ML19253B660

Insp Rept 50-289/74-25 on 740603-06.Noncompliance Noted: Contrary to QA Plan,Makeup & Purification Sys & Decay Heat Removal Sys Were Lined Up W/Superseded Operating Procedures

ML19253B660
Person / Time
Site:	Crane
Issue date:	06/20/1974
From:	Davis A, Spessard R, Sternberg D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML19253B649	List: ML19253B648 ML19253B651 ML19253B653 ML19253B656 ML19253B658 ML19253B660
References
50-289-74-25, NUDOCS 7910160746
Download: ML19253B660 (20)

See also: IR 05000289/1974025

Text

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U.S. ATO !IC ENERGY COM:!ISSION

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DIRECTORATE OF REGUL\\ TORY OPERATIONS

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REGION I

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fic Inspection Report No:

50-289/74-25

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Docket No:

50-289

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Licensee:

Metropolitan' Edison Company

License No:

DPR-50

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Three lille Island Unit l'

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Priority:

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Category:

B-2

Location:

Middletown, Pennsylvania

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Typa of Licensee:

PWR 871 We (B&W)

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Type of Inspection:

Routine, Announced

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Dates of Inspect ~ ion:

June 3 through June 6, 1974

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Dates of Previous Inspection:

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Reporting Inspector

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R. L.,Spessard,' Reactor I.nspector

Date

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Accompanying Ir.spectors:

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D. M. Sternberg, Reactor Inspector

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.E.*I

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W. A. Ruhlman, Reactor Inspector

Date

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Date

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Date

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Other Accompanying Personnel:

A. B. Davis, Senior Reactor Inspector

(June 4&5,

Date

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Reviewed By:

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A. B. Davis, Senior Reactor Inspector

D'a t e

q9101607y[

Reactor Operations Branch

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SUFSIARY OF FINDIMCS

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Fnforcement Action

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Violations

1.

Contrary to the requirements of the FSAR Section lA, Operating

Quality Assurance Plan Section VI, the Makeup and Purification

System and the Decay Heat Removal System were lined up with

superseded operating procedures.

(Details 13.e. (1) and (2))

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2. Contrary to the requirements of the FS AR Section lA, Operating

Quality Assurance Plan,Section VI, the Waste Gas Disposal

System was lined up with a procedure-containing unauthorized

changes.

(Detail 13.e.(3))

Safety

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None Identified

.

Licensee Action on Previously identified Enforcement Items

Not Inspected

Unusual Occurrences

A.

Superseded valve checklists and startup checklists were used during

the course of plant startup.

(A0 74-7, lht Ed Telegram dated June

,

6, 1974 to Director, DRO, RO:I).

(Detail 13.e)

B.

A socket weld leak was discovered on the discharge header of the

reactor coolant system makeup pumps.

(A0 74-6, lht Ed Telegram

dated June 4, 1974 to Director, DRO, RO:1).

(Detail 5)

C.

A pinhole sized leak was discovered in the "B" Makeup Pump recircu-

lation flow orifice (A0 74-5, Fht Ed Telegram dated June 3,1974 to.

,

Director, DRO, RO:I).

(Detail 4)

D.

The "C" tbkeup Pump sustained severe damage during operation.

(Detail 6)

E.

The Sodium Thiosulfate Tank was diluted with Sodium Hydroxide during

recirculation of the Sodium Hydroxide Tank.

(Detail 7)

F.

A burned terminal was discovered on one of the control rod drive

power transformers.

(Detail 12)

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Other Sinnificant Findings

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A.

Current

1.

Initial Criticality

a

The inspectors witnessed the initial criticality event which

was accomplished in an orderly and controlled manner in

accordance with the procedure.

Criticality was achieved at

10:36 pm on June 5,1974, and the actual critical conditions

were in close agreement with the predicted values.

(Detail

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13.u)

2.

The Sodium Thiosulfata Tank is now operable.

(Detail 7)

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3.

Repairs and testing of the control rod drive system have been

completed.

(Detail 8)

"14.

Design changes to both Diesel Generators have been completed

using approved maintenance procedures.

(Detail 9)

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5.

The reactor . building sump and feeder drains were observed to be

clean and free of debris.

(Detail 10)

B.

Status of Previousiv Reported Unresolved Items

1.

The following items are resolved:

A procedure has been implemented which describes the

a.

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actions to be taken to indicate the status of out of

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servicc/ calibration gauges, meters and instruments.

(Detail 11)

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b.

The discrepancies noted in the implementation of the

surveillance t - program have been corrected.

(Detail 13.e)

The licensee's review and evaluation of the Diesel

c.

Generator circuitry design changes has been completed.

(Detail 9)

C.

New Unresolved Items

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None Identified

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Management Interview

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An exit interview was held onsite on June 6, 1974, at the conclusion of

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the inspection with the following attendecs:

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Metropolitan Edison Company

Mr. J. G. Herbein, Station Superintendent

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Mr. J. R. Floyd, Supervisor of Operations

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Mr. W. E. Potts, Supervisor of Quality Control

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Mr. J. J. Colitz, Station Engineer

Mr. J. P. O'Hanlon, Nuclear Engineer

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Mr. C. E. Hartman, Electrical Engineer

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Mrr R. L. Summers, Junior Engineer

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General Public Utilities Service Corporation

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Mr. R. J. Toole, Assistant Test Superintendent

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The following summarizes items discussed.

A.

