IR 05000369/1987025

Insp Repts 50-369/87-25 & 50-370/87-25 on 870803-07.No Violations Noted.Major Areas Inspected:Licensee Actions on Previos Enforcement Matters

ML20238F735
Person / Time
Site:	Mcguire, McGuire
Issue date:	09/10/1987
From:	Blake J, Chou R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20238F727	List: IR 05000369/1987025 ML20238F724
References
50-369-87-25, 50-370-87-25, NUDOCS 8709160310
Download: ML20238F735 (8)

Text

.- . _

q Se nicog UNITED STATES

.

'o NUCLEAR REGULATORY COMMISSION

~'

[" -

n REGlON 11 101 MARIETT A STRE ET, g j ,

• 't ATL ANTA, GEORGI A 30373 t

o

%q . . . . . s Report Nos.: 50-369/87-25 and 50-370/87-25 Licensee: Duke Power Company 422 South Church Street Charlotte, NC 28242 Docket Nos.: 50-369 and 50-370 License Nos.: NPF-9 and NPF-17 Facility Name: McGuire 1 and 2

) Inspection Conducted: General Office of Duke Power Co., Charlotte, N.C. on August 3 - 7, 1987 Inspector:

R ktk .

C, u

[ u 7'f47 Date Sig ed Approved by: / - 9 47 J. IdKe, Chief 'Dhte' Signed te als and Processes Section iv, sion of Reactor Safety SUMMARY Scope: This routine, announced inspection was in the areas of licensee actions on previous enforcement matter Results: No violations / deviations were identifie DO K O O DH G

_ _ _ _ . _ _ - _ _ _ _ _ _

.

REPORT DETAILS Persons Contacted Licensee _ Employees

• S. B. Hager, Chief Engineer, Civil Engineering Division

• C. L. Ray, Principal Engineer, Civil Engineering Division l

• P. F. Guill, Supervisor, Licensing Engineer R. E. Hardin, Supervising Design Engineer W. H. Taylor, Jr., Supervising Design Engineer J. Cobb, Supervising Design Engineer J. E. Thomas, Senior Engineer

• J. F. Snipes, Design Engineer

• N. G. Atherton, Compliance /NPD Specialist III/McGuire

• R. L. Gill, Licensing Specialist R. L. Williams, Analytical Engineer II D. A. Jenkins, QA Specialist

• P. B. Nardoci, Associate Engineer / Licensing ,

• P. V. North, Assistant Engineer / Licensing )

Other licensee employees contacted included engineers, technicians, and office personne ']

• Attended exit interview Exit' Interview The inspection scope and' findings were sunnarized on August 7,1987, with those persons indicated in paragraph 1 above. The inspector described the ,

areas inspected and discussed in detail the inspection finding No I dissenting comments were received from the license The following new  !

item was identified during this inspectio Unresolved Item (L?T) 369/87-25-01, Loads Used in Support Calculations Differing from Loads ;n Piping Stress Analyses, paragraph 3 The licensee did not identify as proprietary any of the material provided to or reviewed by the inspector during this inspectio . Licensee Action on Previous Enforcement Matters i (Closed) Unresolved Item 369/87-16-04, Records for Improper Structural I Restraints per LOCA Load Mismatc This concern came from the licensee's Significant Deficiency (SD)

Reports 369/80-20 and 370/80-15 which reported the following problems: LOCA loads not used in piping analysis, valve discharge l load used improperly, and support / restraint loads shown on design

_ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _  !

___ _ _ _ - _ _ _ _ _ - _ _

'

y

'

. 2

'

' drawings not matching piping analysis loads. The SD Reports

. estimated. that the total number of support / restraints requiring

. analytical . review and possible revision was 1591 and the number of support / restraints requiring hardware ' change was 450. The inspector who . conducted Inspection 81-03 reviewed . the installed systems required for Mode 6 which were affected by the reanalysis. The matter was left open for NRC review of other example The inspector held discussions with the licensee's responsible engineers and . reviewed the . information provided. ' The license engineers stated that a total 'of - 2632. support /restriants was L reviewed for Unit I for possible revisio Per licensee internal

'

g memoranda to file, SRG-81M-101 and 103, dated February 2 and 3.1981,

.a total numbers of 55 supports required structural changes for Unit 1, which was dramatically .less than the estimated 450 supports for both -

units ' The licensee also performed a similar review for Unit 2. -The inspector randomly selected the following math analyses and supports for Unit 1 to check the loads in stress analysis against the loads shown in the.as-built drawings and loads used in calculation Total Supports Checked Analytical Math Model or List of Supports-Method or Isometric Location Checked Rigorous NC-003 -

31 supports j R1gorous NV-001 -

38 ' supports i j

Al ternate 1-MCR-S-NI-100-6 E MC-AA-5235 l

-

MC-AA-6005 )

