ML20132E116

Forwards Insp Rept 50-445/84-45 Containing Corrected Pages 14-17 Due to Inadvertent Omission of Items 6.b.(3) & 6.c. Related Correspondence

ML20132E116
Person / Time
Site:	Comanche Peak
Issue date:	09/24/1985
From:	Mizuno G NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Bloch P, Grossman H, Jordan W Atomic Safety and Licensing Board Panel
References
CON-#385-622 OL, NUDOCS 8509300395
Download: ML20132E116 (28)

See also: IR 05000445/1984045

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20555

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September 24, 1985

'85 SEP 26 R2:39

Peter B. Bloch, Esq., Chairman

Dr. Kenneth A. McCollom

$CNE AN b

Administrative Judge

Administrative Judge

ERMCH

Atomic Safety and Licensing Board

Dean, Division of Engineering,

U.S. Nuclear Regulatory Commission

Architecture and Technology

Washington, DC 20555

Oklahoma State University

Stillwater, OK 74078

Herbert Grossman, Alternate Chairman

Elizabeth B. Johnson

Administrative Judge

Administrative Judge

Atomic Safety and Licensing Board

Oak Ridge National Laboratory

U.S. Nuclear Regulatory Commission

P.O. Box X, Building 3500

Washington, DC 20555

Oak Ridge, TN 37830

Dr. Walter H. Jordan

Administrative Judge

881 W. Outer Drive

Oak Ridge, TN 37830

In the Matter of

Texas Utilities Generating Electric, et al.

(Comanche Peak Steam Electric Station, Units T and 2)

Docket Nos. 50-445 and 50-4460t-

Dear Administrative Judges:

By letter of March 6,1985, NRC Staff counsel transmitted copies of

NRC Inspection Report 84-45 (March 5, 1985) to the Board. Subsequently,

the Staff determined that the discussions for items 6.b.(3), 6.c. and

d. were incorrect or omitted in pages 14-16 of the inspection report.

The most significant item was the inadvertent omission of item 6.c.,

which is a discussion on Deviation 445/8415-01. Accordingly, the Staff

has issued new pages 14-17 to replace original pages 14-16. See Enclo-

sure 1.

Copies of Inspection Report 84-45 which include new pages 14-17

are enclosed for the information of the Board as Enclosure 2.

Hence,

the original version of Inspection Report 84-45 should be discarded.

Sincerely,

8509300395 850924

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ADOCK050004j5

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PDR

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Gea y S

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Counsel for NRC Staff

Enclosures:

As stated

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Service List

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ENCLOSURE 1

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_ ENCLOSURE 1

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In Reply Refer To:

SEP 18 25

Dockets:

50-445/84-45

Texas Utilities Generating Company

ATTN: Mr. W. G. Counsil

Executive Vice President

400 North Olive, L.B. 81

Dallas, Texas 75201

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Gentlemen:

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This refers to the NRC Inspection Report 50-445/84-45.

Enclosed are new

pages 14-17 to provide corrections to the report details, Section 6, Applicant

Action on Previous Inspection Findings (items 6.a.(3), b.c., and 6.d).

Please

replace the original pages 14-16 with new pages 14-17.

Should you have any questions, please contact us.

Sincerely,

,

Oricinal signed Byi

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incne.rd P. Denise

R. P. Denise, Director

Division of Reactor Safety

and Projects

.

Enclosure:

As stated

cc:

Texas Utilities Electric Company

ATTN:

J. W. Beck, Manager,

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Licensing

Skyway Tower

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400 North Olive Street

Lock Box 81

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Dallas, Texas

75201

Texas Radiation Control Program Director

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b.

(Closed)OpenItem 445/8415-02:

Minor discrepancies found during NRC

inspection of station administrative procedures.

During a previous

inspection (445/84-15) of station administrative procedures, the RR1

found several minor discrepancies and made some suggestions to

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preclude future problems.

The applicant took action on those items

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that the applicant considered justification for implementing a

procedure change.

F,or example:

(1) STA-401, " Station Operation Review Committee," Revision 5,

Section 4.4 did.not fully implement the responsibilities of the

committee as stated in the CPSES Unit 1 Technical Specifications

(final draft).

This was corrected in Revision 8 of STA-401.

.(2) STA-203, " Control of Station Manuals," Revision 7. Section 4.3.3

required a notification memo to be sent to each onsite holder of

controlled station manuals to alert recipients of a revision or

new procedure.

This was not being done for holders of the

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manual who incorporate their own changes because they sign a

receipt for the changes or new procedures anyway.

Revision-9

clarified this such that the applicant is in compliance with the

procedure.

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(3) STA-307, " Forms Control," Revision 2, allowed minor changes to

forms without revising the parent procedure containing a sample

of the form as an attachment.

However, instead of changing the

revision number of the form itself, the office services staff

misinterpreted Section 4.2.6 of STA-307 and changed the revision

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of the parent procedure attachment page, which caused a conflict

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with the rest of the parent procedure pages.

This was corrected

by the applicant and STA-307 was revised to preclude

misinterpretation.

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This item is closed.

c.

(Clos.ed) Deviation 445/8415-01:

Failure of the applicant to use 50RC

approved instructions to perform work on the emergency diesel

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generators.

