IR 05000445/1988068

Insp Repts 50-445/88-68 & 50-446/88-64 on 881005-1101.No Violations Noted.Major Areas Inspected:Action on 10CFR50.55(e) Deficiencies,Followup on NRC Bulletins,Plant Tours & Comparison of as-built Plant to FSAR Description

ML20206H523
Person / Time
Site:	Comanche Peak
Issue date:	11/17/1988
From:	Bitter S, Burris S, Joel Wiebe NRC OFFICE OF SPECIAL PROJECTS
To:
Shared Package
ML20206H515	List: IR 05000445/1988068 ML20206H510
References
50-445-88-68, 50-446-88-64, IEB-88-002, IEB-88-2, NUDOCS 8811230219
Download: ML20206H523 (8)

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S. NUCLEAR REGULATORY COMMISSION OFFICE OF SPECIAL PROJECTS NRC Inspection Report:

50-445/88-68 Permits: CPPR-126

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50-446/88-64 CPPR-127

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Dockets: 50-445 Category: A2 50-446 Construction Permit Expiration Date:

Unit 1: Extension request submitted.

Unit 2: Extension request submitted.

Applicant:

TU Electric j

Skyway Tower 400 North Olive Strcot Lock Box 81

Dallas, Texas 75201 Facility Names Comanche Peak Steam Electric Station (CPSES),

Units 1 and 2 Inspection At:

Comancho Peak Site, Glen Rose, Texas Inspection Conducted:

Octobor 5 through November 1, 1988 Inspectors d _M Il

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S. D(,Dittor,Rog0d'c~nt Inspector, bate Operations

!If n {?B Inspector:

S.

P. Burris, Senior Resident Inspector, dato Operations l

Reviewed by:

ItMV3 YO IIf11[Et J.

3. Wicbo, Lodd Project Inspector Date Ckh$$dd[43 PDC

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Inspection Summary Inspection Conducted:

October 5 through November 1, 1988 (Repcrt 50-445/88-68: 50-446/88-64)

Areas Inspected:

Unannounced resident safety inspection of applicant action on 50.55(c) deficiencies, followup on NRC Bulletins, plant tours, and comparison of as-built plant to F7A'4 description.

Results:

Within the areas inspected, no violations or de"lations were identified.

No significant strengths or weaknesses were noted.

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DETAILS 1.

Persons Contacted

• R.

W. Ackley, Jr., Director, CECO'

• R.

P.

Baker, Licensing Compliance Manager, TU Electric

• J.

L. Barker, Marigor, Engineering Assurance, TU Electric

• D.

P. Barry, Man

'cr, ESG, Stone and Webster Engincoring Corporati).

(SWEC)

osident, Nucicar Engincoring, TU Electric-J.

W. Beck, Vice i

• M.

R. Biovins, Ma...ger, Technical Support, TU Electric

• H.

D. Bruner, Soniar Vice President, TU Electric

• W.

J. Cahill, Consultant, TU Electric

• J.

T. Conly, APE-Licensing, SWEC

• G.

G. Davis, Nuclear Operations Inspection Poport Item Coordinator, TU Electric

• R.

D. Delano, Licensir.g Engineer, TU Electric

• D.

E. Devincy, Deputy Director, Quality Assurance (QA),

TU Electric

• G.

L.

Edgar, Attorney, Newman

.1d Holtzinger

• G.

E. Grabruck, QA, Impell

• W.

G.

Guldemond, Executive Assistant, TU Electric

• P.

E. IIalstead, Manager, Quality Control (QC), TU Electric

• T.

Heatherly, Licensing Compliance Engineer,

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TU Electric

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• C.

B. Hogg, Engincoring Manager, Bochtel

• R.

T. Jenkins, Manager, Mechanical Enoineering, TU Electric

• J.

J. Kelley, Manager, Plar.t Operations, TU Electric

• O.

W.

Lowe, Director of Engineering, TU Electric

• F.

W. Maddon, Mcchanical Engineering Manager, TU Electric

• G.

M. McGrath, TS/SP Manager, Startup, TU Electric

• J.

C. Miller, Site Manager, TENERA

• J.

W. Muffett, Manager of Civil Engineering, TU Electric

• L.

D. Naco, Vice President, Engineering & Construction, TU Electric

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• E.

F. Ottney, Representative, CASE

• S.

S.

Palmor, Project Manager, TU Electric

• J.

D. Redding, Executivo Assistant, TU Electric

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• D.

M. Reynorson, Director of Construction, TU Electric

• M.

