duration, and no later than the end of the outage. ,
BI Vents. Drains, Instrument Connections These they didconnections not were not covered by the original computer analysis so fall-under the scope of IE Bulletins 79-07 and 79-14 It was determined they should be evaluated to give reasonable assurance that they did not significantly affect the process piping.
The remaining small bore piping was handled by a samp
\pproximately
- o. rstress or thehalfprocess of these connections were analyzed with no cases of pipe.
signtricant It was thus concluded that no impact on the-analysis of the parent lines existed.
C.
Unanalyzed Loads Due to ' Valve Eccentricity -
In our letter of February 25, 1982, approximately 25 motor-operated All but four have been evaluated based on UELC estimat .
,s operator weights and centers of gravity.
values with vendors have indicated that the estimated values are asEfforts to v accurate samfac (t 10. percent) an any values which could be supplied by the turers a 10 percent varianca in weights and the vendor's estimates wi improve the accuracy of the analysis, the vendor verification program was terminated.
analyzed and The remaining four valves were not originally computst therefore, are not encompassed by IE Bulletin 79-07 However, a ge,neric analysis was performed on these lines which verifi .
that the piping stresses are within ANSI B31.1 limits .
L _ _ _ -_-
Mr. O'Reilly July 26, 1982 D. Verification of Accentable Containment Penetration Nozzle Loads All penetration nottle loads-have been verified as acceptable per the requirements of IE Bulletins 79-07 and 79 14.-
E. Vender Supplied and Vender - A/E Interface Piring After a review of vendor documentation, we have concluded that these lines were not cooputer analyzed. IE Bulletins 79-07 and 79-14 thus do not apply.
F. Smell Nozzle Loads on Safety-Related Components The only lines encompassed in this category are the vent and drain lines off the HPCI, RCIC and core spray pumps, which were ana,1yzed.
The IE Bulle' tin 79-14 condition has been reviewed and no short-term fixes were required. Long-term fixes are scheduled. on. the same basis discussed for Item A. .
G. Seismic Requirements Inconsistencies Only two lines under this category were found- to require analysis; one is the surge line in the Diesel Fresh Vater Cooling System, the other is a drain line in the Standby Liquid Control System. These lines are small ,
and were not originally computer analyzed. Therefore, the 79-07 Bulletin is not applicable. However, in order to completely close out all outstanding items, these lines were as-built and evaluated as part of Item A.
H. CRD System Baseplate Flexure Analysis In regard to our bulletin r' quirements e for _the CRD System supports, we
, stated in our February 25, 1981 letter, " Completion of baseplate flexure l \ analysis on CRD piping not essential to safe shutdown is scheduled for l completion as part of the Phase II Program." CP&L has determined that nonessential portions of the CRD System are'not safety related or seismically qualified;- therefore, this analysis was not required.
I. . Anchor Bolt Testinx -
As stated in our February 25, 1981 letter, the scheduled anchor bolt
. testing per IEB-79-02 of all the additional supports identified for testing is now complete for Unit No. 2.
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Mr. O'Reilly -4* . July 26, 1982 The testing of self-drilling anchors included application of a torque representing a pull out load equal to or greater than the allowable casign load for the anchor. Concrete embedment and thread engagemen were also measured whenever it was possible to remove the bolt / stud from the anchor. It must be noted here that this phase of test program covered many floor mounted supports employing self-drilling anchors with all thread rod studs and grout. Because of moisture conditions during plant 'peration several studs were found to be frozen in anchors and could not be removed for measurement of depth. All of these anchors, however, either passed the preload test or were replaced.
All supports that did not meet the test acceptance criteria were conservatively evaluated for the load values generated by IE 79-07 j reanalysis effort. Repairs were made to deficient supports and the frozen studs broken during test. Rusted self-drilling anchors in service water intake structure were replaced by stainless steel wedge anchors.
4 A total of 163 baseplates containing 433 anchors were included under this phase of the program. All baseplates and anchor bolts were tested to the extent as was reasonably pos'sible. The primary test verifying adequate preload was performed on 88 percent of all anchor bolts and on at least one anchor bolt on all of the baseplates except two. One'of these baseplates had a seismic load of one pound and the other a safety factor of 20. These loads are sufficiently low 't hat the satisfactory inspections of their condition when testing was attempted was adequate to 1
assure their reliability. The preload test demonstrated the actual chility of each anchor bolt to withstand its design load.. The failure rate for this test was 2.4 percent. The inability to back off the leveling nut was the predominent reason for not testing all of the anchors. A. stuck leveling nut does not indicate any structural deficiency with an anchor, .it just prevented any meaningful testing.