The CPSES FSAR commits to Regulatory Guide (RG) 1.22,

Revision 2, February 1978 with no exceptions. RG 1.33 and ANSI

N18.7-1976Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1976" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. to which it refers, requires maintenance to be performed

using procedures / instructions receiving the same review and approval

as operating instructions, i.e., review and approval by the SORC.

During two previous inspections (50-445/84-07 and 50-445/84-15), the

NRC inspectors noted that the applicant had defined " instructions" as

procedures wnich do not require SORC approval, and had issued

" instructions" to perform work on safety-related equipment such as

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the emergency diesel generator (EDG). The apparent basis was that

EDG work performed by the maintenance department had no significant

impact on other departments, and/or was work unique to the

maintenance department.

Since the above NRC inspections, the issue

has been resolved as evidenced in Station Administrative Procedure

STA-707, " Safety Evaluations," (Revision 2) STA-202, " Preparation,

Review, Approval, and Revision of Station Procedures" (Revision 10),

and the final draft'of the CPSES Unit 1 Technical Specifications

(TS).

In. essence., all. safety-related procedures and instructions

will receive.a SORC review by virtue of the requirement that the SORC

review the related safety evaluations, as stated in the TS and

STA-401, which both list the responsibilities of the 50RC.

This deviation is closed.

d.

(Closed) Violation 445/8421-02:

Failure of preoperational test

procedures to provide adequate prerequisites.

During a previous

inspection (445/84-21), the RRI noted that during conduct of

preoperational test ICP-PT-29-02,RT1, " Diesel Generator (DG) Control

Circuit Functional and Start Test" the DG barring device was

connected to a portable air comprassor instead of the Service Air

System. There was no' prerequisite step in the test procedure to

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provide either temporary or permanent air for the barring device, yet

it needed air to be tested.

Also, during testing of the Service

Water System in accordance with ICP-PT-04-01, RT 1, " Station Service

Water (SSW)," a Barton D/P gage did not function due to air binding.

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There was no prerequisite in the test procedure to ensure the gage

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was recently filled and vented to assure accurate test data, nor did

the Startup Administrative Procedures for writing the test require

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it.

This wa:: e notentially generic problem. The applicant has since

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revised CP-SAP-7, " Format and Content of Test Instruction / Procedures"

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to require the appropriate prerequisites.

Each organization

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responsible for review of preoperational test procedures has been

instructed to ensure that test prerequisites receive a comprehensive

review to ensure system readiness and correct component configuration

to assure validity of the test results.

This item is closed.

7.

Plant Tours

During this reporting period, the SRRI and RRI conducted several

inspection tours of Unit 1.

In addition to the general housekeeping

activities and general cleanliness of the facility, specific attention was

given to areas where safety-related equipment was installed and where

activities were in progress involving safety-related equipment.

These

areas were inspected to ensure that:

Work in progress was being accomplished using approved procedures.

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Special precautions for protection of equipment were implemented, and

additional cleanliness requirements were being adhered to for

maintenance, flushing, and welding activities.

Installed safety-related equipment and components were being

protected and maintained to prevent damage and deterioration.

Also during these tours,'the SRRI and RRI reviewed the control room and

shift supervisors' log books.

Key items in the log review were:

plant status

changes in plant status

tests in progress

documentation of problems which arise during operating shifts

No deviations or violations were found.

S.

plant Status as of December 31, 1984

a.

The applicant was at the end of the Thermal Expansion Test sequence

and making preparations to roll the main turbine-generator.

Details

of the testing sequence and problems encountered are discussed in

paragraph 2 of this report.

b.

Unit No. 1 is 99% complete with 403 of 422 areas and 323 and 332

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subsystems turned over to operations custody.

" Custody" means having

immediate authority and responsibility for operational control of

system or equipment.

The applicant has accepted 260 of 332 subsystems for final

acceptance.

c.

Of the 199 preoperational tests, one is not yet completed on field

testing, and 21 are pending review and approval of completed data.

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Eighteen are pending NRC completed data inspections,

d.

The following items related to NRC resident operations office

findings are open pending applicant action and NRC followup

inspection to confirm completion of closure:

Violations

10

Deviations

0

Open items

100

Unresolved

7

Total

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Action is underway to complete these items.

Closure will be

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documented in future inspection reports.

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e.

Unit No. 2 is 65% complete. The preoperational test program on

systems associated with NRC.. inspections has not yet started.

9.

Exit laterview

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An exit interv.iew was conducted January 4, 1985, with applicant

represeatatives identified in paragraph 1.

During this interview, the RRI

and Mr. D. M. Hunnicutt of the Region IV NRC office reviewed the scope and

discussed the inspection findings. The applicant acknowledged the

findings.

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ENCLOSURE 2

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ENCLOSURE 2

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Texas Utilities Electric Company

kR 2 8 %

In Reply Refer To:

Docket:

50-445/84-45

Texas Utilities Electric Company

ATTN:

M. D. Spence, President ,TUGC0

Skyway Tower

400 North Olive Street

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Lock Box 81

Dallas, Texas

75201

Gentlemen:

This refers to the inspection conducted by Messrs. D. L. Kelley and W. F. Smith

of this office during the period November 1 through December 31, 1984, of

activities authorized by NRC Construction Permit CPPR-126 for the Comanche Peak

Facility, Unit 1, and to the discussion of our findings with Messrs. B. R.

Clements and J. C. Kuykendall and other members of your staff at the conclusion

of the inspection.

