ML20210N950

Insp Rept 50-346/86-04 on 860114-0402.Violations Noted: Failure of Bechtel to Rept Noncompliance W/Fsar Conditions Re Stress Calculations to Util & Insp & Evaluation of Piping & Supports Not Conducted

ML20210N950
Person / Time
Site:	Davis Besse
Issue date:	04/30/1986
From:	Danielson D, Fair J, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20210N927	List: IR 05000346/1986004 ML20210N923 ML20210N934 ML20210N950
References
50-346-86-04, 50-346-86-4, CAL-85-13, IEB-79-14, NUDOCS 8605050334
Download: ML20210N950 (18)

See also: IR 05000346/1986004

Text

-____ _ _ _ _ - _______ __

. .

.

.

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-346/86004(DRS)

Docket No. 50-346 License No. NPF-3

Licensee: Toledo Edison Company

Edison Plaza

300 Madison Avenue

Toledo, OH 43652

Facility Name: Davis-Besse Nuclear Power Station, Unit 1

Inspection At: Dasis-Besse Site, Oak Harbor, OH

Bechtel Power Corporation, Gaithersburg, MD (Bechtel)

Inspection Conducted: January 14-16, February 5-6 and 19-20, 1986 at the site

January 29-30 and April 1-2, 1986 at Bechtel

Inspectors- . T. Yin oh

W

Da~te

(April 1, 1986 only)

duin

Date

Approved By:

e-

D. H. Danielson, Chief d[3o[/t

Materials and Processes Section Date

Inspection Summary

Inspection on January 14 through April 2, 1986 (Report No. 50-356/86004(DRS))

Areas Inspected: Special, announced inspection of the auxiliary feedwater pump

turbine steam supply (AFPTSS) piping modifications; the Facility Change Request

(FCR) system; the implementation of Region III (RIII) Confirmatory Action Letter

(CAL) 85-13 actions; actions on Licensee Event Reports (LER); the status of

completion of IE Bulletin (IEB) 79-14; the Bechtel control of High Energy Line

Break (HELB) analyses; and followup on previous inspection findings.

Results: Of the areas inspected, two violations were identified; (failure

of the Bechtel staff to follow procedures and failure of TED to follow site

procedures - Paragraphs 4.b and 8.b(1); failure of the licensee to take

adequate corrective action on identified problems - Paragraph 8.b(2)).

WonBM M8@r

O

_ . .

. .

.

.

DETAILS

1. Persons Contacted

Toledo Edison Company (TED)

• +*T. J. Bloom, Senior Licensing Specialist

• T. Chowdharm, Manager, Engineering Services Department

• +*P. H. Straube, Senior Engineer

J. Dunne, Senior Engineer

T. J. Myers, Nuclear Safety and Licensing Director

• L. Ramset, Quality Assurance Director

F. R. Miller, Staff Engineer

P. W. Jacobsen, Senior Engineer

J. F. Helle, Nuclear Facility Engineering Director

• S. J. Osting, Senior Assistant Engineer

D. R. Wyokko, Regulatory Affairs Supervisor

• D. Kies, Manager, Mechanical / Structural Engineering

• H. Brinkmann, Director, Nuclear Facility Engineering

C. Merkbel, Civil and Structural Systems Engineer

Bechtel Associates Professional Corporation, Ohio (Bechtel)

J. W. Brothers, Chief, Quality Engineering

N. Tolani, Senior Engineer

+M. S. Wasserman, Mechanical Engineer Supervisor

• M. L. Murphy, Senior Engineer

• W. C. Lowery, Project QA Engineer

A. T. Vieira, Engineering Technical Specialist

• +D. C. Kansal, Deputy Division QA Manager '

J. M. Ogle, Civil Engineer Supervisor

• +D. L. Gill, Project Quality Engineer

+E. J. Ray, Project Engineer

C. H. Abutaa, Senior Engineer

R. Lee, Engineer Supervisor

+T. I. Gillespie, QA Manager, Projects

+S. R. Kalavar, QA Manager, Audit

S. A. Bernsen, Division Manager of QA

J. B. Wallis, Senior Engineer

• V. R. Marathe, Assistant Project Engineer

, K. I. Patel, Engineering Supervisor

U.S. Nuclear Regulatory Commission, Region Ill (RIII)

• W. Rogers, Senior Resident Inspector

*D. Kosloff, Resident Inspector

+ Denotes those attending the management exit meeting on January 30, 1986

. at Bechtel.

• Denotes those attending the management exit meeting on February 20, 1986

at the site.

!

j

!

2

<

. .

.

.

• Denotes those attending the management exit meeting on April 2,1986 at

Bechtel.

