ML20246H750

Insp Rept 50-416/89-10 on 890327-31.No Violations or Deviations Noted.Major Areas Inspected:Piping Sys Procedures & Calculations

ML20246H750
Person / Time
Site:	Grand Gulf
Issue date:	04/26/1989
From:	Blake J, Chou R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20246H740	List: ML20246H737 ML20246H750
References
50-416-89-10, NUDOCS 8905160184
Download: ML20246H750 (10)

See also: IR 05000416/1989010

Text

'

. >

l3

i: ,

UNITED STATES -

t. 48V3 EEcu *o

' '

'

' NUCLEAR REGULATORY COMMISSION - )

, .: $v ..

. REGION ll

g

8

l j; 101 MARIETTA STREET.N.W.

'* ATLANTA, GEORGI A 30323

4 ..... /

U

Report No.: 50-416/89-10

Licensee: System Energy Resources, Inc. 1

Jackson, MS -39205 1

Docket.No.: 50-416 License No.: NPF-29 .

Facility Name: Grand Gulf.

Inspectio on te : arch 27-31,-1989

Inspect r:

'

/ Y( f:

,4L R. . D a~t e . igned

Appro ed by ( _s 5/ 16 9

J. . Jake, Chief .Date Signed

Mt ials and Processes Section

n neering Branch

ivision of Reactor Safety

SUMMARY-

Scope: This routine, unannounced inspection was conducted in the area of

review of piping system procedures and calculations.

Results: In the area inspected, no violations or deviations were identified.

One Unresolved Item (UNR) was identified concerning the piping system

calculations, involving five generic problems were identified. The

licensee was very cooperative. They are aggressive and agree to

-resolve the problems- identified and improve their calculation-

quality.

"

8905160184 890428

PDR ADOCK 05000416

Q PDC

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

.

__

.

r

-

7 .

REPORT DETAILS

~1. Persons Contacted

Licensee Employees

• J. D. Bailey, Compliance Coordinator

• L. F. Daughtery, Compliance Supervisor

• N. Deshpande, Pipe Support Design Supervisor - Nuclear Plant Engineering

• W. C. Eiff, Principal Quality Engineer

• C. R. Hutchinson, Station General Manager

D. S. Pace, Nuclear Design Manager

• J. Summers, Compliance Coordinator

• S. Tanner, Quality Services Manager

• F. W. Titus, Nuclear Plant Engineering Director I

'

• M. J. Wright, Plant Support Manager

• J. W. Yelvesta, Plant Operation Manager

Other licensee employees contacted during this inspection included

craftsmen, engineers, mechanics, technicians, and administrative

personnel.

Other Organization l

INPO

• A. W. Davis, INPO Trainee

NRC Resident Inspector

• C. Christensen, Senior Resident Inspector

• Attended exit interview

2. Piping System Calculations Review

a. Description

During an audit of pipe supports design calculations for the

alternate decay heat removal system (ADHRS) on January 4 and 5, 1989,

by NRR, the staff found that one of the calculations presented had

not been revised to bring it up to date. When questioned about this, l

the licensee stated that their procedure does not require a revision  !

be made if there is evidence that the change would result in larger

factors of safety. The staff views this procedure to be a potential

safety problem because the basis for the determination that the

change is in the conservative direction is not documented, reviewed,

or approved to the same QA criteria as the original calculations.

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

.

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

"'

.,

.

, .

,

2

b. Stress Calculations Reviews

The support loadings summary sheets of Stress' Calculation Problems

Nos. 63, -Issue No. 09 for Isometric .No. H-1349A, Rev. M and 141A,

Issue No. 10, for Isometric No. H-1351H, Rev. F were partially

reviewed to check the actual load combinations based on Supplement

No.1, Rev. '0, to . Specification No. 9645-M-300.2,,- Rev.18 against the

Maximum Design ~ Loads listed in those1 summary sheets. All calculated

' actual loads are lower or equal to' the Maximum Design Loads l_isted.

The inspector questioned why the lower loads were- different. . .The

licensee's engineers replied that three cases applied, but none of

the calculated actual. loads is higher than the Maximum Design Loads q

generally used for the pipe support design calculations. Fi rst , the .,

calculated actual' loads were multiplied by 110% for conservatism and.

rounded up to the closet figures. Second, the calculated actual

loads (without adding 10's) were-close to the previous Maximum Design

Loads and the previous Max Design Loads were used. Third,. the

calculated -. actual loads af_ter adding 10% were still lower than the

previous Maximum Design Loads and the previous Maximum Design Loads

were kept since the support calculations were designed using the

previous loads. Since . the calculated actual loads were lower and

were not reflected in the Max Design Loads the support design

engineers were not supposed to review or evaluate the' load changes.