J. Riggs, Plant Evaluation Manager, Operations, TU Electric

• J.

C.

Smith, Plant Operations Staff, TU Electric

• P.

D.

Stevens, Manager, Electrical Engincoring, TU Electric

• J.

F. Streeter, Director, QA, TU Electric

• C.

L. Terry, Unit 1 Project Manager, TU Electric

• T.

G.

Tyler, Director of Projects, TU Electric

• J.

R. Waters, Licensing Complianco Engineer, TU Electric The NRC inspector also interviewed other applicant employces during this inspection period.

• Deno; s personnel present at the November 1, 1988, exit intereicw.

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2.

Action on 10 CFR Part 50.55(c) Deficiencies Identified by the Applicant (92700)

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(a)

(Closed) SDAR CP-86-02, "Turbino Driven Auxilia ?

Foodwater Pump Performance":

This issue arose when the initial preoperational test results for the pump indicated that it could not deliver the required flow rate at the requirad head.

This deficiency stemme( from the implementation of a system design change that altered the capabilitics as well as the requirements for the

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pump.

Reviews conducted by the Westinghouse and applicant engineering staffs determined that when the system design change is taken into account, the head / flow requirements for the pump becomo less stringent.

In order to reflect these new requirements, revisions to the FSAR and draft Technical Specifications have been approved and issued.

Thorofore, based on ovaluation of the initial prooperational test results, the current system design has boon found to satisfy the FSAR requirements and the pump is capable of performing its intended safety function.

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The applicant determined that this apparent deficiency

would not have adversely affected safety if it had remained uncorrected, and thorofore, is not reportable under the provisions of 10 CFR 50.55(c).

Based on this review, the inspcctors considor this itom closed.

(b)

(Closed) SDAR CP-86-28, "Service Water Discharge Piping":

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This issue stemmed from the reclassification (from safety Class 3 to a nonsafety class) of a particular portion of

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i the service water piping located outside buildings.

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i piping involved was originally designed, purchased, and installed as safety Class 3 and was buried in Class 1

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backfill up to the discharge structare.

Subsequently,

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the portion of the piping from the discharge structure to

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the first wcld upstream (40 foot upstream) of the

discharge structure was reclassified as nonsafety.

l In accordance with 10 Crn 21, Gibbs and Hill expre.. sed a t

concern to thc Une that a break in the newly rec 14 :sified portion of the service water system could compromise the T

service water system's ability to act as an ultimate heat sink when required.

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The applicant addressed this issue by evaluatins as both 10 CFR 50.55fo) and 10 CFR 21 concerns.

evaluation indicated that the change in classitt>. tion would not compromise the service water system's ability

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to provide safety-related cooling; even if a moderate energy leakage crack were to occur in the nonsafety-related portion of the piping, the ensuing leakage would flow downhill to the Safe-Shutdown impoundment.

Therefore, the applicant han concluded that this is neither a 10 CFR 21 nor a 10 CFR 50.55(c) issue.

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The NRC inspectors consider this item closed.

(c)

(Open) SDAR CP-86-34, "Defective Fire Stop Installations":

This issue originated from a 10 CFR Part 21 notica concerning penetration fire seals supplied by B&B Promates.

The applicant's review determined that the notice is not applicable to the fire seal

configuration used at CPSES.

However, during this

review, the applicant did identify a fire test in which the unexposed surface of the fire seal reached

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369 degrees F.

This exceeds the 325 degree F limit specified in NUREG-0800.

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Tbc applicant resolved this 'ssue by determining that tney are not committed to NLAEG-0800; that ASTM E-119 specifies a limit of 325 degrees F "rise" for a single point; and that, assuming a 75 degree F initial temperature plus a 325 degree F temperature rise, the actual absoluto limit is 400 degrees F.

Using these assumptions, the applicant determined that the test temperature of 369 degrees F is acceptable.

The inspector disagrees w'th the applicant's resolution.

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Although the applicant is not committed to NUREG-0800 and

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the applicant is committed to Appendix A to Branch

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Technical Position (BTP) APCSB 9.5-1; Generic Letter 86-10, "Implementation of Fire Protection Requirements" dated April 24, 1986, indicates that the guidance in Generic Letter 86-10 takes precedence over prior

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guidance.

In addition, Generic Letter 86-10 states that

the guidance in Generic Letter 86-10 represents recent NRC assessment of questions and provides acceptable

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l methods of satisfying regulatory requirements.

Other proposed methods for complying with Commission regulations will be considered on their own merits, i

Generic Letter 86-10, Enclosure 2, Section 3.2.1, provides guidance for fire barrier acceptance criteria.