2. Licensee Action on Previous Inspection Findings

a. (Closed) Unresolved Item (346/83-17-04): Several new vintage

Grinnell Corporation hydraulic snubbers with Miller cylinders,

including PSP-1-H4 and PSP-1-H6 installed on the Pressurizer Spray

Piping System, were observed installed without fluid reservoir

breather and filter units. The NRC inspector reviewed the site

Temporary Modification Request, dated January 10, 1986, and

Section 8.2.5 of Procedure MP1410.02.04, " Maintenance of Hydraulic

Snubbers," and considered the licensee's measures for reinstalling

the filter units to be acceptable. A purchase order procuring

50 new filter units was issued on January 10, 1986.

b. (Closed) Violation (346/85013-01): The licensee failed to document

nonconformances in accordance with procedure requirements. The NRC

inspector reviewed Item IV.A.2 of the TED response letter (Serial

No.1-604) to the NRC, dated January 27, 1986, and considered it

acceptable. TED corrective actions are documented in RIII Inspection

Reports No. 50-346/85013, Paragraph 8; No. 50-346/85033, Paragraphs 2

through 5; No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this

report.

c. (Closed) Violation (346/85013-02): After the AFPTSS problems were

identified, the TED evaluations did not investigate the cause of the

problem and consequently measures to prevent recurrence were not

developed. The NRC inspector reviewed Item IV.A.3 of the TED response

letter (Serial No. 1-604) to the NRC, dated January 27, 1986, and

considered it acceptable. TED corrective actions are documented in

RIII Inspection Reports No. 50-346/85013, Paragraph 10;

No. 50-346/85035, Paragraphs 4.a and 6; and Paragraph 6 of this

report.

d. (Closed) Violation (346/85013-03): Inadequate piping suspension

system QC inspection and ineffective implementation of the IEB 79-14

! walkdown inspection program. The NRC inspector reviewed Item IV.A.1

l of the TED response letter (Serial No. 1-604) to the NRC, dated

{ January 27, 1986, and considered it acceptable. TED corrective

i actions are documented in RIII Inspection Reports No. 50-346/85033,

i_ Paragraphs 2 to 7; No. 50-346/85035, Paragraphs 4.b and 5; and

L Paragraph 4 of this report.

!

! e. (Closed) Violation (346/85013-06): The licensee failed to report

l AFPTSS component deficiencies in accordance with 10 CFR 50.73

l requirements. The NRC inspector reviewed Item IV.B of the TED

l' response letter (Serial No. 1-604) to the NRC, dated January 27,

1986, and considered it acceptable. TED corrective actions are

documented in RIII Inspection Reports No. 50-346/85033, Paragraph 2;

l No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this report.

!

,

3

L

'

. .

.

.

,

f. (Closed) Violation (346/85031-01): Lack of a design interface

procedure between Bechtel and Grinnell for evaluating pipe hangers

in accordance with IEB 79-02 and IEB 79-14. The NRC inspector

reviewed TED response letters, Serial No. 1-593, dated November 25,

1985, and Serial No. 1-616, dated February 27, 1986, and considered

the matter resolved. The need to review Grinnell calculations is

discussed in Paragraph 10 of this repo:t.

g. (Closed) Violation (346/85031-02): TED did not effectively implement

its FCR system in that a number of safety-related supports were not

restored to their FSAR condition in a timely manner. The NRC

inspector reviewed TED response letters, Serial No. 1-593, dated

November 25, 1985, and Serial No. 1-616, dated February 27, 1986,

and considered the licensee actions to be acceptable. The near term

support modification work to assure FSAR conditions were met, was

conducted in accordance with RIII CAL 85-13, Item 1.a(4)

requirements (see Paragraph'4.d of this report). The licensee's

long term upgrade of the FCR program are being reviewed by RIII and

NRC Headquarters personnel. See Paragraph 2.k of this report for

details concerning piping design and support modifications.

h. (Closed)Viofation(346/85035-02): The licensee failed to use the

appropriate allowable stresses specified in Bechtel Evaluation

Procedure MGP-04 for evaluating stresses at weld attachments to the

piping pressure boundary. The NRC inspector reviewed the TED

response letter, Serial No.1-314, and Bechtel Procedure CGP-04,

" Procedure for Evaluating Nonconformance Reports Related to Pipe

Supports, Pipe Anchors, and Seismic Restraints at Davis Besse

Nuclear Power Station, Unit 1," Revision 1, dated January 28, 1986.

The NRC-inspector noted that Procedure CGP-04 allowed higher

allowable stresses for the SSE load combination than Bechtel

Procedure MGP-04. Bechtel reanalyzed all affected piping using

the revised allowable stresses. The results were documented in

a Bechtel letter to TED, BT-16555, " Procedure GCP-04; Faulted

Condition," dated April 2, 1986. Procedure CGP-04 was subsequently

revised to reflect the lower allowable stresses for the SSE load

combination on April 4, 1986 as Revision 2.

i. (0 pen) Unresolved Item (346/85035-03): -The NCR evaluations for weld

deficiencies designed to the AISC specification do not require meeting

the specification minimum weld sizes which correspond to the base

material thicknesses. The NRC inspector reviewed the licensee's

response contained in an intracompany memorandura (File 0093, T-0294)

dated January 24, 1986. The NRC inspector will discuss this matter

with NRC-NRR to determine if it represents a position acceptable to

the NRC staff.

