ML19323G507

Forwards Repts Re Status of Responses to IE Bulletin 79-14, Seismic Analysis for As-Built Safety-Related Piping Sys, & IE Bulletin 79-02, Attempted Extortion - Low Enriched U.

ML19323G507
Person / Time
Site:	Fort Calhoun
Issue date:	04/21/1980
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
IEB-79-02, IEB-79-14, NUDOCS 8006040024
Download: ML19323G507 (42)

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April 21, 1980 Mr. K. V. Seyfrit, Director U. S. Nuclear Regulatory Comission Office of Inspection and Enforcement Region IV 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76011

Reference:

Socket No. 50-285

Dear Mr. Seyfrit:

This letter is to advise you that the Omaha Public Power District will have completed inspections, analyses, and modifications (as des-cribed in Enclosures 1 and 2) for the Fort Calhoun Station, performed pursuant to the requirements of IE Bulletin 79-14 and IE Bulletin 79-02, prior to unit startup currently scheduled for May,1980. Enclosure 1, attached hereto, provides the District's response to IE Bulletin 79-14, including information in regard to implementation status, methods, criteria, and schedules. A final report will be submitted after com-pletion of the voluntary verification program (see Enclosure 1), which will provide a compilation of required data and sumarize results of the verification program.

Enclosure 2 provides a status sumary of work performed in ac-cordance with IE Bulletin 79-02. A final report sumarizing the dis-position of inspections performed will be submitted in conjunction with the IE Bulletin 79-14 final report.

Sincerely, l, (WP

, W< C. Jones Division Manager Production Operations WCJ/KJM/BJH:jm Enclosures cc: U. S. Nuclear Regulatory Comission Office of Inspection & Enforcement Washington, D. C. 20555 LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.

Washington, D. C. 20036 8006040 O N

ENCLOSURE 1 April 18, 1980 IE BULLETIN 79-14 RESPONSE Request 1 Identify inspection elements to be us'ed in verifying that the seismic analysis input information conforms to the actual configuration of safety-related systems. For each safety-related system, submit a list of design documents, including title, identification number, revision, and date, which were sources of input information for the seismic analyses. Also submit a description of the seismic analysis input information which is contained in each document. Identify systems or portions of systems which are 91anned to be inspected during each sequential inspection identified in Items 2 and 3. Submit all of this information within 30 days of the date of this bulletin. (Includes safety-related piping 2-1/2 inches in diameter and greater per IE ,

Bulletin 79-14, Rev. 1, dated July 18,1979.)

Response

The District responded to this request in a letter dated August 3, 1979. This information is reproduced in Attachment 1 for your reference, except for the list of drawings. The drawings listed below were origi-nally included but were later found to be non-safety-related:

IC-160 Plant Air ~

IC-161 Plant Air IC-159 Plant Air In addition, all drawings previously listed, and discussed on pye 4 of Attachment 1, as unavailable have been located or redrawn. The'.e were on the following systems / components:

Hydrocen Purge (Auxiliary Building)

Fire Protection (Intake Structure)

Miscellaneous Supports -

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• s Requests 2 and 3 For portions of systems which are normally accessible, inspect one system in each set of redundant systems and all non-redundant systems for conformance to the seismic analysis input information set forth in design documents. Include in the inspection: pipe run geometry; sup-port and r' estraint design, locations, function and clearance (including floor and wall penetration); embedments (excluding those covered in IE Bulletin 79-02); pipe attachments and valve and valve operator locations and weights (excluding those covered in IE Bulletin 79-04).

Inspect all other normally acces'ible s safety-related systems and normally inaccessible safety-related systems.

Response

All normally accessible and inaccessible safety related systems have been thoroughly inspected in accordance with the bulletin, in-cluding all revisions and supplements, except for portions of systems included on the following isometric drawings which were located in areas inaccessible both during power operation and cold shutdown:

IC-228 Auxiliary Coolant (AC)

IC-278 (Partial) Chemical & Volume Control (CVCS)

IC-352 CVCS IC-359 (Partial) CVCS IC-365 CVCS IC-248 Waste Disposal (WD)

IC-250 WD IC-315 WD IC-226 (Partial) AC IC-277 AC IC-275 (Partial) AC

. IC-332 (Partial) AC IC-334 (Partial) AC After reviewing each 'of the above noted exceptions, it was con-cluded that the consequence of not performing inspectians presents no significant safety problem or in any way endangers the health and safety of the public. Generally, for the sections of piping which were not accessible for inspection, the following assurances were provided by the safety review. In the event of a breakage of the affected section of piping: l a) dafe shutdown can be achieved.

b) Safe shutdown can be maintained.

c) Equipment and methods required to be used to effect a safe shutdown are evaluated in the FSAR.

d) The break can be detected.

e) The break can be isolated without impairment of the function of essential safety systems.

Response (Continued) f) No safety system required to mitigate any FSAF analyzed accident is affected.

These inaccessible portions of piping are associated with the following systems and/or subsystems: liquid waste disposal (within-waste holdup tank vaults), spent fuel pool cooling, and chemical and volume control (letdown).

The criteria referenced below were used for guidance in performing inspections and to identify potentially significant discrepancies for normally accessible and inaccessible systems:

a) For accessible systems inspected prior to issuance of sup-plement #2 of the bulletin, the criteria used is provided in Attachment 2.

b) For inaccessible and accessible

• systems inspected after issuance of supplement #2 (all items addressed in supplement

1. 2 were incorporated), the criteria is provided in Attachment 3.

In addition to inspecting all safety related systems, portions of certain non-safety related systems were inspected and evaluated for their impact on safety related systems (i.e., pipi :g which could damage nearby safety related equipment in the advent of failure).

• For consistency and accuracy, all previously inspected accessible systems were reinspected using the revised criteria noted above.

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Reques. 4 If nonconformances are identified:

a) Evaluate the effect of the nonconformance upon system oper-ability under specified earthquake loadings and comply with applicable action statements in your technical specifications including prompt reporting.

b) Submit an evaluation of identified nonconformances on the validity of piping and support analyses as described in the Final Safety Analysis Report (FSAR) or other NRC approved documents. Where you determine that reanalysis is necessary, -

submit your schedule for: (i) completing the reanalysis, (ii) comparisons of the results to FSAR or other NRC approved acceptance criteria,' and (iii) submitting descriptions of the results of reanalysis, c) In lieu of b), submit a schedule for correcting nonconformino systems so that they conform to the design documents. Also submit a description of the work required to establish con-formance.

d) Revise documents to reflect the as-built conditions in plant, and describe measures which are in effect which provide as-surance that future modifications of piping systems, including their supports, will be reflected in a timely manner in design

. documents and the seismic analysis.

Response

Philosophy of Evaluation Program Determining system operability and establishing conformance, as referenced in the bulletin, would require a detailed pipe rupture analysis and/or making modifications to return everything to the origi-nal design configuration. It was decided that a more expeditious approach to evaluating nonconformances and establishing system oper-ability would be to analyze "as-L - slt" pipe and support stresses and perform modifications as necessary to ensure that these stresses were within original design stress limits. Modifications made as a result of this approach were not necessarily required for system operability, but it was axpected that less time would be needed to make these modifica-tions than to establish conformance (which might be physically impractical due to construction interferences, etc.) or to determine system oper-ability. In addition, this conservative approach ensures operability of the system by maintaining all calculated induced stresses at less than or equal to conservative allowable design limits.