Areas examined during the inspection included:

(1)witnessingofthethermal

expansion test conducted in November and December 1984, (2) review of initial

startup test procedures (3) verification of completion of human engineering

deficiencies (4) Review of completed preoperational test data (5) applicant

actions on previous inspection findings (6) plant tours, and (7) plant status.

Within these areas, the inspection consisted of selective examination of

procedures and representative records, interviews with personnel, and

observations by the inspectors.

These findings are documented in the enclosed

inspection report.

During this inspection, it was found that certain of your activities were in

violation of NRC requirements.

Consequently, you are required to respond to

these violations, in writing, in accordance with the provisions of

Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of

Federal Regulations.

Your response should be based on the specifics contained

in the Notice cf Violation enclosed with this letter.

This violation maybe related to findings identified by the NRC Technical

Review Team (TRT).

If the issue is considered to be similar, you may respend

to this item separately or as part of the Comanche Peak Response Team Action

Plan.

One open item is identified in paragraph 2 and one in paragraph 5 of the

enclosed inspection report, which will require closure by the NRC inspectors at

a later date once the actions are completed by the applicant and a followup

inspection has been completed.

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Texas Utilities Electric Company

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The response directed by this letter and the accompanying notice is not

subject to the clearance procedures of the Office of Management and Budget as

required by the Paperwork Reduction Act of 1980 PL 96-511.

Should you have any questions concerning this inspection, we will be pleased to

discuss them with you.

Sincerely,

" Original $igned By:

D.R. HUNTER"

Dorwin R. Hunter, Chief

Reactor Project Branch 2

Enclosure:

Appendix A - Notice of Violation

Appencix B - NRC Inspection Report

50-445/84-45

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cc w/ enclosure:

Texas Utilities Electric company

ATTN:

J. W. Beck, Manager

Licensing

Skyway Tower

400 North Olive Street

Lock Box El

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Dallas, Texas

75201

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Texas Utilities Electric Company

ATTN:

B. R. Clements, Vice President, Nuclear

Skyway Tcwer

400 North Olive Street

Lock Box 81

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Dallas, Texas

75201

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APPENDIX A

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NOTICE OF VIOLATION

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Texas Utilities Electric Company

Docket:

50-445/84-45

Comanche Peak Steam Electric

Construction Permit: CPPR-126

Station, Unit 1

Based on the results of an NRC inspection conducted during the period of

November 1, 1984, through December 31, 1984, and in accordance with the NRC

Enforcement Policy (10 CFR Part 2 Appendix C), 49 FR 8583, dated March 8,

1984, the following violation was identified:

Failure to provide adeouate procedures appropriate to

circumstances

-10 CFR 50, Appendix "B", Criterion V requires that, " activities affecting

quality shall be prescribed by documented instructions, procedures, or

drawings, of a type appropriate to the circumstances and shall be

accomplished in accordance with these instructions, procedures, or

drawings."

Contrary to the above, an Instrument and Control (I&C) technician

received a first degree thermal burn on his forearm while attempting to

fill the reference leg on a pressurizer level detector (1-LT-0460)'during

hot plant conditions using a procedure that did not contain sufficient

detail to accomplish the task.

The I&C technician was using Instruction

No. 1C1-2007, " Filling and Venting Level Transmitters and Level Indicating

Switches (Wet Leg)" which is a generic procedure that provides general

guidelines for filling and venting level instruments.

This use of a

generic procec'ure is inappropriate for the circumstances, and appears to

have directly contributed to the technician receiving thermal burns

because he connected the low pressure fill equipment incorrectly and

manipulated the wrong valves.

This action resulted in the low pressure

fill equipment being blown off and releasing hot reactor coolant to the

containment atmosphere.

The I&C technician received thermal burns to his

arm from the hot reactor coolant.

This is a Severity Level IV Violation.

(Supplement II.E) (445/8445-02)

Pursuant to the provisions of 10 CFR 2.201. Texas Utilities Electric Company is

hereby required to submit to this office, within 30 days of the dates of this

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Notice, a written statement or explanation in reply, including: (1) the

corrective steps which have been taken and the results achieved; (2) corrective

steps which will be taken to avoid further violations; and (3) the date when

full compliance will be achieved.

Consideration may be given to extending your

response time for good cause shown.

Dated:

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APPENDIX B

U. S. NUCLEAR RE3ULATORY COMMISSION

REGION IV

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NRC Inspection Report: 50-445/84-45

Construction Permit CPPR-126

Docket: 50-445

Category: A2

Applicant:

Texas Utilities Electric Ccmpany (TVEC)

Skyway Tower

400 North Olive Street

Lock Box 81

Dallas, Texas

75201

Facility Name:

Comanche Peak Steam Electric Station (CPSES)

Unit 1

Inspection At:

Glen Rose, Texas

Inspection Conducted:

November I through December 31, 1984

Inspectors:

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D. L. Kelley, Senior Resident Reactor

'Date

Inspector (SRRI)

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(paragraphs 1, 2, 7, and 8)

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- W. - F. Smith, Resident Reactor Inspector

(Tats

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(RRI)

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(paragraphs 1, 2, 3, 4, 5, 6, 7, 8, and 9)

Approved:

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A/5/55'

D. M. Hunnicutt, Section Chief,

Oafe

Reactor Project Section B

Inscection Summary

Insoection Conducted:

November 1 through December 31, 1984 (Report

50-445/84-45)

Areas Inspected:

Routine, unannounced inspection of (1) the Thermal Expansion

Test conducted during November and December 1984, (2) Initial Startup

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Procedures (3) corrected Human Engineering Deficiencies, (4) completed

preoperational ttst data, (5) applicant actions on previous inspection

findings, (6) plant tours, and (7) plant status.