! J. (Closed) Unresolved Item (346/85035-04): Bechtel exhibited

questionable design control for conducting HELB analyses and whip

restraint designs. See Paragraph 7 for details of the followup

'

inspection.

"

.

4

i

-.

1

. .

.

.

k. (0 pen) Open Item (346/85035-05): The TED MWO and FCR systems require  :

further evaluation and improvement. A temporary organization named

Engineering Services Department (ESD) was formed to focus the

company's attention on closing out open FCRs and to establish better

ways of handling future FCRs. ESD is presently manned by a full time

technical staff of ten and occasionally by engineers of various

disciplines depending on specific needs. The NRC inspector met with

the Manager of ESD and reviewed the ESD organization chart and the

" Project Proposal for Closeout Backlog Evaluation" to evaluate the

scope and provisions of the project and had no adverse comments.

A more detailed review of ESD will be conducted by the NRC inspector '

to assess the effectiveness of the new system. In addition to the

above effort, an audit was conducted by Stone and Webster Engineering

Corporation of the FCR system to identify system deficiencies and to

recommend improvements.

'

3. Licensee Action on Licensee Event Reports (LERs)

a. (Closed) LER (346/85019-LL): "PORV Discharge Line Overstressed Due to

Inadequate Heat Trace," reported on November 6, 1985. See Paragraph 8

for inspection details,

b. (Closed) LER (346/85023-LL): " Error in the High Energy Line Break

Analysis in the Auxiliary Building," reported on December 28, 1985.

During a TED review of the environmental qualification (EQ) and single

failure analysis for a proposed modification to the AFPTSS piping,

TED discovered that portions of the system upstream of the MS

admission valves 106, 106A, 107 and 107A would not be isolated during

a postulated high energy line break event. A break in these pipe

'

sections would affect rooms 500, 501 and other connected rooms. The

licensee's corrective actions are included in RIII Inspection Report

No. 50-346/85035, Paragraph 6.c.

The NRC inspector noted that similar situations could exist in other

safety-related high energy piping systems. TED stated that efforts

to expand the scope of the HELB review had been initiated, and that

deficiencies had been discovered in the main feedwater lines. TED

also indicated that nonconforming conditions will be documented in

either an amendment to LER 85023 or in a new LER.

4. Implementation of RIII CAL 85-13_[ction Items

As a result of a meeting con @;cte at the site on October 9, 1985 (RIII

Inspection Report No. 50 Af ' 150 , Paragraph 4) RIII CAL 85-13 was

issued on October 17, 19t5

The licensee's implementation of the actions set forth in the CAL was

reviewed by the NRC inspector. The status of CAL Item 1 (action items

prior to plant restart) is as follows:

J

e

5

1

- - - . , - , . - ~ , , - , , - , . . , - , , - , _ , . - - - - - - - - - - - , , - - . - . - - - - , - - . - , - . . , - - , ,. -

. .

,

.

I . .

, ] a. Item 1.a(1)-(Closed)

,

'

,

s Reinspections per_ CAL 85-13 Item 1.b

. ,

• All 2365 hangers have been inspected; 921 require evaluation and

,f, , clospout

... ,3; prior to restart.

-IStatus of Engineering Evaluations as of March 25, 1986

• 858 NCRs', were writteE ,af ter evaluation of the 921 inspected

hangers

s .

• ,Of the 858 NCRs, 656 were dispositioned to "Use-As-Is"

• Of the 858 NCRs,202 required corrective actions; rework has

been completed for 180 of the NCRs

,- s

b. Item 1.a(2) (Closed)

i

The following piping stress analyses were rerun by Bechtel to

determine system operability:

Systems Affected Bechtel Calculation Nos.

'

LPI 188, 180, 80A, and T-010A

LPI/ Core Flood T-008

l / HPI 56D, 56F, and T-009B

'

l Containment Spray 22F

.. Containment Sump to ECCS 32E

' Hydrogen Dilution 119L

q MS-Exhaust from AFPT 161

The NRC inspector selected the following calculations for review at

Bechtel and concluded that the evaluations were technically sound

and conservative:

• No. 56F, " Davis-Besse High Pressure Injection System,"

Revision C2, dated December-20, 1985.

. .:-

• No. 188, " Davis-Besse Low Pressure Injection System,"

Revision C3, dated January 3, 1986.