This practice of not revising the Max Design Loads to- reflect the

calculated actual loads will blind the review and documentation of

the pipe support design calculations due to stress load. changes and

is considered a generic problem for the stress calculations.

The individual load' figures shown on the support loadings summary

sheets for Support Nos. S-16, S-9, and'S-10 on the stress calculation

problem No. 63 were checked against the computer printout to verify

the accuracy. All the figures were picked up correctly from the

computer printout,

c. Support Calculation Review

The stress calculations on four isometrics were randomly selected to

compare their Maximum Design Loads, listed in the support loadings

summary sheets, to the design loads used in the support design

calculations. The four isometrics with the corresponding stress

calculation are listed below:

1-

_ - _ - - _ .- _ -__ -___- - - _ _ _ - - - - _ _ _ _ _ - _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ -

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

.

.

'

.

3

1

Table 1

_l l l Stress l

l Isometric lRev l Calculation lRev.

Line Description l No. lNo. I No. lNo.

I I I I

HPCS PP Discharge to Containment l HL-1349A 1161 63 l 9

l -1336C l l l

1 -1336L l l l

1 I I I

RHR, LPCI "A" & "B" & CTMT Spray I HL-1348F l 15 l 141-A. l 13

l l l 1

Main Steam Line IFSK-P-1013M.001-Cl 4l l

1 I I I

Standby Surface Water Loops l H-1358K l 7l 170 1 5

l H-1358L l 8l l

A total of about 80 support design calculations from the above four

stress calculations were randomly selected and the design loads used

checked against the Maximum Design Loads listed in Stress

Calculations. The support calculations for stress calculation No. 63

were also reviewed and evaluated in detail.

'

(1) Stress Calculation No. 63

This stress calculation contains 18 support calculations. I

Seventeen support calculations were mainly reviewed for

comparison of loads between the latest stress calculations and

the latest support design calculations. One support calculation

was not available for comparison. All support calculations were

completely or partially reviewed and evaluated for member sizes,

weld sizes, anchor bolts, base plates, standard components,.

thoroughness, clarity, consistency, and accuracy. In general,

the design calculations were of good quality. The calculation

package contained cover sheet, computer input, computer output,

and verification of critical members, deflection, member

flexibility, welds, anchor bolts, and base plates. The

discrepancies or comments are listed in Table 2. More than 90%

of the standard components, such as struts and clamps were not

verified in the support calculations for their actual loads

against the manufacturer's load capacity sheets or catalogs.

l

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

, _ _ _ __. _ ____ _ . _ _ _

4

.

- -

p ..-

4

l

l

i

i.

'

Table 2

Support l 1 l

Calculation l Rev. l Stress l

No. I No. I No. I Discrepancy / comment

! I l  !

Q1E22G001C08 l B l S-1 l Note 1

l l 1

01E22G001C01 l G l S-2 l Note 1

l l l

Unknown l l S-3 l Calculation was not received.

-l I l

Q1E22G001C03 l C l S-4 I Clamp was not verified in i

i I l calculation.

I I i  !'

Q1E22G001R03 I C l S-5 l Note 1

I I I

Q1E22G001C04 l B l S-6 l The resolutions of forces on

l l l p.5 were wrong. Therefore ,

l l l the computer input and inter-

I l l actions of bolts were wrong.

I I I

l l l Note 1

I I I

Q1E22G001R02 l D l S-7 l DCN #2 was not incorporated  ;

.i l l in calculation. '

l l 1

l l l Note 1

I I I

Q1E22G001C02 l D l S-8 l Note 1 ,

d

I I I

l l l Reaction on Joint 1 on p. 10 ,

l l l of computer output should be

l l l bearing or compression ,

l l l only. The wrong computer

l l l model produced the wrong

l l l output. <

l 1 I

Q1E22G001C05 l 8 i S-9 l Note 1

I l I

Q1E22G001C06 l C l S-10 l Note 1

I I I

Q1E22G001C07 l D l S-11 l None j

l l l

Q1E22G001R04 l C l S-12 l Note 1

l l 1 )

__ _ _ _ _ _ _ _ - . _ __-_______________ _____ _ _ ___ - -

_ _ - _ _ _ _ _

'

.