As'a first step toward analyzing pipe / support system stresses, discrepancies were identified between the "as-built" and "as-designed" configurations of pipe / support systems. Discrepancies are defined as being dimensional or geometric differences (between the as-built and as-designed pipe / support systems) which exceed previously identified acceptance criteria and could potentially affect system operability.

. Philosophy of Evaluation Program (Continued)

After analyzing pipe / support system stresses to account for the reported discrepancies, modifications were made when necessary to bring calculated stresses within appropriate limits. Very few calculated pipe overstress problems were identified (some were due to thermal) and the rajority of support modifications resulted from calculated compressive loads on rod hangers and lateral loads on U-bolts (in excess of catalog rated capacities).

The detailed program which has resulted in establishing system

integrity and thereby ensuring operability is as follows

Implementation Methods The sequence of events, which concluded with assurance of system operability, is. outlined below. See Logic Diagram, Attachment 4.

a) Field inspections were made to determine the "as-built" configurations of all safety related systems. The as-built configurations were shown by marked up design drawings pre-pared by the field teams.

b) Discrepancy reports were generated for all dimensional dif-ferences between the design isometric and the as-built con-figurations judged to have potentially significant impact on the original seismic analysis and/or system operability as follows:

1) For the normally accessible systems, the judgement of system operability was made by members of the Plant Review Committee (PRC). The qualifications of these people were forwarded tc the Commission by letter dated February 19, 1980.

2) In response to.certain NRC staff members' concerns in

, regard to the qualifications of PRC members to make these i judgements, the District has voluntarily elected to have

, all work performed under 1) verified by the District's A/E.

3) For the normally inaccessible systems, all judgements were made by the District's A/E. Their qualifications are provided in Attachment 5.

c) Discrepancy reports were forwarded to qualified piping analysts for review as follows:

1) Where discrepancies were judged to be within the original design criteria, the piping was classified as being in conformance with the original design. The guidelines used by the A/E in evaluating the . impact of discrepancies on piping integrity are provided in Attachment 6.

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Implementation Methods (Continued)

2) Where discrepancies were not within the guidelines identified in 1) above, a dynamic analysis for thermal,

, deadweight and seismic induced stresses using TPIPE code was performed to ascertain pipe stresses.

3) Where pipe stresses were found to be within the original design limits, no changes were necessary in support location. The original design limits are shown in Attachment 7.

4) Where pipe stresses were found to exceed the design allowables, modifications were recommended by the piping analyst and coordinated with field personnel to ensure constructability as proposed.

d) Pipe support discrepancies (other than location which was analyzed in c) above) were evaluated by the District's A/E to determine if support stresses were within original design limits. Criteria used to evaluate support stresses are pro-vided by Attachment 8. Evaluations were performed as follows:

1) Where a new piping analysis was not performed, supports with discrepancies were analyzed using original design loads as taken from the support detail drawings.  ;

' 2) Where a new piping analysis was performed, the supports with discrepancies were analyzed using new TPIPE loads.

In addition, supports having no discrepancies were analyzed when their new loads exceeded their original design loads.

3) Where support stresses were found to be within original design limits, the support was judged to be in conformance.

4) Where support stresses exceeded design limits, modi-fications were recommended which would ensure support integri ty.

e) All modifications required to ensure system operability were made in compliance with applicable limiting conditions for operation in the Technical Specifications.

f) As-built sketches for all field inspected items, which were found to differ from the original design or were modified as a result of analysis, will have been produced prior to unit startup. These sketches are verified by second party review with signatures and dates for documentation.

g) All new and revised documents generated during work perfonned for compliance with the bulletin will be supplied to the District by the A/E at the completion of this task.

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Status of Work All modifications on accessible piping systems have been completed, as required by initial inspection and evaluation. All modifications required in normally inaccessible areas will be completed prior to startup of the Fort Calhoun Station, except for those evaluated in the response to Requests 2 and 3 above. All modifications required as a result of the verification work being performed in accessible areas will be completed prior to unit startup following the 1981 refueling outage.

Any nonconformances (discrepancies) identified in the verification phase

! of inspections in the auxiliary building will be evaluated to determine the effect upon system operability under specified earthquake loadings and applicable action statements in the Fort Calhoun Technical Specifi .

cations, including proper reporting, will be complied with.

Justification for Continued Operation

1) All essential modification work, in compliance with the bulletin, will have been completed prior to unit startup.

" Essential" modifications are defined in Attachment 9. Any modification work resulting from the District's voluntary verification work (referenced in b)2) above) will be performed in accordance with the applicable Technical Specifications.

A system of recordkeeping was established which documents all modifications recommended by the A/E and when they were per-formed. This system assures the District that all work will have been completed.

2) Modifications performed to date have resulted in some up-grading over the original design basis for the Fort Calhoun Station.

3) The Fort Calhoun Station will continue to be operated in conformance with the Final Safety Analysis Report and the Technical Specifications. No unreviewed safety question has been identified with regard to continued operation and the health and safety of the public are not jeopardized in any way by return to power before the final report is submitted.

4) Verification work performed to date on accessible piping systems has shown that no seismic overstressing of piping is present. The District believes the remainder of the normally accessible systems which remain to be verified will show similar results.

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ATTACHMENT 1 August 3, 1979 RESPONSE TO REQUEST IN IE BULLETIN 79-14, DATED JULY 2, 1979 Request

1. Identify inspection elements to be used in verifying that the

, seismic analysis input information conforms to the actual con-figuration of safety-related systems. For each safety-related system, submit a list of design documents, including title, identi-fication number, revision, and date, which were sources of input information for the seismic analyses. Also submit a description of the seismic analysis input information which is contained in each document. Identify systems or portions of systems which are planned to be inspected during each sequential inspection identi-fied in Items 2 and 3. . Submit all of this information within 30 days of the date of this bulletin. -

Response

Inspection Elements The elements to be inspected are the geometry in 2-1/2 inch or 2 larger diameter, safety-related piping plus associated valve and valve operator locations and types; pipe attachments; and seismic restraint configurations, dimensions, embedments, and locations.

Description of Seismic Analysis Input Information Contained in Design Documents and Verifications to be Made The design documents to be checked are the piping and instrument-ation diagrams (P&ID's), piping isometrics with seismic restraint locations identified, and seismic hanger / support drawings. The P&ID's indicate the piping isometric numbers for each segment of the various systems. The isometrics will be verified by measuring the as-built piping geometry to include locations of pipe attachments, valves, and

. seismic hangers / supports. The isometrics give hanger / support identi-i fication numbers for each specified restraint location. The hanger /

support configuration, dimensions, clearances, embedments, and type will be compared with the individual drawings for each one. Embedments will J.

be visually inspected for correct location, type, and general condition.

i The valves will be checked for specified manufacturer and type by com-paring information stamped on the valve body and actuator with that contained in the valve files and on the valve lists. On valves with ia identifying numbers visible, e.g., check valves totally covered by insulation, a cross check between the valve file information and valve lists will be made.