In Calculation No. 18B there were two restraints that did not meet

FSAR commitments. They were not reported to TED in accordance with

Bechtel Procedure MGP-04, " Procedure for Control of Interim /Short-Term

Allowable Stress Criteria for Seismic Category I Piping Systems at

Davis-Besse Nuclear Power Station Unit 1," Revision 1, dated

September 27, 1985. Procedure MGP-04, Paragraph 6.1.a and 6.1.b

states:

"For each nonconformance determined by calculation to require the use

of interim stress criteria per Section 4.0, TED Facility

Engineering shall be notified as follows:

6

l

. .

.

.

(1) Bechtel shall notify by telephone the Director, Nuclear

Facility Engineering, or General Supervisor, Facility

Engineering.

(2) The notification by telephone shall be followed up by a

written confirmation. The written confirmation shall

include the NCR number, specific deviation evaluated,

and the specific analyzed results. Modifications required

to meet design /SAR requirements will be included. The

written confirmation shall also include a comparison of

the interim versus SAR allowables involved with the design

and the available safety margin."

During the inspection, the NRC inspector noted that the following

systems and pipe restraints required the use of interim stress

criteria (exceeded FSAR allowables):

System Stress Calculation No. Restraint No.

Containment Spray 22F(C2) -

LPI T-010A(C6) -

LPI 18B (C3) * GCB-1-H13

• A-59

Hydrogen Dilution 119L 29-HBB-15-H7

Containment Sump

to ECCS 32E * 33A-GCB-8-H3

• A-43

The failure to follow the approved procedure is a violation of

10 CFR 50, Appendix B, Criterion V (346/86004-01A).

Prior to the conclusion of the NRC inspection, the licensee

initiated a number of corrective actions which are included in

the following documents:

• TED letter to Bechtel, "NCR Resolutions," advanced copy dated-

February 11, 1986; formal letter dated February 20, 1986. The

letter requested a listing of all pipe support evaluations

where FSAR commitments were exceeded but interim requirements

were met.

• Bechtel letter to TED, BT-16335, dated February 14, 1986

provided the list requested by TED.

• Bechtel Interoffice Memorandum from Project Engineer to the

Group Supervisors, "NRC Violation - NCRs Interim /Short-Term

Allowables," dated March 15, 1986 provided additional

instructions and training for the performing of evaluations,

c. Item 1.a(3) (Closed)

See RIII Inspection Report No. 50-346/85035, Paragraph 4.b.

.

7

, . _ _ _ _ _ . _ . _ _ _ . - . . _ - _ . _ -._ _ ._ ._

. .

.

.

d. Item 1.a(4) (0 pen)

The FCRs that could impact safety-related piping system operability

are listed in RIII Inspection Report No. 50-346/85035, Paragraph 4.b.

An update of this list (as of January 15, 1986) is as follows:

(1) FCR No.77-213, 77-398,80-221, and 80-276: No Maintenance

Work Orders (MW0s) were issued for these FCRs. Since related

modification work will not affect system operability, the MW0s

will be developed after restart.

4

(2) FCR 78-360: This FCR was voided.

(3) FCRs78-126, 79-308,83-151, 85-086,85-010, 85-126,85-163,

85-176, and 85-224 were closed.

(4) Status of the remaining FCRs:

!

General Description

-

FCR No. (No. of Supports Involved) Status79-421 AFW pump turbine modification (15) Work completed,

in process of

being closed

out.

,

83-136~ Replace FW pump governor (2) Work in progress

,.83-138 Change of 14 valves (4) Work in progress- 85-025 Motor driven FW pump (103) Work in progress

-

85-143 Relocate steam admission Work in progress

valves (6)85-160 PORV loop seal drain (8) Work in progress

e. Item 1.a(5) (0 pen)

,

The documentation in RIII Inspection Report No. 50-346/85035,

! regarding Paragraph 4.b remains unchanged.

5. Status of Completion of IE Bulletin 79-14

RIII Inspection Report No. 50-346/85031, Paragraph 6.a, documents that

TED's ineffective utilization of the FCR system resulted in some support

{. component rework not being completed as stated in a TED letter to RIII.

'

In response to the RIII findings, TED reported in a letter to RIII (Serial

No. 1-598, dated December 20, 1985) that several FCRs which were originally

issued as a result of the I&E Bulletin walkdown were identified as still

being open, and for several of the work items which were originally

identified as closed, the work was not yet fully completed.

l

l

.

8

i

_ _ . - - . . . . , - . _ _ . - . , . _ . _ . ~ . , . . . _ _ . . _ _ . _ . . - . . . _ _ _ _ _ - . _ _ _ _.. .._._ _ _ _

. .

.

.