.

-

.

5

Table 2

Support l l l

Calculation l Rev. I Stress l

No. I No. I No. I Discrepancy / comment

(cont'd)

Q1E22G001R05 I B l S-12 i Note 1

<

I I I

Q1E22G001R06 l C l S-13 l Note 1

l l l

Q1E22G001R07 l A l S-14 l Note 1

l l l

Q1E22G001H01 l B l S-14 l None

1 l 1

Q1E22G001H04 l A l S-15 l None

i I l

Q1E22G001H05 I B l S-16 l Note 1

l l l

l l l The critical weld between

l l l bracket of Items 3 and 2

l l l (w4x13)was not reviewed.

Note

1. Sway strut and clamp were not verified in support

calculation.

(2) Stress Calculation No. 141-A

This stress calculation containing about 42 support calculations

15 support calculations were randomly selected and reviewed for

comparison of loads between the latest stress calculation and

the latest support design calculations. Table 3 lists the

discrepancies or comments. The standard components such as

spring cans, sway struts, clamps, etc. were also not verified in

support calculations for their actual loads against the

allowable loads stated in manufacturer's load capacity sheets or

catalog. Seven out of 15 support design calculations had the

latest lower stress loads than the latest support design loads.

The support design calculations were not reviewed for the latest

lower stress loads and documented the evaluation.

.

. l

.

.

6

Table 3

Support l l l

Calculation l Rev. l Stress l

No. l No. I No. l Discrepancy / comments

l I l

Q1E22G015C04 l D l S-12 l None

i I I

Q1E12G015R31 l D l S-37 l None

1 l 1

Q1E12G015C06 l C l S-13 l None

l I l

Q1E12G015C08 i B l S-5 l Note 1

l l 1

Q1E12G015H07 l B l S-29 l Note 1

I I I

Q1E12G015H08 i B l S-30 i Note 1

l l 1

Q1E12G015H09 l B l S-32 l None

l l l

Q1E12G015H06 & H10 l B l S-28 l Note 1

i l I

Q1E12G015H11 l B l S-4 l Note 1

l I l

Q1E12G015H14 l B i S-6 l Note 1

I I I

Q1E12G015H15 l 0 l S-33 l None

I l l

Q1E12G015H17 l C l S-37 l None

I l l

Q1E12G015H19 l G l S-7 l None

l l l

Q1E12G015H2O I D l S-2 l None

l I l

Q1E12G015R10 l B l S-6 l Note 1

l l 1

Q1E12G015R13 l D l S-14 l None

Note:

1. The latest stress loads were lower than the support

design loads and the effects of the lower loads on the

support design calculations were not reviewed and

documented.

(3) Main Steam Line

Both the Main Steam Line and the Recirculation, Line do not have

support design calculations. All supports an both lines are

standard components such as snubbers or springs. Therefore, the

i

__ _ -_ __

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

• i

.

.

i

7

l

licensee generally did not create support design calculations to

'

verify the actual loads against the component allowable loads

set by the manufacturer's load capacity sheets or catalog.

Twenty-four snubber loads shown on Drawing i

No. FSK-P-1013M.001-C, Rev. 4 were compared to MPL No. B21-G006, i

Installation and Instruction Manual for Main Steam System,

Rev. A, Grand Gulf I and 2 Montek Division of E-Systems, dated

November 1980. The Manual illustrated parts listing, snubber,

and accessories. The snubber loads shown on drawing were

correct. The snubber loads were also compared to the wmputer

printout of stress analysis performed by General Electric

Company and all loads were correct.

Ten variable springs (or supports) shown on the above drawing

were also compared to MPL B21-G002, ITT Grinnell Corporation I

Installation Manual for Grand Gulf I & II and the attachment,

MSS SP-89, Pipe Hangers and Supports Fabrication and

Installation Practices, by Manufacturers Standardization

Society, Inc., 1978 Edition. The spring loads showed on drawing

were correct.

(4) Stress Calculation No. 170 j

Fourteen out of 35 support calculations were randomly selected

and reviewed for comparison of loads between the latest stress

calculation and the latest support design calculations.