Systems or Portions to be Checked During Each Phase The systems or portion of systems that will be checked during the two phases of the inspection are listed below. -Phase I is for piping which is normally accessible during reactor operation; Phase II is for

August 3, 1979 Systems or Portions to be Checked During Each Phase (Continued) all the rest. Some of the items to be checked are either in very high radiation fields or are inaccessible due to geometry considerations.

Pipe and components in these areas will be viewed from a safe distance, if possible, to confirm proper configuration. In no case will buried piping be inspected.

1. Reactor Coolant System Phase I - None -

Phase II - All accessible piping.

2. Chemical and Volume Control System Phase I - Piping for one charging pump, one boric acid pump, and one concentrated boric acid storage tank. All the re maining non-redundant piping outside the reactor con-tainment building, with the exception of that in very high radiation areas; e.g., piping for the ion exchangers, purification filters, and volume control tank.

Phase II - All the remaining accessible piping.

3. Safety Injection and Containment Spray System Phase I - Piping in the auxiliary building for one high pressure

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safety injection pump, one low pressure safety injection pump, two containment spray pumps, one shutdown cooling heat exchanger, and one recirculation line.

Phase II - All the remaining accessible piping. Containment spray rings and supply pipes will be inspected via field glasses or some other similar scheme.

4. Main Steam Phase I - Piping in auxiliary building up to the containment I isolation valves.

4 Phase II - All the remaining piping.

5. Feedwater/ Auxiliary Feedwater Sy: tem Phase I - Feedwater and steam generator blowdown lines in the auxiliary building between the containment penetration and containment isolation valves. Piping for the l motor-driven auxiliary feedwater pump up to the con- l tainment penetration.

Phase II - All the remaining piping.

August 3, 1979 Systems or Portions to be Checked During Each Phase (Continued)

6. Component Cooling Water System Phase I - Piping for two component cooling water pumps, one control room air conditioner, one shutdown cooling ,

heat exchanger, the spent fuel pool heat exchanger, '

waste evaporator, three component cooling heat ex-changers, cooling water supply and return headers l to safety injection / containment spray pumps, sample heat exchanger, and component cooling water surge  ;

tank. Piping in auxiliary building for the safety  ;

injection tanks leakage coolers and seal coolers / '

lube oil coolers for the reactor coolant pumps. I Phase II - All the remaining piping.

7. Raw Water System l

1 Phase I - Piping for two raw water pumps and three component  :

cooling heat exchangers.

Phase II - All the remaining piping.

8. Spent Fuel Pool Cooling Phase I - All accessible piping. Some restraints in the spent fuel pool are virtually inaccessible; others underwater may be visible from the surface. The storage pool demineralizer and filter are always very high radi-ation areas. Piping in that area will be viewed from a safe distance to confirm configuration.

9. Radioactive Waste Disposal System Phase I - All 2-1/2 inch or larger, nuclear class piping in the auxiliary building. Exceptions include piping around the spent resin storage tad and concentrate tanks because they are very high radiation areas.

Phase II - All remaining 2-1/2 inch or larger, nuclear class piping.

10. Fire Protection System Phase I - All piping within intake structure.

11. Containment Hydrogen Purge Phase I - All piping in auxiliary building.

Phase II - All remaining piping.

. August 3, 1979 Systems or Portions to be Checked During Each Phase (Continued)

12. Plant Air System Phase I - Piping in auxiliary building between the containment penetration and the containment isolation valve.

Phase II - Piping inside containment.

List of Design Drawings A list of design drawings that will be checked is attached.

Drawings that have not been located are listed by title only; identi-fication numbers, revision, and date are left blank. We will continue our attempt to locate these drawings. On the fire protection and con-tainment hydrogen purge systems there are no piping isometrics with restraint locations marked available in the files.

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ATTACHMENT 2 CRITERIA FOR IE BULLETIN 79-14 INSPECTION The following criteria were utilized in determining recordable discrepancies for IE Bulletin 79-14 for normally accessible safety related systems:

1) Restraint location: Restraint locations which deviated from the design documents by >l foot were recorded on the marked-up isometrics; restraint locations which deviated from the design documents by >2 feet were recorded on a discrepancy report form.

2) Valve type and manufacturer: Deviations from the design documents in either type or manufacturer were recorded on a discrepancy report form.

3) Restraint configuration: Restraint dimensions which deviated from the design documents by >15% were recorded on- a discre-pancy report form; restraint material deviations were recorded on a discrepancy report form; restraint orientations which visual inspection found to be different from the design docu-ments were reported on a discrepancy report form.

4) Anchor location: Measurable deviations from the design documents were reported on a discrepancy report form.

5) Piping geometry: Deviations in geometry were recorded on a discrepancy report form; dimensions which deviated from the design documents by >15% were recorded on a discrepancy report form.

(Please note that all discrepancies recorded on discrepancy report forms also appear on the marked-up isometric drawings.)

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ATTAChi4ENT 3 SP-SEISMIC-1-1 4

CRITERIA FOR IE BULLETIN 79-14 INSPECTION The following criteria were utilized in determining discrepancies for IE Bulletin 79-14 for normally inaccessible safety related systems:

1.0 PURPOSE To establish inspection requirements for 21/2" and 1:rger piping and piping support assemblies for safety related piping systems including Seismic ' Category 1 piping.

2.0 References 2.1 NRC's IE Bulletin 79-14, dated July 2, 1979.

2.2 NRC's IE Bulletin 79-14, Rev. 1, dated July 18, 1979.

2.3 NRC's IE Bulletin 79-14, Supplement, dated August 15, 1979.

2.4 Letter

T.E. Short: The District to K.V. Seyfrit, NRC, dated August 3, 1979.

2.5 Fort Calhoun Station, Unit No.1, Technical Specification.

2.6 0maha Public Power District Containment Ccmposite Piping Drawings.

2.7 -

Cmaha Public Power District System P&I Diagrams.

2.8 Dravo Corporation pipe fabrication drawings (piping isometric drawings) that are stamped by Bergen Paterson Pipesupport Corporation which ir. corporate the pipe supports / restraints.

2.9 Bergen Paterson Pipesupport Corporation support / restraint drawings.

3.0 Prerecuisites Sicn Off/Date 3.1 Notify Q.C. at the start of procedure.

Q.C. is to spot check the work to ensure that the work is performed in accordance with procedure.

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Q.C.

3.2 Notify the shift supervisor at start of procedure.

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Q.C. refers to OPPD Q.C. unless otherwise: noted.

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R1 2-1-80 SP-SE!3MIC-1-2 Sion Off/Date 3.3 Obtain proper radiation work permit (RUP) ior work in radiation controlled areas.

RWP No. /

3.4 This work is covered by:

M.0. No. /

4.0 Precautions / Limitations 4.1 Precautions and limitations shall be as per the radiation work permit.

5.0 Equicment Recuired 5.1 Miscellaneous tools required will be flashlights, fifty-foot steel taper and six-foot folding rules.