Alternate MC-1683-NR.01 -

MC-AA-3328

- MC-AA-3327 l R-13 MC-AA- 844 Among the support loads checked. Support Nos. 1-MCR-NC-530 and 1-MCR-NC-531 for math model NV-001 were identified to have been revised and smaller loads used to qualify the supports than reported in the stress analysi The supports were orignally designed based on the current stress loads required. The subsequent revision based 4 on the incorrect load transmittal generated the mistake. Two sup- i ports are overlap supports on Duke Power Class E pipe (nonsafety-related) which was included on Math Models NC-0S and NV-01 which included safety-related pipin The supports were originally designed for. loads from Math Model NV-01 and later the loads were i revised per the incorrectly revised loads enveloped from NV-01 and 1 NC-09. The licensee immediately reviewed the two supports against the current stress loads and determined that the two supports were adequat The licensee also indicated that the necessary

- _ _ _ _ _ _ _ _

_ _ _ _ _ _ _ - . . _ - _ - _ - _ _ _ _ _ -

t t-i

'

. 3 revisions for support calculations were in progress. This unresolved item 369/87-16-04 is considered closed. Pending licensee resolution on the above two supports, this problem is identified as unresolved item 369/87-25-01, Loads Used in Support Calculations differing from Loads in Piping Stress Analyses, (Closed) Unresolved Item 369/87-16-05, Records for Incorrect Design of Bellows Penetration Structural Restraints and LOCA Restraints on Small Piping Inside Containmen This matter concerns Containment Pressure and Movement (dynamic load) j due to LOCA effects that were not considered in design criteria and piping analyses of lines penetrating containment through bellows  !

type cold penetration These were documented in the licensee's i SD Reports 369/80-21 and 370/80-16 and Supplement 1, dated December 17,  !

198 An evaluation of the piping analysis procedures, the piping stresses, support / restraint locations and loads, and higher loads on penetration structures and attachments was required. Thirty seven (37) piping analysis math models required the licensee reanalysi An estimated 399 support / restraints required revisions and hardware change The inspector held discussions with the licensee's responsible engineers and reviewed the' information provided. Per licensee internal memorandum to file, SRG-81M-0090, dated January 29, 1981, and other information for Unit 1, the licensee reviewed 26 math models for penetration pressure effects and 10 math models for LOCA dynamic j effects with one math model eliminated due to nonsafety-related i piping. The total of 154 support / restraints required structural l changes as a result of the penetration pressure and LOCA dynamic -

review and analysi The licensee also stated that analysis  :

procedures were reviewed to confirm that LOCA pressure loading was f properly included in subsequent piping analysis. Subsequent piping

'

analyses included Unit 2 piping analyse To ensure the licensee effort in resolving this problem, the inspector randomly selected tha following math models and supports to check loads in stress analyses against loads shown c,n detail drawings and loads used in support calculations. If the new loads were less  !

than design loads and displacements were within tolerances, the loads in detail drawings and support calculations might not be revised since the old loads were conservative and the supports could withstand new load .

I

!

i

_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - _ - _ _ -

~

. 4 Total Supports Analytical Math Model Checked or List of Method or Isometric Location Supports Checked Alternate 1-MCR-S-NI-104-2 L- MC-AA-6209 0 MC-AA-6213 Z MC-AA-336-AA MC-AA-337 Alternate 1-MCR-S-VE-100-1 D MC-AA-6151 H MC-AA-6195 I MC-AA-6196 f

Alternate 1-MCR-S-VI-102-1 A MC-AA-793 S MC-AA-6183 Q MC-AA-6216 Rigorus VQ-002 -

15 supports No discrepancies were identified among the above support / restraint Unresolved item 369/87-16-05 is considered close (Closed) Unresolved Item 369/87-16-06, Records for RVLIS Penetration Not Installed to Safety-Related QA nor Traceable Material This item expressed concern with regard to apparently unsatisfactory conditions that had been identified by the license A licensee Quality Assurance (QA) review had revealed that the Reactor Vessel Level Indication System (RVLIS) was not installed under the proper QA program in that tubing connecting to the containment penetration fitting was not ASME Code Material and had no traceability. The above problems were reported on Incident Investigation Report No.1-84-13 and 2-84-13 and Licensee Event Report (LER) No. 84-0 The system drawings for above tubing indicated that the entire system was QA condition one, which required that the tubing materials should be safety-related with traceabilit The inspector held discussion with the licensee's engineers and QA specialist for the above problem The licensee's engineers acknowledged that some information contained in the Incident Investigation Report and other information, previously provided,

'

misled the inspector to believe that the nonsafety-related materials for capillary tubing were ordered and purchased from Westinghouse Electric Company based on Duke Power Company's specification. The actual situation was that Duke Power Company ordered and purchased the entire RVLIS system based on Westinghouse's specificatio Therefore, the nonsafety-related tubing materials were supplied by Westinghouse and installed by Duke based on Westinghouse's specificatio _____ _-

_ -- - .- .. .- . _ _ _ . . _ _ _ _ _ - _ _ _ _ - .