The inspection involved 109

inspector-hours by two NRC inspectors.

Results:

Within the 7 areas inspected, one violation was identified (failure

to provide adequate procedures, paragraph 2).

In addition, two open items

exist; one in paragraph 2 and one in paragraph 5 pending applicant action.

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DETAILS

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1.

Persons Contacted

Acolicant Personnel

• B. R. Clements, Vice Pres'ident, Nuclear Operations

• J. C. Kuykendall, Manager, Nuclear Operations

• C. H. Welch, Quality Assurance Supervisor

"J. C. Smith, Quality Assurance

• R. B. Seidel, Operations Superintendent

• H. A. Lancaster, Startup Quality Assurance Specialist

• J. M. Ward, Startup Quality Assurance Specialist

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• R. E. Camp, Startup Manager

"S. M. Franks, Special Project and Technical Support Lead

R. R. Wistrand, Administration Superintendent

J. J. Allen, Operations Engineer

• R. A. Jones, Manager, Plant Operations

• J. T.

Merritt, Assistant Project General Manager

• L. G. Barnes, Operations Supervisor

• T. Gosdin, Coordinator, Public Information

D. W. Braswell, Engineering Superintendent

J. C. Zimmerman, ISU Coordinator

D. B. Allen, ISU Test Coordinator

B. J. Browning, Thermal Expansion Test Engineer

M. R. Blevins, Maintenance Superintendent

B. Taylor, I&C Supervisor

M. D. Deen, Shift Supervisor

A. W. Rosette, Operations Engineer

• Denotes those present at exit interview.

The NRC inspectors also interviewed other applicant employees during this

inspection period.

2.

Witnessina of Thermal Expansion Test

During the period of this inspection the applicant conducted a series of

pre-fuel load initial startup tests at reactor system temperatures and

pressures ranging from ambient to hot standby.

The principle test

conducted in the 1984 was Thermal Expansion Testing (and Retesting) that

was not completed during the 1983 hot functional test (HFT).

In addition,

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other retests requiring hot standby conditions were completed thereby

reducing the extent of hot plant testing that would be deferred until

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after initial fueling.

The sequence was of approximately 54 days duration,

as planned by the applicant.

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The objectives of this inspection were to establish through observations,

records rev'fews, and independent checks that the testing was conducted in

accordance with approved procedures, and to evaluate the performance of

the applicant's personnel involved in test performance.

The final NRC

inspection of test results will be conducted during a subsequent period

after the Station Operations Review Committee (SORC) has completed its

review of the data.

The performance of these objectives were accomplished

on a sampling basis.

The SRRI and RRI determined that testing appeared

to be conducted.in a careful and controlled manner, with minimal problems

as noted in paragraph 2.b below.

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a.

The following tests were conducted:

(1)

ISU-300A, " Pre-Fuel Load Initial Startup Test Sequence." The

objectives of this test were to provide an overall sequencing of

all the other tests to be conducted, to establish and maintain

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the plant conditions for testing, and to verify acceptable

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reactor coolant pump seal flow.

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(2)

ISU-008A, " Thermal Expansion".

The purpose of this test was to

verify that the ASME Code Class 1, 2, and 3 systems and other

nonsafety class systems which operate at temperatures greater

than 200*F were not restrained during heatup to normal operating

temperature or during cooldown to ambient conditions. This

procedure included verification that loads and clearance gaps

of selected piping system snubbers, spring hangers and pipe

rupture restraints were properly set for free pipe movement.

Component checks consisted of items requiring retest after the

preoperational test conducted during the 1983 HFT (ICP-PT-55-11,

" Thermal Expansion") and items which were not covered by that

test. Measurements will be taken at an initial ambient temperature,

and plateaus of 250'F, 350 F, 450 F and at normal operating

temperature.

A final set of readings was taken after cooldown

to ambient temperature.

(3)

ISU-206A, " Auxiliary Feedwater Performance" The purpose of this

test was to verify that five consecutive cold quick starts of

Turbine Driven Auxiliary Feedwater (TDAFW) pump could be

performed and that 1-LV-2383 (condensate drain valve) functioned

properly during each of these starts to drain condensate from

the steam supply lines.

This test also verified that the time delay from receiving a

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start signal until the TDAFW pump delivered rated flow at rated

pressure was less than 60 seconds.

The time delay was

determined for both trains and will be a summation of all

system delays including channel sensor and actuation logic.

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(4)

ISU-220A, " Turbine Generator Initial Synchronization and

O'verspeed Test".

The objectives during this testing sequence

were to obtain turbine baseline data and to verify the proper

operation and adjustment of the turbine generator system and its

associated auxiliary and support systems to the extent

practicable during noncritical hot plant conditions.

(5)

ISU-234A, " Main Steam Isolation Valves Operability and Response

Times." The purpose of this test was to verify that the full

stroke closure times of the Main Steam Isolation Valves and Main

Steam Isolation Bypass Valves were within the limits specified

in the Comanche Peak Steam Electric Station (CPSES) Final Safety

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Analysis Report and Technical Specifications for Comanche Peak -

Unit 1.