Calculation Nos. Q1P41G010R01, Rev. A for Stress No. S-3 and

Q1P41G010C01, Rev. A for Stress No. S-4 were found to have

latest stress loads lower than the support design loads and the

support design calculations had at been reviewed and the

evaluation documented.

d. Procedures and Specifications Review

The following Nuclear Plant Engineering Administrative Procedures and

Technical Specifications were reviewed during this inspection:

(1) Procedure No. 305, Engineering calculations, Rev. 11

(2) Procedure No. 307, Engineering Drawing

(3) Procedure No. 315, Updating /As-Building Grand Gulf Nuclear

p Station Design Documents, Rev. 7

(4) Procedure No. 323, Design Inputs, Rev. 1 l

(5) Specification No. 9645-C-103.1, Design and Installation of

Concrete Expansion Anchors, Rev. 9

__-_-_-_ _ --____----_______ -

_ _ _ _ - . _ _

___ _ _ - - _ _ _ _ ________ _ _ _ _ _ _

,

. I

"

w. <

t

. ['

.

i

'(6) Specification ' No. 9645-M-300.2, Design Specification for Pipe

Hangers, Supports, Restraints,-and Anchors, Rev. 18

.

1

'(7) Specification No._9645-M-220.0, Design Specification for Nuclear  ;

Piping Systems, Rev. 15  ;

None of above procedures or specifications required. action, when. the

. latest stress loads were lowe- than the previous stress. loads.or the

design -loads. The licensee's engineers replied that common sense

required them to review and revise. the support design calculation g

when the latest stress loads were higher than the previous stress

loads or design loads, but for the stress loads. lower than the.

previous stress . loads or design loads, the design engineers thought

'that the safety factor would be higher, and while they did' review the'. .

design calculations they did not document the evaluation since. no 4

'

procedures required them to do so, _ There were no evidence.in the

i

design calculations to prove that the design calculations were

reviewed to compare the latest lower stress loads against the higher  ;

1' design.. capacity.

e. Findings and Results

After reviewing the . stress and support calculations, five potential- )

generic problems were identified: R

(1) The latest stress loads were lower than the previous stress

loads, the stress calculations were not revised. The. problem is ,

that the support design engineers do not know the new l stress  !

loads are lower- and do not review the support . design. j

calculation; and document the review evaluation. 4

(2) The latest stress loads on the Maximum Design Loads were . lower

than the previous stress loads or support design loads. The j

support design calculations were not reviewed or their ' i

evaluation documented to determine the safety significance. l

(3) Base plate flexibility was not considered in anchor bolt

qualifications per requirements of IE Bulletin 79-02. The

licensee stated that the premise that base plate flexibility .

'

could be neglected in support design calculations was approved

by NRR, but no ' approval letter was available. The NRR l

inspectors also found out this problem during their second

inspection on March 20-24, 1989. The NRR may take care of this

problem by inspecting the licensee A/E firm-Bechtel power i

Corporation.  ;

(4) Standard Components such as snubbers, sway struts, clamps,  !

springs, etc. were not verified in nearly all of the support ,

design calculations. A few of the support design calculations l

did verify the standard componer,ts.

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

_ _ - _ _ _ _ _

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

..

.

.

9

(5) Main Steam Line and Recirculation Line do not have the support

design calculations.

The licensee acknowledged the above problems and agreed to solve them

by reviewing and regenerating calculations. The procedures or

specifications will be revised to add the procedures for reviewing

and documenting the evaluation in support design calculations when

the stress calculations are revised, even when the new loads are

lower than the previous design loads. The licensee did take quick

action and issued " Pipe Support Load Reconciliation Guidelines" on

March 29, 1989, during the inspection, to provide instructions for

the review of pipe support calculation design loads - and the

corresponding piping stress load sheets. This review will include

all "Q" designated pipe supports. Pending the licensee's action to

resolve the above five potential generic problems and discrepancies

found on Table 2, this item is identified as UNR 50-416/89-10-01,

Piping System Calculation Concerns.

3. Exit Interview

The inspection scope and results were summarized on March 31, 1989, with

those persons indicated in paragraph 1. The inspector described the areas

inspected and discussed in detail the inspection results listed below.

Proprietary information is not contained in this report. Dissenting

comments were not received from the licensee. The inspector expressed

concerns about the five generic problems identified during this

inspection. The management said they would solve the problems in a proper

way by reviewing, revising and regenerating the calculations and

procedures.

UNR 50-416/89-10-01, Piping System Calculation Concerns.

!

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