The low failure rate and the extensiveness of the test program for both baseplates and anchors provides a high confidence in the' ability of the 1 existing anchor bolts to accommodate the required loads.
Tests for proper installation were performed on 59 percent of the anchor bolts. A failure rate of 1.6 percent was obtained for improper engagement and 1.6 percent for inadequate embedment. Problems with anchor bolts or studs which could not be removed (27 percent of all anchors), in addition to the previously mentioned frozen leveling nut problems (9 percent of all anchors),- were the' ovarriding reasons '
preventing full testing. All of the anchors with unremovable bol.ts or studs were successfully tested for preload, however, demonstrating the .
load capability of the anch' ors. This satisfactory demonstration and the low failure rates indicate there is no concern for inadequate embedment and engagement. In addition, 31 percent of these anchor bolts which were not fully tested were subsequently replaced for other reasons further reducing the number of not fully verified anchors.
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Mr. O'Rti,11y /
Ser+
gy gp/ y:-/6 /6 July 26, 1982 An program overall failure on Unic No.rate
- 2. of 5.6 percent was obtained from the testing failure rates from the tests, and the number of anchors which were replaced, leave a small cpportunity for inadequate baseplates.
Anchor Bolt Testing Results Summary on Unit No. 2 , -I #3 Total number of baseplates
- 79 Number of baseplates tested 163 Total number of anchors 161 433 Number of anchors tested for preload Number of anchors failed preload 360 9
Preload test failure rate 2.4%
' Number of anchors not tested for preload 53-Number not tested due to frozen leveling nut 39t Number tested for other reasons 14 Number of anchors tested for embedment 254 Number of anchors with inadequate embedment 4
Embedment test failure rate 1.6%
Number of anchors not tested for embedmont .
179 Number not tested due to frozen leveling nut 39 Number not tested due to frozen stud 117 Number not tested for other reasons 23 Number of anchors tested for engagement 25 6 Number of anchors with inadequate engagement 4 Engagement test failure rate 1.6%
. Number of anchors not tested 'for angagement 177 s Number not tested due to frozen leveling nut 39 Number not tested due to frozen stud 117 Number not tested for other reasons 21 Total failure rate 5.6%
As Unitrequired No. 2. Unit by IENo.Bulletin 79-02, CP&L has completed the test program for I testing is essentially cemplete. The results are -
being tabulated and untested. _During the upcoming Unit No.
checked to assure no $dentified supports remain .
September 1982, any supports not yet tested in the primary contairment1 outa will be tested and results triasmitted to your office.
l yd) #e kd1 T to 4- 2n n rs = h 0 o
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.1 Mr. O'Rolllyi July.26,- 1982- [
" In our February-25,'1981 letter, we committed; to perforcing a weld . ;
verification sampling-program for seismic pipe supports-as-part of the Phase I program. _We have completed this sampling prograa with greater '
than a 95 percent confidence level;that the original QC inspection 1 program was adequate. This-95 percent confidence:1evel is-c'onsistent
- with that required for th'e:IE8 79-02' sampling programs and thus we
~
believe our: pipe support welds are acceptable.
In conclusion, upon the completion of the -long-term fixes discussed previously, thez Pipe Stress.' Analysis Summary Tables will:be updated to-indicate completion. of the field: modifications. - This update will signify our completion of work and compliance with the above bulletins.- We '
anticipate this milestone v111-occur in mid-1983,'at which. time you will be-notifled in writing.
' t ruly . yours ', --
INAL SIGNED BY[ -
C. ! R tzheneral. .'.*anager Brunswick-Steam Electric Plant-JSB/ds cc: Mr. R. C.-DeYoung- .-
bec: Mr. D._ L. Bensinger/ File: BC/t.-4 Mr. L. H. Martin -
Mr. F. R. Coburn- Mr. J. - A. McQueen, Jr./ File: B-X-544 Mr. A. B.-Cutter Mr.-D. O. Myers Dr. T. S. E11eman Mr.-C. H.-Moseley_
Mr. B.-'J. Furr Mr. R.-B. Starkey,LJr.
Dr. J. D. E. Jeffrios :Mr.'L. V. Wagoner--
'Mr. I. A.-Johnson - . Mr. J. - L~.- Willis --
INPO Ms. M.:Si Wingo
- Nuclear Operations-File 13510(E)-
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