This test also was to demonstrate the operability of

each Main Steam Isolation Valve and each Main Steam Isolation

Bypass Valve.

(6)

ISU-282A, " Containment & Feed Water (FW) Penetration Room

Temperature Survey" With the RCS at the normal operating

temperature and pressure, the objective of this test was to

demonstrate that the various cooling systems were maintaining

temperatures at or below their design limits in the following

areas:

(a) reactor coolant pipe penetrations; (b) containment

average air temperature; (c) neutron detector wells; (d) each

steam generator compartment; (e) the pressurizer room at the

905 foot elevation; and (f) supply air to each reactor vessel

support.

(7) EGT-712A, " Reactor Coolant System Pressure Isolation Valve

Leakage Testing." This was a retest of repaired or replaced

Safety Injection System check valves which did not meet the

acceptance criteria while being tested during the original HFT

of 1983.

In addition to the specific tests above, the applicant took the

opportunity to exercise several integrated plant operation and

standard operating procedures to confirm or correct their accuracy

and adequacy. Also a few dry runs were conducted on pending initial

Startup test procedures to help minimize procedure problems after the

fuel is loaded.

b.

The applicant conducted weekly status and problem review meetings

between the NRC resident inspectors and key managers including the Manager,

Plant Operation, Engineering Superintendent, Maintenance Superintendent,

Operations Superintendent, and Operations Supervisor.

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This meeting provided an opportunity for the RRI and SRRI to assess

applic' ant management involvement in the test sequence, and to address

NRC inspection concerns and actions taken by the applicant, and to

keep abreast of management decisions which affect testing plans for

the week that followed.

The following problems were encountered:

(1) Failed Reactor Coolant Pump Motor:

During the early phases of

reactor coolant' system (RCS) fill in preparation for this

testing sequence, No. 4 reactor coolant pump motor tripped due

to arcing in the' stator. This was apparently caused by a

foreign piece of metal resembling a washer which may have

damaged the stator insula. tion.

This appears to be an unusual,

isolated occurrence provided an electrical path to ground. The

motor has an open type enclosure.

The motor was replaced and

retested.

(2) The No. 1 Residual Heat Removal (RHR) pump tripped upon starting

due to an apparent upper wear ring failure. The applicant is

reviewing the problem as to cause and will report it as required

by the regulations.

(3) There were two cases of failure to maintain adequate procedural

control of plant conditions:

ISU-008, " Thermal Expansion" did not address the required

charging / letdown path and as such the paths were selected in

accordance with the plant operating procedures. As a result,

the lineup had to be changed to accommodate the test.

Plant

temperature stability, as defined in the test, was lost.

The

only consequence was about a 4-hour delay in reestablishing

stable temperature conditions which are prerequisites to the

test.

ISU-008 also failed to address the fact that RHR cross

connection valve 8716A was to be open for the test, because the

integrated plant operating procedure used to establish

conditions required the valve to be shut. When it became

apparent that the valve should be opened, verbal

miscommunications between test and operating personnel

resulted in a second delay in establishing stable temperatures

for the expansion test.

These problems were discussed with the ISU Coordinator, as well

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as TUEC management.

TUEC committed to ensure that all test

procedures will be checked and revised as necessary to identify

any valve or breaker positions required that are not normally

provided by the operating procedures referenced by the test

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p,rocedures.

The NRC inspectors did not observe this problem any

further and thus considered the corrective action of the

applicant to be' adequate in this area.

(4) Main Steam Isolation Valve (MSIV) cycling met the acceptance

criteria, after correcting minor mechanical problems pending

final review and. approval of the data, but the bypass valves

(MSIBV) did not.

The' applicant is evaluating and has made

informal comments to the RRI that a design change to manual

valves will be implemented.

(5) During thermal expansion ; testing, at the various temperature

plateaus, of ambient, 250*F, 350 F, 450*F, and normal operating

temperature, numerous support snubbers and restraints required

some rework and retesting.

By the end of the sequence most had

been corrected and retested, except in some cases where it was

impractical or unsafe (due to hot surfaces) to make adjustments.

The latter cases have been identified and will become part of

what now appears to be a potentially small test deferral package

for postfueling hot functional testing.

(6) During starting of the Turbine Driven Auxiliary Feedwater

(TDAFW) pump, in accordance with ISU-206A, the pump manual

discharge valve IAF-041 was discovered locked shut. This valve

was recorded by the operators as " locked open" on the valve

lineup sheet provided by the system operating procedure. The

RRI noted this in the shift supervisor's logs about one 8-hour

shift later,and cuestioned the shift supervisor in charge of the

subsequent shift whether or not a deviation report (DR) had been

initiated, as required by Administrative Procedure STA-404,

" Control of Deficiencies." The shift supervisor indicated that

he would check into it and if necessary, initiate the

appropriate reports.

He further stated that the cause appeared

to be confusion over which way to turn the valve handwheel due

to the reach rod linkage, and the valve being overhead, rather

than a violation of the system operating procedure.

IAF-041 is

an 8" rising stem overhead valve with several reach rod links

with universal joints to get the handwheel within easy access.

Tne RRI inspected the valve. Although it may be difficult to

check whether the valve is shut, the operator should not have a

problem checking whether the valve is open by looking for inward

stem movement because he can always attempt to shut the valve

and see some stem movement.

There is no apparent reason for

-

this problem, other than failure on the part of the operator

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to check the valve position in a positive manner.