Due to NRC questions concerning whether or not TED had made misleading

statements to the NRC, the NRC inspector discussed the matters with the

TED licensing and QA directors. Through their cooperation, the NRC

inspector obtained the following information to aid in his assessment

of the matter:

a. TED Letters to RIII Documenting the Status of IEB 79-14

Implementation Scope or Status

Serial No. (Date) Connitment Date of Work

1-137 (6/16/80) 3/1/81 Approximate 210 supports for

accessible areas

1-177 (12/20/80) 12/31/81 207 supports for accessible

areas

3/82 outage 15 supports for inaccessible

areas (only 60% analyses

completed)

1-187 (2/13/81) 3/82 outage Modification will be made

for inaccessible areas

1-201 (5/22/81) 3/82 outage 45 supports for ipaccessible

areas

1-223 (11/13/81) 12/82 133 supports for inaccessible

areas

1982 outage Remaining supports inside

containment

1-289 (8/19/82) Next outage 11 support modifications

inside containment

1983 Remaining supports outside

containment

1-429 (5/31/84) 12/31/83 All supports " mechanically

completed"

b. Causes of Support Modifications not Being Completed As Stated in TED

Letter 1-429

Actual Completion

Document No. Support No. Date Causes

FCR 80-87 FSK-M-CCB-8-25-H 12/84 Bechtel failed

to provide TED

with modification

package

9

_--

. .

.

.

Actual Completion

Document No. Support No. Date Causes

FCR 80-87 FSK-M-HCD-15-18-H 9/85 Same as above

FCR 80-91 A 399/A 402 4/10/84 TED Licensing

(NCR393-79) (instead of oversight

Feb.'84)

NCR 396-79 EBD-12-19 4/30/84 TED Licensing

and EBD-12-20 (instead of oversight

Feb.'84)

FCR 80-125 SR 33, 35, and 5/1/84 TED FMD

41 oversight

FCR 80-125 SR37 3/30/84 TED FMD

oversight

FCR 80-125 A171 1/16/84 TED FMD

oversight

NCR 524 A8 11/85 TED FMD and

QC oversight

As a result of the review, the NRC inspector concluded that (1) from 1980

to 1982 TED reported the status of work as requested in IEB 79-14, (2) it

appears that substantial funds were spent to implement the actions set

forth in IEB 79-14 and IEB 79-02, and (3) despite the identified

deficiencies, less than 4% of the components were not modified as reported

in the TED letters to RIII. No further action is planned by Region III

at this time.

6. AFPTSS System Modification

Actions taken by TED to modify the AFPTSS system are discussed in RIII

Inspection Report No. 50-346/85035, Paragraph 6. During the site

inspection of January 14, 1986, the NRC inspector performed a waltdown

of the AFPTSS crossover leg piping connecting steam generator No. 1-2 to

auxiliary feedwater pump turbine No.1-1. This crossover leg, with its

many piping expansion loops, is located in the " Fan Alley" (area having a

large amount of HVAC equipment) on floor elevation 623'-0" inside the

auxiliary building.

a. Inspection of Piping Restraints

An inspection of piping restraints and anchor modifications, to

evaluate their functionability, was conducted by the NRC inspector.

The piping section observed was between valve HV 106 and valve

HV 106A. No deficiencies were identified.

10

. .

.

.

b. Questions Concerning High Thermal Lockup Stress

During the walkdown, the NRC inspector observed excessive restraining

within the region bounded by restraints 3A-EBD-19-H 104 and

3A-EBD-19-H 109. This region involved four directional changes in

less than 22 feet of 6" diameter pipe. The restraints are:

H 104 -

X and Y restraints

H 105 -

Z restraint

H 106 -

X restraint

H 107 -

Y restraint

H 108 -

Z restraint

H 109 -

X and Y restraints

The primary reason for all the thermal expansion loops installed in

the crossover leg is to minimize piping thermal stresses. The

placement of so many rigid restraints within the pipe loop is a

contradiction of this design intent.

In January 1986 the NRC inspector reviewed Bechtel piping stress

analysis Problem 40A, " Main Steam," Revision D2, dated September 25,

1985, and observed: (1) a high ANSI B31.1 secondary stress (RIII

Inspection Report No. 50-346/85035, Paragraph 6.d.(2)(a) reported a

maximum ASME Section III secondary stress only), and (2) a piping

dimensional deviation at a critical area between rigid restraint

H 105 (Data Point 122) and pipe elbow (Data Point 124). The Bechtel

analysis showed 25", the Bechtel Hanger Location Drawing HL-203J

showed 15", and the field remeasurement taken during the inspection

was 14". During the NRC inspection the piping stress analysis was

rerun in a simplified configuration based on corrected dimensions.

The results were:

Data Point Thermal and Existing /New Code allowable

Existing /New SAM Stress (psi)  % Change (psi) reference

124/72 33,942/33,298 -1.9% 37,500

128/80 8,858/12,690 +43.3% 37,500

132/90 15,410/24,043 +56% 37,500

Subsequent to the review, the NRC inspector noted:

~(1) The present design met code requirements but system relief

should be provided.