5.2 Equipment for climbing will be. required such as six-foot step ladders, twenty-five to thirty-foot extension ladders and safety belts.

6.0 Precedure 6.1 General 6.1.1 Support / restraints assemblies include: fixed hangers and supports, variable and constant support hangers and supports, seismic restraints and snubbers. The complete hanger, to-gether with the support steel to the embedments or building steel, and the base plate or embedments are included in the inspection. Base plates and concrete expansion anchor bolts are inspected per "Special Proc'edure S. P. Anchor-1".

6.1.2 Inspection criteria which applie:. to all types of supports /

restraints are listed herein.

6.1. 3 Where visual inspection is necessary due to access limitations ,

this condition shall be recorded on all applicable data sheets.

6.2 Isemetric And Succort/ Restraint Insoection Procedure 6.2.1 Pipe run geometry will be inspected to the piping isometric drawings. Actual dimensions shall be taken and recorded on the piping isometric drawing for any' piping not run as per the piping isometric drawing, within plus or minus (+4") tolerance.

This deviation will also be noted on the Isometric Evaluation Data Sheet.

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S?-SEISMIC-I.-3 6.2.2 Pipe support / restraint attachment locations will be inspected and compared to the piping isometric drawing. The actual dimensions locating the attachment shall be recorded on the appropriate piping isemetric drawings, when the following tolerances are exceeded:

ftominal Pioe Size Location Tolerance

- to 6"

+ 4" 8" to 18" T 8" 20" to 36" T_ 12" the support / restraint shall be recorded on the Isometric Evaluation Data Sheet as beirq out of tolerance.

6.2.3 Valve and valve operators shall be located as shown on piping isometric drawing with a tolerance of plus or minus four

(+4"). Actual dimensions shall be taken and recorded on the appropriate piping iscmetric drawing for each valve and operator not meeting or not shown on the piping isometric drawing.

Valves added or valves indicated on the piping isometric drawing as not having proper information in the valve list shall have as much information as possible taken frcm valve and operator at time of inspection and recorded on the piping isometric drawing. The valve and operator not in conformance shall be noted on the Isometric Evaluation Data Sheet. Valve operator orientation relative to cross section of pipe shall be within plus or minus fifteen degrees (~+15') of orientation shown on the piping isometric drawing.

6.2.4 All additional. supports / restraints that are existing on the pipe, that are not shown on the piping isometric drawing shall be recorded. A sketch showing function and location shall be included on the Isometric Evaluation Data Sheet.

6.2.5 All support / restraint deviations found during the inspection  !

shall be recorded on the System Evaluation Data Sheet. l 6.2.6 The piping system shall be checked for anything that may prevent pipe expansion other than that which is designed to i prevent pipe expansion i.e. other pipes, structure steel or a cable. Any interferences not shown on the piping isometric drawing shall be recorded on the System Evaluation Data Sheet. )

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SF-5EI5Icc-1 h 6.2.7 Pipe floor and wall penetration clearance at four ninety degree (90*) locations shall be checked and recorded on the penetration Data Sheet. Each floor or wall penetration shall have a penetration no. assigned on each piping isometric draiwng which is assigned by the reviewer. The reviewer stall assign each penetration no. in the form X-PY, where X is the two or three digit IC no. and Y is 1, 2, 3, etc. Each floor or wall penetration shall have a Penetration Data Sheet. If pipe is cast integral with the wall, this condition should be recorded on the Penetration Data Sheet as well as the thickness of the wall.

6.2.8 Each support / restraint shall be checked in accordance with the Bergen Paterson support / restraint drawings and deviations recorded on the Support /Rrestraint Data Sheet. Deviation.

recorded shall be but not limited to items such as member sizes and shape, all dimensions, plate sizes,. weld sizes, weld types, weld lengths and bolt sizes.

6.2.9 Base plates with anchors which are shown on the support /re-straint drawings shall be checked as a minimum:

a) Plate thickness, plate dimensions and orientation.

b) Anchor location.

c) Edge distance from center of anchor bolt to edge of plate.

6.2.10 , Concrete Anchors for other supports within a radius of eight inches (8") of any anchor on the support / restraint being inspected shall be recorded on the Support / Restraint Data Sheet.

6.2.11 Attachments to base plates shall be checked dimensionally with the support / restraint drawing. Deviations shall be recorded on the Support / Restraint Data Sheet.

6.2.12 Document deviations as precisely as possible. Use blank sheet and attach, if needed.

6.2.13 When unistrut is used, identify and record size of bolt in-

stalled on Support / Restraint Data Sheet.

U R! L1-80 pg3 1 1980

S?-5ZI2MIC-1-3 6.3 Hancer Acceotance Criteria 6.3.1 Supports / restraints shall be of the correct type that is shown on the support / restraint drawings: rigid rod, structural framed restraint, spring, U-bolt, stanchion, trapeze, snubber assembly, etc. Differences in support / restraint type and function shall be recorded on the Support / Restraint Data Sheet as well as on the System Evaluation Data Sheet.

6.3.2 Ricid Rod Hanaers 6.3.2.1 Single rigid rod hangers can be changed to double rod trapeze hangers provided that the new configuration is adequate to carry the resulting loads.

6.3.2.2 Double rod rigid trapeze hangers can be changed to single rigid rod hangers provided that the diameter of the single rod is one and one-half (1 1/2) times the double rod diameter or greater.

6.3.2.3 Rigid rod hangers shall not deviate from design by more than approximately plus or minus seven degrees (17*).

6.3.3 Cantilever Concrete Attachment 6.3.3.1 The elevation of a cantilever concrete attachment can vary by plus or minus twelve inches (+12") provided the angle of the

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hanger rod (s) does not deviate from design by more than approximately plus or minus seven degrees'(1 70).

6.3.4 Sorina Hancers 6.3.4.1 Spring hangers shall be the correct size that is shown on the support / restraint drawing. The angle of the hanger rod shall not deviate from design by more than app.oximately plus or minus four degrees (140).

6.3.4.2 Types "A", "B" and "C" spring hangers can be laterchanged.

6.3.4.3 The elevation of a Type "A", "B", and "C" spring can be changed.

6.3.5 Constant Supoort Hanaers -

6.3.5.1 Constant support hangers shall be the correct size and type that is shown on the support / restraint drawing. The angle of the hanger used shall not deviate from design by more than approximately plus or minus four degrees (140 ).

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6.3.6 Hydraulic Snubbers 6.3.6.1 Hydraulic snubbers shall be the correct bore and stroke sizes that are shown on the support / restraint drawing. A larger bore snubber may be subsititued.

6.3.6.2 The hydraulic snubber cold set dimension, shown on the support /

restraint drawing, shall be within plus or minus one-half inch (11/2" ) . In all cases, one quarter inch (1/4") from full extension or full compression is required to provide functionability.

6.4 Succort/ Restraint Orientation 6.4.1 The vertical support / restraint orientation shall be within approximately plus or minus seven degrees (+0 7 ) of vertical.