Discussions

between the RRI and the applicant brought out a need for the

applicant to take definitive corrective action to preclude

future valvt lineup problems and to ensure that all such

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problems are documented in a timely manner by shift supervisors

with first-hand knowledge of the problems. At the discussion it was

pointed out by the RRI that the shift supervisor appears to be

burdened with an analysis of the problem and possible corrective

actions for the purpose of deciding in what format the problem

must be reported, i.e., Deficiency Report, Nonconformance

Report, or Problem Report.

These reports are controlled by

three different administrative procedures. The applicant

indicated that action will be taken to provide the shift supervisors

with simpler reporting instructions. Tne applicant has committed

to the above corrective actions. This is an Open Item

(445/8445-01).

.

(7) Pressurizer level indicator 1-LT-0460 did not compare favorably

with the redundant level channels as RCS pressure increased to

near normal operating pressure. Troubleshooting the piping for

the detector revealed a leaking drain valve which was tightened

thereby stopping the leak, but the reference leg needed to be

filled.

Upon attempting to fill the reference leg in accordance

with a generic " basic guidelines" procedure, an Instrument and

Control (I&C) Technician connected low pressure fill equipment

incorrectly to the detector piping and then operated the wrong

instrument valves.

This action resulted in the low pressure

fill tubing being blown off and the I&C technician received

thermal burns to his arm from hot reactor coolant.

The

personnel safety and postulated radiological implications of

this type of problem after initial criticality was discussed

with the applicant's representatives. As a result of this

discussion, Deficiency Report 84-127 was written.

Instruction

No. ICI-2007, " Filling and Venting Level Transmitters and Level

Indicating Switches (Wet Leg)" is not adequate to assure proper

controls over quality and radiological safety, and using such a

procedure is in violation of 10 CFR 50, Appendix B, Criterion V.

This is a Violation (445/8445-02).

3.

Review of Initial Startup Test Procedures

During the month of October, 1984, the RRI conducted a review of test and

administrative procedures to be used in the control of the Thermal

Expansion Test and other hot plant tests.

The results are listed in NRC

Inspection Report 445/84-39.

The RRI inspected the following procedures

during November 1984 to complete the review:

ISU-206A, " Auxiliary Feedwater performance" (Revisicn 2)

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ISU-282A, " Containment & F.W. Penetration Room Temperature

Survey" (Revision 1)

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Attributes checked included assurance that:

(1) the procedures were

consistent with regulatory requirements, (2) the procedures contained the

necessary administrative controls, (3) the test objectives would be met

and properly documented, (4) adequate Quality Assurance provisions were

incorporated as committed in the FSAR, and (5) there were no major

technical or editorial errors._

No violations or deviations were identified.

4

Verification of Completion of Human Engineering Deficiencies

The Human Factors Control Room Design Review of CPSES, conducted by the

Human.Facter Engineering Branch of the NRC, identified many Human

Engineering Discrepancies (HEDs).

NRC Inspection Report 445/84-31 reported that as of August 31, 1984, all

but 23 prelicensing HEDs had been closed by the Human Factor Engineering

Branch, and that the remaining HEDs will be verified by the resident

inspectors and documented in future inspection reports.

As of December 31, 1984, 20 of the 23 remaining HEDs have been verified by

the RRI as satisfactorily completed by personal observation of the

installed hardware.

There are now 3 HEDs remaining to be closed.

The

following is a listing of the HEDs remaining to be verified:

88.

HED DESCRIPTION

Trend recorder scale differs from chart paper scale.

ACTION

Confirmatory on recorders having paper matching recorder scales (all

recorders should have paper), including Hot Shutdown Panel (HSP).

Note: HED 122 was closed with exception of " proper paper in

recorders" which will be verified as a part of this HED.

181.

HED CESCRIPTION

The nuclear instrumentation system recorder lacks a scale for

differential power.

ACTION

Confirmatory on installation of a scale for differential power.

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184. HED DESCRIPTION

Counters require calculations by the operator when displayed values

run past 60 minutes.

Other counters require the operator to convert

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displayed values by multiplication factors other than a multiple of

ten.

ACTION

Confirmatory on full scale counters replacing 0.5 scale counters on

CPS-01.

The following is a listing of the HEDs that have been completed and then

verified by the NRC resident inspector:

3.

HED DESCRIPTION

Annunciator alarms are not visually prioritized.

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ACTICN

Verified completion of annunciator prioritization.

68.

HED DESCRIPTION

No storage space has been allocated for essential material.

ACTION

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Verified installation of portable storage unit and storage of

equipment at the HSP.

80.

HED DESCRIPTION

Pointers on "J" handle / star / handle switches contrast poorly with

handle color.

ACTION

Verified "J"

handle / star handle pointers being painted white.

93.

HED DESCRIPTION

No control coding is currently being used for:

o

Mechanical valves, pumps, b.eakers, motors, etc.

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o

Throttle valves

o

Emergency or critical controls

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ACTION

Verified installation of "T" handles on transfer switches at the HSP

(14 handles).

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106.

HED DESCRIPTION

.

Labels are missing.

ACTION

Verified labels on record.ers on CV-Oi, incore panel, and for lights

on CV-03.

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120.

HED DESCRIPTION

Sound powered jack communications are incomplete.

ACTION

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Verified storage of sound powered headset at the HSP (see no. 68

above).

122.

HED DESCRIPTION

The HSP is in the process of complete redesign.