The guidance states that the NRC staff position is that an analysis is required if any temperature on the unexposed side of the barrier exceeds 325 degrees F.

This issue temain s open pending further NRC review to determine if the applicant is deviating from his commitments or from NRC guidance concerning this issu __ _ - _ -

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(d)

(Closed) SDAR CP-88-04, "Overpressurization of the Auxiliary Foodwater Suction Line":

The applicant

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identified this item during a design validation of the auxiliary feedwater (AFW) system.

Specifically, the concern was that backloakage through the reverse flow check valves could allow high pressure water from the operating steam generators to pressurize the suction side of the idle AFW pumps' suctions.

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The applicant reviewed the system design and has determined that, even with a failure of the reverse flow check valves, the AFW suction lines would not be overpressurized bec.nuso any potential overpressure would

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be relieved to the condensato storage tank via the recirculation lines.

I These recirculation lines, originating at the discharge of each AFW pump and cischarging to the condensate storLge tank are maintained unisolated during normal system operation with locked-open manual valves and fail-open pneumatic valves.

orificos in each line prevent condensate storage tank overpressurization by reducing pressure and thus proventing overpressure, in the downstream piping.

Furthermoro, the applicant is implementing design changes that call for the installation of relief valves in the recirculation lines.

In summary, the applicant has determined that this concern is not reportable under the provisions of 10 CFR 50.55(o).

The NRC inspectors consider this item closed.

3.

Inspectiori of NRC Bulletin 88-02 (92700)

This bulletin, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubo," requested that license and permit holders of

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Westinghouse designed nucicar power reactors with Westinghouse steam generator models No. 13, 27, 44, 51, D1, D2, D3, and D4 perform inspections of their specific steam generators to provent a similar occurrence such as the one that occurred at North Anna Unit 1 on July 15, 1987.

The applicant responded to the bulletin requirements in the following manner:

a.

This bulletin is applicable to Unit 1 steam generators only.

b.

Based on a 1982 baseline eddy current test and a 1986 inspection of rows 1 and 2 of each steam generator, no tubo denting was identified.

In addition, during the first refueling outage, further eddy current testing will

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be performed in accordance with CPSES Technical Specifications.

c.

An enhanced primary-to-secondary leak rate monitoring program is not applicable to CPSES because CPSES Unit 1 is not yet operational.

Any denting due to corrosion, should it occur after becoming operational but prior to the first refueling cutage, is not expected to degrade the tubes to the extent that an enhanced monitoring program would be required.

d.

Future Unit 1 steam generator eddy current tost results will be examined for any signs of tubo denting.

If any denting is found, then future corrective actions will be implemented at that time.

The NRC inspectors have reviewed the applicant's response and discussed the bulletin with responsible individuals.

The NRC inspectors have determined that the applicant has adequately impicmented the requir:monts of this bulletin.

Thorofore, this item is considered closed.

4.

Plant Tours (71302)

The NRC inspectors conducted numerous plant tours during this inspection period.

These tours provided coverage during normal, off-normal, and backshift working hours.

NRC inspection activities included reviewing work documentation, witnessing ongoing work activitics, observation and interviews of shift operations personnel, reviewing the status of control room construction work, reviewing the status of system and component completion, determining the status of Units 1 and 2 equipment lay up, observit.g housc. keeping activities, and inspecting for general safety compliance.

To support these activitics, NRC inspectors attended plan-of-the-day meetings and discussed plant status with operations personnel.

During the course of the tours and inspections, no discrorancies were noted.

5.

Comparison of As-Built plant to FSAR Description (373011 During this inspection period, the NRC inspectors continued the walkdown of the accessible portions of the auxiliary foodwater system (AFW).

To assist the walkdown effort, an NRC inspector discussed the AFU system status with the system engineer.

As a result of this discussion and the continuing walkdown, the inspectors have verified that:

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All system valves inspected woro installed c crectly and exhibit no visible damage.

b.

All major system components inspected woro labeled correctly.

c.

Those portions of the system inspected corresponds with the latest drawing revision (when outstanding DCAs are considered).

No items of concern have been identified during this reporting period.

Completion of the AFW system walkdown is anticipated during the next inspection period.

6.

Exit Meeting (30703)

An exit meeting was conducted on November 1, 1988, with the applicant's representativos identified in paragraph 1 of this report.

No written material was provided to the applicant by the inspectors during this reporting period.

The applicant did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.

During this mcoting, the NRC inspectors summarized the scopo and findings of the inspection.