6.4.2 Beam brackets or equal can be rotated ninety degrees (90*) for i rigid rod hanger type supports.

6.4.3 Beam brackets, when used with sw'ay braces, shall be oriented so as not to hinder axial expansion, unless otherwise required on the support / restraint drawing.

! 6.4.4 Restraints installed to act in pairs are to provide the same -

l function as the designed support.

6.5 Welded Pioe Attachments All welded pipe attachments shall conform to the configuration on the support /. restraint drawing. Deviations shall be recorded on Support / Restraint Data Sheet.

6.6 Comconents Substitution

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6.6.1 Comoonents Substitution General a) All components, except spring hangers, constant support hangers, and hydraulic snubber cylinder assemblies, may be substituted provided the size of the substituted material is equal to, or' greater than, the original

, material.

1

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I FEB 1 1980 R1 2-1-20

'.-7

. 5 ? - 3 I I 5: C C b) All ccmponents that are related to the overall adjustment of the support / restraint may be substituted if the substi-tuted material provides overall adjustment.

~

c) All components that are related to the load carrying capacity of the support / restraint shall be installed.

6.6.2 Structural Steel Attachments 6.6.2.1 Two (2) steel plates may be substituted for a beam bracket; dimension should be noted on data sheet.

'6.6.2.2 Hanger attachment plates with forged clevises may be inter-changed with beam brackets with welded eyerads provided the sizes are equal to or greater than that shown on the support /

restraint drawing.

6.6.3 Weldless Eyenuts 6.6.3.1 Weldless eyenuts with threaded rod and welded eyerods may be interchanged.

6.6.4 Pice Insulation Saddles 6.6.4.1 Field fabricated pipe insulaticn saddles may be substituted for vendor supplied pipe insulation saddles; dimension shall be' recorded on Support / Restraint Data Sheet.

6.7 Welding

6. 7.1 ' Welding - Symbol Shcwn 6.7.1.1 Welding support / restraint auxiliary steel to existing steel shall be one quarter inch (1/4") fillet weld all-arcund or shall be welded in accordance with AISC standards as a minimum.

6.7.2.2 Welding of all other support / restraint materials shall be one quarter inch (1/4") fillet weld all-around, or in cases where this is not achieved, the weld must develop, is a minimum, the strength of the weaker of the two members being joined.

? -

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- SP-SEISMIC-1-8 6.8 Insulation 6.3.1 Insulation on piping and valves consist of calcium silicote, reflective, elastimer, and fiberglass. '

6.8.2 Insulation type and thickness shall be noted on the Isometric Evaluation Data Sheet.

6.3.3 Insulation shall be removed from piping as noted by the Engineer.

This includes but is not necessarily limited to the following:

a) Stanchons b) Lugs c) Valves (only where additional information is required) d) Reducers when not visually noticable.

6.8.3.1 Personnel removing and reinstalling insulation will be required to wear at least a half face mask due to the cust and insulation particles that may become airborne. (Some of the old insulation may contain carcinogenic materials.) If the airborne dust and insulation particles become such a problem that it gets into the eyes, either goggles worn with the half face mask or a full face mask may be required.

6.8.4 If insulation is removed, it shall be noted on the Isometric Evaluation Data Sheet, the actual Isometric Drawing, and forwarded to OPPD Q.C.

tiOTE: If the piping is heat traced contact OPPD electrician to move heat tracing.

6.9 Evaluation and Modification 6.9.1 Inspection reports shall be reviewed and appropriate analysis performed.when necessary. The Engineer shall recommend any modificati~ons and/or changes in support configuration or locations. - '

7.0 Maintenance Items The following are items which require a maintenance order (MO) to resolve. These items do not affect the function of the support / restraint as found. This should be recorc'ed in the Comments section of the Support / Restraint Data Sht:et.

7.1 Loose Riqid Rod Hancer 7.2 Locknuts Locknuts inust be installed to a tight condition in accordance with the support / restraint drawing with the following exceptions:

R1 2-1-80 <

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7.2.1 Only one locknut is required at turnbuckles on rigid rod hangers that have eyerods at the steel attachment and at the pipe attachment points. Ths locknut may be either on the top or bottem of the turnbuckle.

7.2.2 Two locknuts are required on any hanger that has a rod threaded at both ends between a clevis or eyenut and a turnbuckle. One locknut can be at the clevis /eyenut end or at the turnbuckle end of the rod. The other locknut shall be on the connection at the other end of the turnbuckle.

7.3 Full Thread Enoacement 7.3.1 " Full hex nut engagement' is defined .:s one full thread exposed beyond the hex nut.

7.3.2 Full thread engagement for clevises and turnbuckles is defined as the extension of one full thread beyond the end of the-clevis / turnbuckle.

7.4 Other items found during inspection which do not affect the function or load capability of the support shall be reported on a maintenance order (MO) for action by the plant staff as necessary. This can be recorded as comments on the Support /

Restraint Data Sheet.

f 1 00DD1h 4-R1 2-1-20 FEB ' t 1980

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LIST SUPPORTS /RESTR AINTS NOS. WITH OEb t ATIONS:

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MISSING SUPPORTS / RESTRAINTS NOS.:

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.OTAL NUMBER OF SUPPORTS / RESTRAINTS:

TOTAL NUMBER OF FLOOR OR WALL PENETR ATIONS:

LIST SUPPORT / RESTRAINT NOS, WITH OEVIATIONS:

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CLEARANCES ADEQUATE EMBEDMENTS SIZE AND LOCATION CHECKED DEVIATICNS:

COMMENTS (NOTE IF VISUALLY INSPECTED)

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LOGIC DIAGRAM (NORMALLY ACCESSIBLE SYSTEMS)

IE BULLETIN 79-14 Comparison with Design Documents ll0 Deviation Deviation l

- 1 l All Other' Deviations: Restraint Location Record on Marked-up Deviation <2 feet Iso's and Discrepancy Report Forms Recordonl Marked-up Isometric for Future y2' Evaluation by GSE g-

~['

l Modify to Comply Forwarld to GSE N,_

With Design Docu- for Analysis -

ments; Schedule -

for Analysis of -

"As-Built" l

Modi 1'y as Accept Required "As-Is" by Analysis -

aliALnMttiI 4

, Page 2 of 2 PROGRAM LOGIC FOR NRC BULLETIN I&E 7914 Y FIELD INSPECTION <

U o ,r O NO 79-02 y FIELD CHECK > AS-BUILT 2 DWG.

O V

k DISCREPANCIES WRITTEN 2

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O g 'I DETERMINE ITS EFFECTS ON 2 EVALU ATION OF ANALYSIS OF FIELD INSPECTION l c DESCREPANCIES BY l 3 STR ESS AN ALYSIS, JACKSON '

E l

2 1r y ANALYSIS

' l 1 m NO YES

'f LOAD TABLES FOR SUPPORT DESIGN r lf CHECK SUPPORTS, CHECK SUNORTS DETERMINE IF WITH DISCREPANCIF.S MODIFICATION IS NEEDED > OK NO jf YES l SUPPORT DESIGN lf FIELD VERIFY Y

MODIFICATION 1f AS 8UILT DWG.

i 1f VERIFICATION I

1f COMPLETE y ATTACHMENT B-2 ,

ATTACHMENT 5 Page 1 of 6 PAUL DAVID OSBORNE Senior Mechanical Engineer Background in mechanical engineering, with concentration in piping and valves for use in nuclear power plants, petroleum and petrochemical process work, 'and mining. Also experienced in inservice inspection for nuclear plants.