ACTION

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Verified completion of Hierarchical labeling at HSP and transfer

panels, labeling of light box, proper paper in recorders (sea no. 88

above), and sound powered headsets at HSP (see no. 68 above), and

transfer panel.

Note:

" Proper paper in recorders" has not been completed.

This

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action was moved to HED no. 88 so that item 122 could be

closed.

130.

HED DESCRIPTION

Controls have unlabeled switch positions.

.

ACTION

Verified new escutcheon plates for 1-HS-2491 through 1-HS-2494 on

CB-09.

214.

HED DESCRIPTION

A rotary control with clockwise-counter clockwise movement is used to -

control a " lower" and " raise" function.

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ACTION

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Verified permanent escutcheon plates on CB-11 (90-1EG2 and 65-1EG2).

225.

HED DESCRIPTION

Thelockingpositionorfbnctionoftheverniercontrollersisnot'

clearly indicated.

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ACTION

Verified " LOCK" position labels on Hagan controllers.

226.

HED DESCRIPTION

Setpoint adjustment knob covers on process controllers can be easily

removed.

. ACTION

Verified more secure attachment of setpoint adjustment knob covers on

controllers.

267.

HED DESCRIPTION

Trend re: orders used frosted glass.

ACTION

Verified replacement of frosted glass with clear glass on recorders

on CB-10.

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HED DESCRIPTION

Annunciator character sizes are inconsistent.

ACTION

Verified re-engraving of annunciator tiles

1-ALB-2:

3.7

1-ALB-3B

2.6

1-ALB-4A

4.4

1-ALB-4B

1.5, 2.6, 3.6

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1-ALB-5B

2.1, 3.4

1-ALB-5C.

3.1, 4.2

1-ALB-6C

1.2, 1.3, 2.1, 2.2, 2.7, 3.2, 3.3, 3.7,

4.2

1-ALB-60

1.4, 1.10, 1.14, 2.4, 2.13, 2.14, 3.13,

3.14, 4.13

1-ALB-8

1.13, 2.13, 2.14, 3.14, 4.14

1-ALB-9

1.4, 1.8, 1.11, 4.1, 7.6

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345.

HED DESCRIPTION

Abbreviations in computer displays do not conform to those in the

Comanche Peak Steam Electric Station (CPSES) " Directory of Acronyms

and Abbreviations."

ACTION

,

Verified revision of point descriptions in P2500 to use CPSES

abbreviations.

5.

Preoperational Test Results Evaluation

The RRI reviewed the followir.g completed test package, ICP-PT-66-01,

" Nuclear Instrumentation System," which has been approved by the Joint

Test Group (JTG). Attributes inspected included: (1) adequacy of the

evaluation of test results, (2) assurance that test data met acceptance

criteria, (3) assurance that deviations were properly identified and

resolved, and (4) the applicant's administrative practices with respect to

test execution and data evaluation were adequate.

The test package met the attributes above, with one apparent exception. A

Test Procedure Deviation (TPD-03) was written to delete the requirement to

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take data in paragraphs 7.1.7.5 and 7.2.7.5 of the procedure, because the

source range meters that were installed in the Hot Shutdown Panel did not

function properly.

The meters did not have the proper signal input

ratings.

This was documented on a Test Deficiency Report (TDR 3014).

TDR-3014 stated that the retest would be per TDR-3547.

TDR-3547 was

written because when the proper meters finally were installed, they would

load down the circuit and cause erroneous readings. The retest specified

on TDR-3547 was lined out, leaving an open-ended paper trail.

The

applicant's representative has committed to take action to correct this

problem. This is an Open Item (445/8445-03).

No violations or deviations are apparent at this time.

6.

Acplicant Action on Previous Inspection Findings

a.

(Closed) Unresolved Item 445/8424-02: Apparent conflict between FSAR

Figure 6.3-5 and the Safety Injection Pump 01 performance curve.

During a previous inspection (445/84-24) of the completed test data

of preoperational test procedure ICP-PT-57-01, " Safety Injection Pump

Performance," the NRC inspector noted that the Safety Injection Pump

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01 performance curve in the completed test data did not meet the

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minimum acceptable performance curve of Figure 6.3-5 of the FSAR.

Since the time of the inspection, Amendment 53 of the FSAR (dated

November 5, 1984) changed the pump curve. The performance data and

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curves in the completed test package for ICP-PT-57-01 now meet the

requirements of the FSAR.

This item is closed.

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b.

(Closed) Open Item 445/8415-02: Minor discrepancies found during NRC

inspection of station administrative procedures.

During a previous

inspection (445/84-15) of station administrative procedures, the RRI

found several minor discrepancies and made some suggestions to

preclude future problems.

The applicant took action on those items

that the applicant consi.dered justification for implementing a

procedure change. .For example:

(1) STA-401, " Station.0peration Review Committee," Revision 5,

Section 4.4 did:not fully implement the responsibilities of the

committee as stated in the CPSES Unit 1 Technical Specifications

(final draft).

This was corrected in Revision 8 of STA-401.

(2) STA-203, " Control of Station Manuals," Revision 7, Section 4.3.3

required a notification memo to be sent to each onsite holder of

controlled station manuals to alert recipients of a revision or

new procedure.

This was not being done for holders of the

manual who incorporate their own changes because they sign a

receipt for the changes or new procedures anyway.

Revision 9

clarified this such that the applicant is in compliance with the

procedure.