(2) The present TED Procedure IP-M-002, "The Piping Support Inspec-

tion and Verification Program: Verification of Support / Component

Location and Quantity," Revision 1, dated November 7, 1985 . states

" Denote all dimensional differences on the walkdown HL drawings."

The dimensional differences are then evaluated by Bechtel to

determine if there will be any effect on the pipe stress analyses.

The NRC inspector stated his view that priority should be given

to critical /high stressed areas.

11

. .

.

.

7. Bechtel Design Control for HELB Analysis

A number of unresolved matters were raised during a previous NRC inspection

conducted at Bechtel on December 4-6, 1985 (RIII Inspection Report

No. 50-346/85035, Paragraph 6.e(4)). A followup review conducted at

Bechtel on January 29-30 and April 2, 1986 revealed the following

conditions:

a. The HELB analysis for the AFPTSS modification was based on desk top

design guides. Bechtel Procedure, MGP-05," Procedure for

Interdiscipline Coordination for High Energy Line Break Evaluation

at Davis-Besse Nuclear Power Station Unit No.1," Revision 1. was

issued for use on March 14, 1986.

b. The desk top design guide, " Pipe Whip Details," was determined to be

unacceptable by Bechtel upon reevaluation. Among the 21 postulated

AFPTSS break locations, five indicated that the plastic hinges will

form at the third elbow instead of the second elbow. However, new

whip restraints were not required due to the location of the impact

areas. The impact areas are as follows:

Break No. Area of Impact

6 ceiling

21 wall penetration

36 wall

38 wall

69 no safety-related equipment

A Bechtel QA Management Audit, No. 12501-05, issued QA Finding No. 1

on February 28, 1986, documenting a similar finding.

c. Design tables utilized in the AFPTSS HELB component design were not

approved for the specific applications. These design tables also had

not been evaluated for applicability. Subsequent Bechtel evaluation

determined that the tables presented in RIII Inspection Report

No. 50-346/85035, Paragraph 6.e(4)(c), were conservative for the

application.

d. The remaining two unresolved matters documented in RIII Inspection

Report No. 50-346/85035, Paragraphs 6.3(4)(d), and 6.e(4)(f) were

adequately addressed in MGP-05.

e. The NRC inspector questioned whether or not all the previous Bechtel

HELB analyses were adequate. Bechtel's response was as follows:

• The original design criteria were very conservative. The FSAR,

Revision 13, dated June 1975, states in Paragraph 3.6.2.2.2,

" Pipe Restraint Design Criteria to Prevent Pipe Whip Impact

Outside the Containment - The basic philosophy used to prevent

high energy pipes from whipping is to provide restraints of

12

.

.

.

sufficient capacity and with such spacing that pipe whipping

cannot develop. These restraints are independent of operating

and seismic supports. Consequently, they are designated for

pipe rupture loads only. Pipe rupture of either guillotine

or side-split type are postulated to occur in accordance

with Table 3-4b for the 36 inch main steam and 18 inch main

feedwater lines. Allowable pipe spans are calculated, assuming

that the force developed during the accident experience is

transferred to the pipe restraint. The pipe restraints are

then designed to withstand this force. The restraints are

further designed to prevent the pipe from shearing off and

generating missiles. Jet effects from ruptured pipe are

, considered in designing pipes, walls, and shield. The

concurrent effects of jets and pressure differentials are

also considered in designing walls and shields."

• Bechtel reevaluated the Steam Generator Blowdown System modifica-

tion conducted in 1982. Of the approximate 20 postulated pipe

break locations, one was found having a plastic hinge location

at the third elbow instead of the originally determined second

elbow. Field inspection observed no safety-related equipment

in the force impact areas.

The NRC inspector considers the Bechtel response acceptable.

8. Followup on Licensee Event Report (LER)

The inspector reviewed the TED LER 85019, "PORV Discharge Line

Overstressed Due to Inadequate Heat Trace," dated November 6, 1985, and

questioned the TED actions taken to identify and resolve the issue.

a. TED Inspection Findings

During recent hanger reinspections the following damage and

deficiencies were identified on the PORV discharge piping:

Hanger No. NCR No. (date) Damages / Deficiencies

30 GCC-9-H10 0920 (10/3/85) Concrete cracked; plate

separated from wall.

30 GCC-8-H6 0921 (10/3/85) Concrete cracked and

fell; plate separated

from wall.

30 GCC-8-H5 0914 (10/4/85) Concrete spalling.

30 GCC-8-H7 0822 (10/1/85) Pipe clamp slipped off

and binded.

13

. .

.

.

Upon removal of the H10 and H6 baseplates for repair, the following

abandoned drilled anchor bolt holes were discovered.