~_.: "

EXPERIENCE: Gilbert / Commonwealth since 1977 .

~ 1979 to Present Supervising Inspection Engineer on Omaha Pub!ic Power District's Fort Calhoun, Unit 1. Responsible for technical and administrative

~

supervi-ion of thirty people inspecting the containment building for compliance with the NRC Bulletin IE 79-14. Includes inspection of piping and supports and assuring that seismic analysis meets the design conditions.

1979 Inspection Engineer on Florida Power and Light's Crystal River Power Plant for compliance with NRC Bulletin IE 79-14.

1977-70 Lead Mechanical Engineer - piping and valves on Ohio Edison's Erie Nuclear Plant, Units I and 2. Responsible for all piping, valves and piping specialties, including specifications. procurement, design specifi-cations and design guidelines. Also responsible for client and vendor coordination and bid evaluation, including coordination and implementation of the Inservice Inspection Program (ISI).

._ 1977-74 Stearns-Roger, Inc., Denver, Colorado

' Piping Engineer responsible for piping design and engineering and supervision of designers and drafters on petroleum and petrochemical projects.

Fiping Engineer on field assignment responsible for checking installa-tion of all piping and instrumentation on nuclear power plant. Also prepared schedule for plant start-up procedures.

Piping Engineer charged with design of all types of piping, as well as stress analysis for molyblenum mill site project.

1972-74 Newport News Shipbuilding and Dry Dock Company Design Engineer in the nuclear submarine department with duties of piping and structural design, as well as acting liaison between engi-neering office and actual ship construction.

EDUCATION: B.S., Engineering Technology, University of South Florida,1972 A.S., Engineering and Drafting Technology, Brevard Community College,1970 Additional ASME courses in valve design, nuclear piping, quality I assurance, inservice inspection, maintainability, and research and 4 1 development for nuclear power plants.

SOCIETIES: American Nuclear Society t

! CLEARANCES: Department of Defense " Secret",1973 Atomic Energy Commission " Secret",1973 1/80 Gilbert / Commonwealth

l ATTACHMENT 5 Pagn 2 of 6 f

LEE C. LINDGREN Senior Associate Mechanical Engineer Over six years experience in design and layout, inspection, and drawings verification of

)

piping, ducting. pollution, and emission control systems for new and existing power and commercial projects.

1 EXPERIENCE: Gilbert / Commonwealth since 1977 I 1979 to Present Pipe Support Engineer, responsible for pipe support design and modi- -

1 fication in response to the NRC I & E Bulletin on 455 MW PWR Omaha Public Power District's Fort Calhoun Plant. i 1978-79 Associate Mechanical Engineer, responsible for design and layeut of piping, equipment selection, technical specifications writing, an.1 bid evaluations for Northern States Power Company's Tyrone Energy Park 1160 MW nuclear power plant project (SNUPPS).  !

1977-78 Assistant Mechanical Engineer, responsible for design and system design description writing for site-related systems for Northern States Power Company's Tyrone Energy Park.

1976 77 Teller Environmental Systems, Worcester, Massachusetts Mechanical Designer, responsible for design, layout, and arrangement of equipment, and design and detailing of bag houses and scrubbers.

1974-77 Riley Stoker Corporation, Worcester, Massachusetts - -- - - -

Assistant Systems Engineer, responsible for support design of primary and secondary air duct, cold piping, and miscellaneous small piping for Southern Mississippi Electrical Utilities. Dairyland Power Co-op, and Carolina Power and Light Company projects. Also responsible for development of engineering studies, and design of new and existing equipment. As Draftsman, responsible for design of fuel-burning equipment, foundations, dampers, pipe and duct systems.

1966-73 Unites States Air Force

' Staff Sergent - Weapons Technician, responsible for a load crew trained in loading air munitions of jet aircraft and for maintenance of equipment for U. S. Air Force's Tatical Air Command.

EDUCATION: B.S.E.T., Mechanical Engineer ~ing. Central New England College of Technology,1974 A.E., Mechanica! Engineering, Worcester Jr. College,1965 Additional Coursas:

~ Advanced Strer. A of Materials, Worcester Poly-Technical Institute Environmental Cheinistry, Worcester State College 4/80-Gilbert / Commonwealth -

• n m -, . -,,

ATTACHMENT 5 Page 3 of 6 ANANT D. NAYAKWADI Manager of Piping and Applied Analysis Section Background of fourteen years in mechanical and structural engineering with particular emphasis on stress analysis and design and fabrication of piping systems for nuclear power plants.

EXPERIENCE: Gilbert / Commonwealth since 19'l6, ,

.1980 to Present Manager of Piping and Applied Analysis Section. Responsible ~ for

._- ?C - 7'f" , technical' and administrative supervision of section personnel respon-n ..s- sible for' reliability engineering, nuclear safety analysis, education

, analysis', compressible flow transient analysis, pipe stress and support Design . Analysis. Also es a Project Engineer on TVA SEQUOYAH Unit I and 2 and Watts Bar Unit Number I design basis accident pipe

^* * '~

stress analysis. Responsibility included client coordination, budget control and scheduled completion of the pipe stress efforts.

1978-79 Supervising Piping Engineer on a tivo-unit 1280 MWe (each) nuclear power plant. Responsible technical and administrative supervision of the following mechanical design groups: piping and valve group, pipe stress group, pipe support group, analytical group and mechanical drafting. Reviews and approves design and procurement specifications for piping and valves. Performs bid evaluations and makes recommen-dations to the client. Reviews and approves pipe stress analysis and pipe support designs. Prepares manpower estimates and work schedule

' projections. Reviews plant design for protection against postulated pipe failures in the high energy fluid system. Review and implemen-tation of the NRC Standard Review Plans (SRP) on the nuclear project.

1975-76 Bechtel Power Corporation, Gaithersburg, Maryland Senior Engineer and Assistant Stress Group Leader on Standarized Nuclear Unit Power Plant Systems (SNUPPS). Responsible for super-vising the Stress Group. Performed time-history dynamic anaylsis of main steam system subjected to turbine trip-out forces. Reviewed stress efforts on other piping systems. Also responsible for all Class I Piping Systems Analysis and postulated design basis pipe breaks on high energy systems. Reviewed procurement specifications, pipe stress and pipe support design calculations.

1972-75 Stone and Webster Engineering Corporation, Boston., Massachusetts Piping Engineer on 1250 MW PWR Millstone Unit 3 Nuclear Power Plant Project. Responsible for preparing technical specifications for design, engineering and procurement of piping and valves; review and approval of piping drawings, vendor's isometrics, pipe stress reports; and coordination of pipe break and protection effort ^. Also prepared class I transient Listograms for pipe stress analysis.