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(3)

STA-307, " Forms Control," Revision 2, allowed minor changes to

forms without revising the parent procedure containing a sample

of the form as an attachment.

However, instead of changing the

revision number of the form itself, the office services staff

misinterpreted Section 4.2.6 of STA-307 and changed the revision

of the parent procedure attachment page, which caused a conflict

with the rest of the parent procedure pages.

This was corrected

by the applicant and STA-307 was revised to preclude

misinterpretation.

This item is closed,

c.

(Closed) Deviation 445/8415-01:

Failure of the applicant to use SORC

approved instructions to perform work on the emergency diesel

generators.

The CPSES FSAR commits to Regulatory Guide (RG) 1.22,

Revision 2, February 1978 with no exceptions. RG 1.33 and ANSI

N18.7-1976Property "ANSI code" (as page type) with input value "ANSI</br></br>N18.7-1976" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. to which it refers, requires maintenance to be performed

using procedures / instructions receiving the same review and approval

as operating instructions, i.e., review and approval by the 50RC.

During two previous inspections (50-445/84-07 and 50-445/84-15), the

NRC inspectors noted that the applicant had defined " instructions" as

procedures which do not require 50RC approval, and had issued

" instructions" to perform work on safety-related equipment such as

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the emergency diesel generator (EDG). The apparent basis was that

EDG work performed by the maintenance department had no significant

impact on other departments, and/or was work unique to the

maintenance department.

Since the above NRC inspections, the issue

has been resolved as evidenced in Station Administrative Procedure

STA-707, " Safety Evaluations," (Revision 2), STA-202, " Preparation,

Review, h,pproval, and Revision of Station Procedures" (Revision 10),

,

and the final draft'of the CPSES Unit 1 Technical Specifications

(TS).

In essence, all safety-related procedures and instructions

will receive a SORC review by virtue of the requirement that the SORC

review the related safety evaluations, as stated in the TS and

STA-401, which both list the responsibilities of the SORC.

This deviation is closed,

d.

(Closed) Violation 445/8421-02:

Failure of preoperational test

procedures to provide adequate prerequisites.

During a previous

inspection (445/84-21), the RRI noted that during conduct of

preoperational test 1CP-PT-29-02,RT1, " Diesel Generator (DG) Control

Circuit Functional and Start Test" the DG barring device was

connected to a portable air compressor instead of the Service Air

System.

There was no' prerequisite step in the test procedure to

'

provide either temporary or permanent air for the barring device, yet

it needed air to be tested.

Also, during testing of the Service

Water System in accordance with ICP-PT-04-01, RT 1, " Station Service

Water (SSW)," a Barton D/P gage did not function due to air binding.

There was no prerequisite in the test procedure to ensure the gage

was recently filled and vented to assure accurate test data, nor did

the Startup Administrative Procedures for writing the test require

it. This was a potentially generic problem.

The applicant has since

revised CP-SAP-7, " Format and Content of Test Instruction / Procedures"

to require the appropriate prerequisites.

Each organization

responsible for review of preoperational test procedures has been

instructed to ensure that test prerequisites receive a comprehensive

review to ensure system readiness and correct component configuration

to assure validity of the test results.

This item is closed.

7.

Plant Tours

During this reporting period, the SRRI and RRI conducted several

inspection tours of Unit 1.

In addition to the general housekeeping

activities and general cleanliness of the facility, specific attention was

given to areas where safety-related equipment was installed and where

activities were in progress involving safety-related equipment.

These

areas were inspected to ensure that:

Work in progress was being accomplished using approved procedures.

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Special precautions for protection of equipment were imolemented, and

additional cleanliness requirements were being adhered to for

maintenance, flushing, and welding activities.

Installed safety-related equipment and components were being

protected and maintained to prevent damage and deterioration.

. Also during these tours,'the SRRI and RRI reviewed the control room and

shift supervisors' log books.

Key items in the log review were:

-

plant status

changes in plant status

tests in progress

documentation of problems which arise during operating shifts

No deviations or violations were found.

8.

plant Status as of December 31, 1984

a.

The applicant was at the end of the Thermal Expansion Test sequence

and making preparations to roll the main turbine-generator.

Details

of the testing sequence and problems encountered are discussed in

paragraph 2 of this report.

b.

Unit No. 1 is 99% complete with 403 of 422 areas and 323 and 332

subsystems turned over to operations custody.

" Custody" means having

immediate authority and responsibility for operational control of

system or equipment.

The applicant has accepted 260 of 332 subsystems for final

acceptance.

c.

Of the 199 preoperational tests, one is not yet completed on field

testing, and 21 are pending review and approval of completed data.

Eighteen are pending NRC completed data inspections,

d.

Tne following items related to NRC resident operations office

findings are open pending applicant action and NRC followup

inspection to confirm completion of closure:

Violations

10

Deviations

0

Open items

100

Unresolved

7

Total

117

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Action is underway to complete these items.

Closure will be

documented in future inspection reports.

e.

Unit No. 2 is 65% complete. The preoperational test program on

systems associated with NRC inspections has not yet started.

9.

Exit Interview

.

An exit interview was conducted January 4, 1985, with applicant

representatives identified in paragraph 1.

During this interview, the RRI

and Mr. D. M. Hunnicutt of the Region IV NRC office reviewed the scope and

discussed the inspection findings.

The applicant acknowledged the

findings.

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