Hanger No. Abandoned Holes

H10 Three holes were outside the three

bolt diameter limit (will not

affect strength of installed bolts)

One hole was at the 2.25 diameter

limit. Evaluation determined

the spacing to be acceptable.

H6 One hole was outside three bolt

diameter limit.

Chip void in concrete. Evaluation

determined this void to be

acceptable.

Since two out of two baseplates that were removed for repair revealed

abandoned anchor bolt holes, the NRC inspector indicated to the

licensee that additional inspections of the areas behind other

baseplates in the vicinity of H10 and H6, including Hanger

6C-EBB-4-H12 (see Paragraph 9 for justification), should be

considered. Depending on the hole / void configuration and relative

distance to the affected anchor bolts, these holes / voids could mean

reduction in support strength or fracturing of concrete at a loading

much lower than full load capacity. The TED engineering department

did not share the NRC inspector's view. TED's justification for not

inspecting for possible abandoned drilled holes existing in other

baseplates will be discussed further during a future inspection.

This is ar. unresolved item (346/86004-02).

b. TED Correr_t13 e Actions

As a resul' ;f the TMI event, the NRC issued NUREG-0737 in 1980

requesting vtilities to reevaluate the pressurizer safety and relief

valve operations. The Teledyne Engineering Services (TES) issued

Technical Report (TR), TR-5639-2, " Davis-Besse Analysis and

Evaluation of the Safety / Relief Valve Discharge System per NRC

NUREG-0737," Revision 0, dated January 1983, to address this issue.

This report indicated that the PORV loop seal temperature should be

maintained at 500 F. As a result of the damaged supports observed

to date, the NRC inspector reviewed the operation records and

procedures and concluded that the TED action to ensure design

implementation was ineffective. The bases for the determination were:

(1) From 1976 to 1979 there was a total of 82 PORV lifts where

the loop seal temperature was less than 500 F. Of these, 46 were

assumed by TED to be 130 F. The 46 lifts with the temperature

less than 400 F is in violation of Davis-Besse Periodic Test

14

. .

'

.

1

Procedure, PT 5164.03, " Pressurizer Relief Valve Heat Trace

Test," Revision 5, dated August 20, 1982, which indicates that

the test acceptance criteria for heat trace (T772) must be

above 400 F, to allow PORV (RC 2A) to be lifted 650 times and

25 times for a heat trace below 400 F.

This is a violation of 10 CFR 50, Appendix B, Criterion V

(346/86004-01B).

(2) Contrary to the conclusion stated in TES TR-5639-2, PT 5164.03

was not revised to reflect the latest loop seal temperature of

500 F and no attempt was made to inspect and evaluate the

condition of the piping and supports.

This is a violation of 10 CFR 50, Appendix B, Criterion XVI

(346/86004-03).

c. TES Design Control

The NRC inspector reviewed the following TES reports docun.enting

their evaluations of effects on PORV inlet and discharge piping and

supports:

TR-5639-2, " Analysis and Evaluation of the Safety / Relief Valve

Discharge System per NUREG-0737," dated January 1983.

TR-6388-1, " Analysis of Davis-Besse, Unit 1 Pressurizer Relief

Line 400 F Loop Seal Blowdown," dated September 26, 1985.

TR-6388-2, " Davis-Besse Nuclear Power Station Reconciliation of

ASME Section III Evaluation of Class 1 Pressurizer Relief

Piping," dated November 15, 1985.

TR-6388-3, " Davis-Besse Nuclear Power Station Evaluation of

Class 3 Pressurizer Relief Piping," dated January 21, 1986.

Subsequent to the review, the NRC inspector had the following

concents:

(1) The maximum transient dynamic loading locations where piping

restraint danage was observed, differed from the TES analytical

prediction.

(2) Review of the reference lists revealed what appeared to be an

analytical bases that were either preliminary or interim in

nature.

(3) Several support design loads, based on the present dynamic

transient without loop seal (at 400 F subcooled water condition),

were many magnitudes higher than the original design with loop

seal. The present support loads should have been bounded by

the original design.

15

_ _ _ _ _ __ __ _ _- _ _ _ , _

_ _ _ _ . . _ _ _ _ , _ _

. .

.

.

(4) The support design load combination for rigid restraints should

be the larger of (Thermal + Weight + SSE) or (Thermal + Weight

+ Blowdown). The snubber design load should not consider the

thermal and weight loads. Some of the TES support design loads

could not be verified using the above criteria.

(5) The PORV discharge modes of operation include:

(a) PORV opens on 2450 psig saturated steam.

(b) PORV opens on 2450 psig saturated steam followed

by a transition to subcooled water.

(c) PORV opens on 2450 psig 640 F subcooled water.

(d) PORV opens on 2450 psig 400 F subcooled water.

The present design is based on mode (d). Based on the comment

stated in (3) above, it is not clear that the mode (d) support

loading will bound all modes of operation.