As Lead Stress Analyst on VEPCO's North Anna Unit 3 Plant, was responsible for the complete project pipe stress efforts and supervision of 20 stress analysts. Performed combined deadweight thermal, seismic, and fatigue analysis for ASME Code Class I Piping Systems on North Anna, Units 1 and 2.

(Continued) -

, Gilbert / Commonwealth

ATTACHMENT 5 Pagg 4 of 6 ANANT D. NAYAKWADI(Cont'd.)

1 % 8-72 V.J.T. Institute, Bombay, India

, Lecturer in the Mechanical Engineering Department; teaching course in Power Plant Engineering.

1967-68 Indian Plastic Limited, Bombay, India Mechanical Engineer in Chemical Division. Responsible for preventive maintenance of chemical plants, and erection of piping.

1967-68 M.H.S. Polytechnic, Bombay, India Associate Lecturer in Mechanical Engineering Department. Taught Theory and Practice of Engineering Drawings, Heat Engines, and Refrigeration and Air Conditioning to graduate students.

1966-67 Maharastra State Electricity Board, Bombay, India Junior Engineer in the Inspection Wing of Store and Purchase Division.

EDUCATION: B.S., Mechanical Engineering, Bombay University,1966 M.S., Mechanical Engineering, Bombay University,1971 REGISTRATION: Professional Engineer in Ohio (1977), Missouri (1976), Virginia (1974),

Massachusetts (1974) and New York (1979).

SOCIETIES: American Society of Mechanical Engineers Attends ASME Boiler and Pressure Vessel Code, Section 3 meetings of Working Group on Piping and Subgroup on Fabrication and Exami-nation. - -- --

?

4 2/80

ATTACHMENT 5 Page 5 of 6 VINUBHAI SHAH Consulting Mechanical Engineer -

Background of professional experience in mechanical engineering including stress analysis, finite element analysis, heat transfer analysis and fluid transient analysis for numerous power engineering projects. Also specialized in the area of computer programming.

EXPERIENCE: Gilbert / Commonwealth since 1977 1978 to Present: Supervising Piping Engineer responsible for supervision and coordination of pipe stress analysis support design selected to NRC I & E Bulletin i 79-14 for Ft. Calhoun nuclear plant.

1978-79 Senior Mechanical Engineer responsible for the analysis of transients in piping network and component stress analysis using finite element technique. This also included special problems relating to stress analysis

, and development of computer programs. As a Lead Engineer on TVA SEQUOYAH Unit 1 and 2 design Basis Accident pipe stress analysis, responsible for supervision and coordination of stress analysis using

,, T-pipe Computer program. Developed seismic spring criteria for Safety related piping system on the Erie Nuclear power plant.

1917-78 Mechanical Engineer responsible for piping analysis including steam hammer analysis on fossil and nuclear power plants. Clie +-included Ohio Edison Company, Comumers Power Company and Pennsylvania i

Power Company.

1972-77 Stone and Webster Engineering Corporation, Boston, Massachusetts

. Engineer responsible for the maintenance of MARC finite element program including thermal transient analysis, stress analysis and water hammer analysis. Stress analysis of ASME Section III Class 1 and 2 piping system.

1971 72 Brown University, Providene, Rhode Island Research Assistant involved in mechanical and thermal analysis using MARC program. Also involved in finite element analysis and com-puter program development.

1969-70 IJttleton Research and Engineering Corporation, Littleton, Massa-chusetts

, Mechanical Engineer involved in ship structure vibration analysis.

1965-66 Gujaret Electrical Board, India Engineer -Involved in the operation of a fossil power dant.

EDUCATION: B.M.E., Birla Vishvarkarma, Mahavidyalaya,1965 M.S.M.E., Worcester Polytechnic Institute,1969 Additional advanced courses in system programming and finite element analysis.

REGISTRATION: Mechanical Engineerin Massachusetts, (1976)

I (Continued) -

caben/ common =aith

ATTACHMENT 5 Pag 2 6 of 6 VINUBHAI SHAH (Cont'd.)

SOCIETIES: American Society of Mechanical Engineers PUBLICATIONS: Co-author with S.P. Ying, " Transient Pressures in Boiler Steam Lines,"

presented at the Ninth Annual Winter Meeting of the ASME in San Francisco, California, December 1978.

ASME Code Committee:

Attends code committee meeting on piping analysis and support design.

O W

3/80

ATTACHMENT 6 The following guidelines shall be used in evalua' ting the discrepan-cies for computerized pipe stress analysis to satisfy the requirements of NRC IE Bulletin 79-14.

A computerized piping stress analysis using TPIPE code will be performed if as-built piping system does not conform to the original design as follows:

a) The as-built piping is missing supports or supports have been added.

b) The support locations on as-built piping are different than the design locations. Reanalysis is generally required if support location is more than 12 inches away from the original design. An engineering judgement depending on the pipe size shall be used in such cases.

c) The as-built support geometry does not conform to the original design geometry of the support, resulting in a change in the degree of restraints on the piping system.

d) The as-built piping system geometry is different from the original piping geometry used for seismic analysis. It is required to use an engineering judgement to evaluate the -

geometry change in deciding whether reanalysis of the piping system is required.

If reanalysis of the piping system is required, the first analysis should be based on the as-built geometry and support systems of the piping. The pipe stresses of as-built piping should be checked against the allowibles set forth in The Stress Analysis Criteria. If stresses are not within the original design allowables, the piping system shall be reanalyzed adding or deleting supports.

O e

ATTACHt4EtiT 7 PIPE STRESS AtiALYSIS CRITERIA The piping systems listed as Class I in F-1.3 of Appendix F to the FSAR of Fort Calhoun #1 are considered seismic category I systems. These Class I systems are covered by NRC IE Bulletin 79-14.

The Class I piping systems are analyzed for seismic effects and designed / fabricated to USAS B 31.7 Nuclear Power Piping Code.

The primary stress allowables and design loading combinations are given in Table F-1 of Appendix F. The terminology used in this table is as follows:

Design Loading: This is the same as sustained loads defined in the TStiE Section III f1C-3652-1 which include the effects of pressure, weight and other sustained mechanical loads.

Design Earthquake: This term is used to define Operational Basis Earthquake (0BE).

flormal Operating Loadings: This includes the effects of pressure, weight, system operating transients and other sustained mechanical loads. The system operating transient primary effects are such as thrust from relief and safety valve loads from pressure and flow transients, the dynamic effect due to sudden closure of valves in the system.

Maximum Hypothetical Earthquake: This earthquake is known as De-sign Basis Earthquake (DBE). The modern term for DBE is known as Safe Shutdown Earthquake (SSE). 4 Primary Stress Calculation: The pipe stress analysis for primary i stress consideration includes the effects of pressure, weight, other I sustained mechanical loads, system operating transient and earthquake (0BE and DBE).

The allowable stress limits as given in Table F-1 of Appendix F for piping are:

1. OBE and Sustained Loads Design Loading + Design Earthquake = PB + Pm 1 1.2Sh (See Table F-1, FSAR, for definition of terms.)