,

'

. 5 Furthermore, the licensee's engineers claimed the material )

requirements specified by Westinghouse utilized the ASME,Section III, NA-1130, Paragraph C exemption. The licensee also stated the entire RVLIS system installed in McGuire Nuclear Station was identical to i one installed in Catawba Nuclear Station. The only difference was that Westinghouse installed the whole system in Catawba Nuclear Station for Duke Power Company. Due to the new information provided, the inspector requested that the licensee provide the following information from Westinghouse Electric Compan (1) The materials for the entire RVLIS system were provided by l Westinghouse based on Westinghouse's specification without the specific materials ordered for capillary tubing from Duke Power Compan ;

(2) The materials used in the RVLIS system were identical, including capillary tubing, between McGuire and Catawb (3) The same RVLIS system were installed in other licensees and ,

approved by NRC's Office of Nuclear Reactor Regulations  ;

(NRR).

On August 20, 1987, the Region II received a telecopy from the licensee who transferred the Westinghouse Letter No. DAP-87-608, dated August 18, 198 The letter stated that all components including all capillary tubing of the Reactor Vessel Level Instrumen-  !

tation System provided by Westinghouse to Duke Power Company for use at the McGuire 1 and 2 and Catawba 1 and 2 Power Plants were standard components and were the same as were provided by Westinghouse for other plant The letter also stated that the generic Westinghouse RVLIS design was presented to the NRC and was found to be acceptabl An additional telecopy from the licensee was received on August 28, l 1987, which forwarded ' Westinghouse letter No. DAP-87-613, dated August 17, 1987. This letter listed other utilities who have used 1 the Westinghouse RVLIS system. The telecopy also forwarded a copy of Generic letter No. 82-28 dated December 20, 1982, which provided NRC approval of Westinghouse's RVLIS system by reference to the NRC ,

review reported in NUREG/CR-2628. Based on the information provided,

,

Unresolved Item 369/87-16-06 is considered close (Closed) Unresolved Item 370/87-16-07, Records for Piping Dynamic

Response Following Loss of Electrical Load Test This matter concerns the licensee's retest of nine hangers found l

deficient in previous dynamic response testing conducted under

'

procedure TP/2/A/1200/27A, Reinspection of Piping Dynamic Response Following Reactor Trip. This testing is required by McGuire FSAR Section 3.9.1.2, and was performed after a 95% or greater reactor trip to verify that the piping system design for the Main Feedwater (CF) and Main Steam (SM) systems take into account dynamic effects due to rapid changes in temperature and pressure. The transient l

'

_ - - _ _ _ _ _ _ _ _ _ _ _ _

- _ _ _ _ _ _ _ - - - - _ . - - . _ _ _ - - _ _ _ _ - - _ - ---

o

', . 6

1. g y

response of the piping is required to be within specified desido !

limits. The test was completed but excluded the support No. 2CF-H44.'. I The inspector had question why that support was excluded from the i test. :The inspector discussed the above concern with the licensge !

engineers and reviewed the information provided. The 11censee'4' {

engineers provided Table 3.2.1-1- and 3.2.2.-3 of the FSAR and DWG- l No. MC-2591-1.0, Rev. 9 and DWG, No. MC-2591-1.1, Rev. 8 for examination of Flow Diagram of Feedwater System (CF). They explained ,

p

'

tables of the FSAR and Flow Diagram which prove that Support i

,

No. 2CF-H44 was not in. safety-related piping system and was exempt j l

from thel test. The saiety-related piping extends from the steam j generator through.and including the outermost containment isolation j valves. Per Table 3.2.1-1 of the FSAR, the supports through the

'

isolations valve and the first support outside the reactor building i

,

require QA ~ and are Category The Design parameters for the !

'

feedwater piping are shown- on flow diagrams, i These flow diagrams clearly define the piping class. Table 3.2.2-3 of the FSAR further defines Duke System Piping Classes. The' piping analysis for the safety piping extends .from the steam generators,

,

l l

through the penetration, through the isolation -' valve -(Safety ,

boundary), and includes piping across the Auxiliary Building roof, !

-into the Turbine Building to the ' A' Feedwater Heaters. Support

~

-

2CF-H44 is on piping from the feedwater pumps to the 'B' feedwater 3 heaters. The piping is Duke Class G, nonsafety piping with design i code of ANSI B 3 !

,

t

'

The originally repeated deformed rigid support was replaced by spring hangers during the last refueling ' outage June 1987. The inspector !

reviewed the installation record and hanger calculation and was . '

satisfied with the licensee explanation and performance. This URI l 370/87-16-07 is considered close Unresolved Items Unresolved Item are matters about which more information is recuired to ;

determine whether they are acceptable or may involve a violations or ;

. deviations. One unresolved item identified during this inspection is  ;

discussed in paragraph !

l

l

' .

.

l

_ _ _ _ _ = _

-rw--,--- - - - - - ,- --- - - - - , ,- -- - , - - - - , - - , - - - - , - - - - - - -

1'

O

'

O =

. ,

m, O 4 i

O O CC Z

m 0-3 N .

. .-

i r

E

)

l

I,

i J

I l

I

!

l

!

i

<

i

,

j

_ _ _ _ _ _ . _ _ . -