Further review of the subject matter is planned. This is an

unresolved item (346/86004-04).

d. PORV Operability Without Loop Seal

The TED decision to remove the PORV loop seal could affect PORV long

term operability due to a continuous steam leak (small amount) and

hydrogen attack of valve disc and seats. TED letter (A85-30681) dated

August 30, 1985, to Crosby Valve and Gage Company, the PORV designer

and manufacturer, requested evaluation of this condition. The Crosby

response was documented in a letter to TED, dated February 28,

1986. The NRC inspector reviewed the Crosby letter and observed the

valve leak detection devices installed on the piping system. The

measures taken to ensure system safe operation due to the design

modification were determined to be acceptable.

9. Auxiliary Feedwater (AFW) Line Transient

During the inspection conducted inside the containment on February 6, 1986,

the NRC inspector observed concrete spalling and grout cracking at load

intensified locations on the baseplate for AFW Train 1-2 hanger

6C-EBB-4-H12. These conditions are indications of excessive transient

loads that may not been accounted for in the Bechtel design and analysis.

The NRC inspector's review of NCR 85-0198, issued on September 3, 1985, and

evaluated by TED engineering on October 5, 1985, identified no indication

that the causes and measures taken to correct and prevent recurrence had

been evaluated.

During the inspection conducted on February 19-20, 1986, the NRC inspector

further discussed his observations. The TED engineer stated that the

concrete damage most likely occurred during construction. The NRC

inspector re-entered the containment on February 20, 1986 for a closer

examination of hanger H12, and observed the following additional

adverse conditions:

16

. .

.

• One of the anchor bolts appeared to be bent.

• One of the baseplates was partially separated from the wall.

• Markings that could have resulted from thermal displacement and

possible loads in the direction the bolt appeared to be bent, were

observed where the restraint contacted the pipe.

During the NRC inspection conducted at Bechtel on April 1, 1986, TED

presented the following documents to address the NRC inspector's

concern:

• Bechtel letter to TED, BT-16470, " Evaluation of Support 6C-EBB-4-H12

Auxiliary Feedwater System," dated March 17, 1986. This letter

documented the Bechtel inspection and evaluation. Bechtel concluded

that there was no evidence of transient loads affecting the support.

• Bechtel letter to TED, BT-16552, "NRC Inspection AFW System Dynamic

Transients," dated April 2, 1986. This letter compared the steam

condensation with the check valve leak rate and concluded that a

severe dynamic transient will not occur.

The TED representative further committed to reinspect the piping system

including Support SC-EBB-4-H12 during the next refueling outage after

restart. The NRC inspector reviewed the above documents and considered

the TED actions to be acceptable.

10. TED Review of Grinnell Calculations

Due to the lack of a formal design interface control etween TED, Bechtel,

and Grinnell (now a Tyco company) and the fact that G 'nnell did not have

final design responsibility for the adequacy of the or., nal 4,000

(estimate) safety-related support calculations and subsequeat IEB 79-02

and IEB 79-14 evaluations, the calculations performed by Grinnell were

in question. Since completion of the IEB 79-02 and IEB 79-14 work, some

of the Grinnell hanger calculations were replaced through the disposition

of Bechtel NCRs and FCRs. The licensee is in the process of obtaining

all hard copies of the Grinnell original calculations and IEB 79-02 and

IEB 79-14 calculations so they can complete their review of this work.

The NRC inspector discussed the matter with the licensee and indicated

that the following areas warrant additional review:

a. In January 1986 TED requested Grinnell transmit all Davis Besse 1

original hanger calculations and subsequent calculations for

IEB 79-02 and IEB 79-14 to TED. No documents have been received to

date. Further efforts are required to obtain these calculations if

they are available,

b. Upon receipt of Grinnell calculations, TED will develop a program

to review these calculations to assure compliance with design

procedures. The NRC inspector concurred with TEDS recommendation

that this work be completed prior to the next refueling outage.

17

_

. .

d

.

c. The TED QA Department recently conducted an audit in the area of

design interface of vendors who provide engineering service to TED.

The NRC inspector plans to revier the audit report during a

subsequent inspection.

This is an unresolved item (346/86004-05).

11. Unresolved Items

An unresolved item is a matter about which more information is required

in order to ascertain whether it is an acceptable item, an open item, a

deviation, or a violation. Three unresolved items disclosed during this

inspection are discussed in Paragraphs 8.a. 8.c, and 10.

12. Exit Interview

The NRC inspector met with licensee representative (denoted in Paragraph 1)

at the conclusion of the inspection. The inspector summarized the scope

and findings of the inspection. The inspector also discussed the likely

informational content of the inspection report with regard to documents

reviewed by the inspector during the inspection. The licensee

representatives did not identify any such documents as proprietary.

18