2. DBE and Sustained Load and System Operating Transients tiormal Operating Loadings + Maximum Hypothetical Earthquake

=PB 1_4SD Cos (w/2 Pm) u SD Pm*30 l

The Sh and SD values should be selected from USAS B 31.7. These Sh values are tabulated in Table A-8 of the code. The SD value which is defined as Design Stress in the Appendix F is equal to Sy at operat-ing temperature for ferritic steels and 1.2Sm for austenitic steel.

The S values are given in Table A-2. y values are given in Table A-3 and Sm Secondary Stress Calculation: The section F.2-2 of :he Appendix F defines that the preliminary thermal analysis was done in accordance with USAS B 31.1. It also says that movements of pipe supports due to seismic or containment post tensioning were considered. There are no secondary stress limits as well as loading conditions due to secondary effects given in the Appendix F. Therefore, the following criteria gbd stress allowable be used for secondary stress calculation in the piping system. ,

SE

• i Me i S A Z

SA = f(1.2SSc + 0.25Sh )

Where SE = Secondary Stress SC = Allowable stress at ambient temperature from Table A-8 of B 31.7.

Sh = Allowable stress as defined under primary stress calcul-ation.

Z = Section modulus of pipe.

ME = Range of resultant moments due to thermal expansion and moment effects of anchor displacement due to earthquake.

! i = Stress intensification factor.

4 If secondary stress criterion as defined above could not be satisfied for certain hot end less flexible system, the following combination of secondary stress plus stresses due to sustained load shall be satisfied.

Pressure stress + dead load and other sustained load stress + expansion stress 1 Sh + SA-Combination of Stresses: The FSAR Appendix F commitment for combining the stresses is as follows:

a. The final combination of stresses is done manually.

b. Stresses are combined at the stress level, rather than moment level as permitted by B 31.7.

However, the reanalysis of the safety related piping systems uses the stress' combination criteria as permitted by B 31.7. The TPIPE com-puter post-processor combines loads at the moment level as required by USAS B 31.7. .

+ 7 e w - -

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Final Stress Summary Report: The following summary sheets shall be completed for the final stress summary report.

4

1. The maximum stress level sheet.

2. Equipment - nozzle, anchor and penetration loads sheets.

3. Support load sheet.

J 5

- , - , ,..w~

'. ,, ATTACHMENT 8 SUPPORT DESIGN CRITERIA

1. CODES & STANDARDS 1.1 USAS B31.7-1969- NUCLEAR POWER PIPING 1.2 ANSI B31.1.0-1967 POWER PIPING 1.3 AISC, SPECIFICATION FOR DESIGN, FABRICATION & ERECTION OF STRUCTURAL STEEL FOR BUILDINGS, 1969 1.4 AWS D1.0-69 CODE FOR WELDING IN BUILDING C0t1STRUCTION 2.- STRUCTURAL STEEL ALLOWABLE STRESSES The follow.ing guidelines will be used in evaluating support steel:

2.1 FSAR, A.PPENDIX F, TABLE F-1, PAGE F-4 2.2 TENSION AISC, " MANUAL OF STEEL CONSTRUCTI0ff , 7th Edition, Section 1.5.1.1, Page 5-16 2.3 SHEAR l AISC, " MANUAL OF STEEL CONSTRUCTION", 7th Edition, Section 1.5.1.2, Page 5-16 1

2.4 COMPRESSION i AISC, " MANUAL OF STEEL CONSTRUCTION", 7th Edition, Section i 1.5.1.3, Page 5-16 2.5 BENDING AISC, " MANUAL OF STEEL CONSTRUCTION", 7th Edition, Section 1.5.1.4, Page 5-17 2.6 AXIAL COMPRESSION AND BENDING AISC, " MANUAL OF STEEL CONSTRUCTION", 7th Edition, Section 1.6.1, Page 5-22 2.7 AXIAL TENSION AND BENDING AISC, " MANUAL OF STEEL CONSTRUCTION", 7th Edition, Section 1.6.2, Page 5-23 3.

COMP 0NENTS STANDARD SUPPORTS ALLOWABLE LOADS The manufacturer's maximum recommended loads will be used as the allowable loads, refer to Bergen-Paterson Catalog No. 66.

4. WELDING ALLOWABLE STRESSES - James F. Lincoln Arc Welding Foundation, " Design of Welded Structpres" by Blodgett, Section 7.1, Page 7.4-6 will be used to determine calculated forces and allowable forces.

r ATTACHMENT 9

.o CRITERIA FOR " ESSENTIAL" MODIFICATI0!!S i

1) All modifications needed to ensure piping integrity based on analytical analysis (support stresses not evaluated).

2) All supports found by " support analysis" to have a Factor of Safety less than 1 based on ASTM yield stress values of materials (structural shapes and bolts).

3) All Unistrut anchors which exceed "Unistrut Catalog Values" (new designs will follow District approved criteria).

4) All snubbers' carrying a load greater than 85% of the relief valve setting as specified in Bergan-paterson hydraulic snubber data.

5) Only supports on safety related piping (and those supports on ncn-safety related piping which effect the 79-14 analysis of safety related pipi;.9) will be modified.

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I . ENCLOSURE 2

SUMMARY

OF WORK RELATED TO IE BULLETIN 79-02 Prior to plant startup, all modifications required because of IE Bulletin 79-02 will be completed in the containment and inaccessible areas of the auxiliary building on the safety related piping 2-1/2 inches in diameter and greater. The scope of the modifications was originally based on the original design loads shown on the Bergen-Paterson support detailed drawings to provide for safety factors of 4 or 5, as applicable. Some modifications have been found to be unnecessary because of decreased loads determined from the IE Bulletin 79-14 analysis.

In addition, all modifications of safety related portions of the small diameter piping (2 inches and under) in the containment and in the inaccessible areas of the auxiliary building will be completed. prior to plant startup. These modifications are based on performance of a visual inspection followed by an engineering evaluation of all those supports that failed the inspection. l Modifica.tions of supports on safety related piping, 2-1/2 inches in diameter and greater, in accessible areas of the auxiliary building will be completed prior to startup to ensure that all anchors have a safety factor of 2 or greater. In addition, approximately 80% of the anchors will have factors of safety of 4 or 5, as applicable. Modification work for base plates / anchors with safety factors greater than 2, but less than 4 or 5, will be integrated. with the IE Bulletin 79-14 verification effort being performed in the auxiliary building (referenced in Enclosure 1). This will ensure that modifications are made which account for the  !

loads as calculated by TPIPE.

Inspection and modification to the accessible safety related portions of piping 2 inches in diameter and less will continue after plant operation. This small piping represents a minor percentage of the total amount of small piping already inspected.

Verification of the calculations for the containment and inaccessible areds in the auxiliary building, based on the original design loads or in some cases the new loads generated by the IE Bulletin 79-14 analysis, will be completed prior to plant startup. Verification of the calcul-ations for accessible areas in the auxiliary building (2-1/2 inch piping and greater), based on the original loads, will continue after plant startup. '

The final report for IE Bulletin 79-02 will provide verification that all seismic support base plates, on the safety related portions of '

systems, will have safety factors in accordance with IE Bulletin 79-02 '

and will include consideration of loads generated by the above referenced IE Bulletin 79-14 verification program.

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