Preoperational Test Program (Detail 2)

B.

Initial Startup Test Program (Detail 3)

C.

Followup Action - A0 74-6 (Detail 5)

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D.

Followup Action - A0 74-5 (Detail 4)

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E.

Follovup Action - A0 74-4 (Detail 8)

F.

Followup Action - A0 74-3 (Detail 9)

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G.

Sodium Thiosulfate Tank Status (Detail 7)

H.

Reactor Building Sump (Detail 10)

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Instrument Status Indication (Detail 11)

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Inspectors' Witness of Initial Criticality and the observed

violations (Detail 13)

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Loss of the "C" Makeup Pump (Detail 6)

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Burned terminal on control rod drive power transformer (Detail 12)

In each area listed above, the licensee representatives acknculedged the

inform. tion presented by the inspectors.

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DETAILS

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Persons Contacted On Site

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Metropolitan Edison Company (Met Ed)

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Mr. T. H. Acker, Control Room Operator

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Dr. T. S. Baer, Station Engineer

Mr. J. C. Banks, Control Room Operator

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Mr. M. L. Beers, Shift Foreman

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$2?? Mr. D. J. Boltz, Control Room Operator

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Mr. R. R. Booher, Auxiliary Operator

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Mr. T. L. Book, Control Room Operator

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Mr. N. D. Brown, Control Room Operator

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Mr. K. P. Bryan, Control Room Operator

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Mr. E. E. Bulmer, Mechanical Engineer

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Mr. H. L. Carr, Jr. , Auxiliary Operator

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Mr. J. J. Colitz, Station Engineer

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Mr. W. W. Cotter, Mechanical Engineer

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Mr. T. L. Crouse, Control Room Operator

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Mr. J. H. Deman, Radiation Chemistry Technician

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Mr. J. R. Floyd, Supervisor of Operations - Unit 1

Mr. A. C. Fredlund, Control Room Operator

Mr. B. C. Getty, Mechanical Engineer - Unit 2

Mr. C. F. Gilbert, Supervisor of Operations - Unit 2

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Mr. K. L. Harner, Radiation Chemistry Technician

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Mr. R. R. Harper, Instrumen't Foreman

Mr. C. E. Hartman, Electrical Engineer

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Mr. R. A. Heilman, Control Roca Operator

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Mr. J. G. Herbein, St.ation Superintendent

Mr. J. F. Hilbish, Project Engineer - Nuclear

Mr. G. R. Hitz, Control Room Operator

Mr. R. S. Hutchison, Auxiliary Operator

Mr. I. P. Hydrick, Shift Foreman

Mr. C. W. Keyser, Control Room 6perator

Mr. B. A. Mehler, C]ntrol Room Operator

Mr. R. P. }Eller, Test Coordinator

Mr. L. G. Noll. Control Room Operator

Mr. J. P. O'Hanlon, Nuclear Engineer

Mr. W. E. Perks, Shift Foreman - Unit 2

Mr. J. F. Peters, Administrator

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Mr. E. L. Pilsitz, Control Room Operator

Mr. R. H. Porter, Shift Supervisor

Mr. C. E. Randolf, Engineering Assistant

Mr. R. A. Rice, Security Specialist

Mr. M. J. Ross, Shif t Supervisor

Mr. D. M. Shovlin, Supervisor of Mait:t*7ance

Mr. B. G. Smith, Shift Foreman

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Mr. J. R. Smith, Shif t- Supervisor

Mr. R. L. Summers, Junior Engineer

tk. J. H. Thomas, Coordinator of Services

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Mr. R. P. Velez, Radiation Chemistry Technician

Mr. G.i C. Wallace, Shif t Supervisor

}k. D. E. Weaver, Instrument Foreman

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Mr. H. L. Wilson, Instrument Foreman

Mr. W. H. Zewe, Shift Foreman

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Ceneral Public Utilities Service Corporation (CPUSC)

5[N1- Mr. J. J. Barton, Startup and Test

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..ac i Mr. W. H. Behrle, Shif t Test Director

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Mr. L. E. Eriggs, Startup Engineer

Mr. H. C. Eisnaugle, Project Administrator

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Mr. C. E. Gatto, Shif t Test Engineer

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Mr. T. M. Hawkins, Shift Test Engineer

Mr. G. P. Miller, Test Superintendent

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p=, Mr. M. A. Nelson, Technical Engineer

- Mr. S. G. Poje, Shift Test Engineer

Mr. I. D. Porter, Shif t Test Engineer

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Mr. J. J. Stromberg, Site Auditor

Mr. R. J. Toole, Assistant Test Superintendent

Mr. J. E. Wright, Site QA Manager

Mr. L. M. Zubey, Mechanical Engineer

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Babcock and Wilcox Nuclear Services

Mr. J. H. Flint , Shift Data Analyst

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Mr. R. A. Govers, Engineer

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Mr. E. R. Kane, Engineer

Mr. J. A. Middleton, Engineer

Mr. M. B. Owen, Engineer

Mr. W. E. Schaid, Shift Data Analyst

NUS Comoany

Mr. J. D. Trotter, Consultant

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2.

Preoperational Test Program

The inspector reviewed TWG bbeting minutes for the period iby 29 -

June 4, 1974 (Meetings 110-115), completed test results (of ficial

Field Copy), and the approved Prerequisite List for Initial

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Criticality.

Additionally, the licensce's written safety evalua-

tions made pursuant to 10 CFR 50.59 for not completing certain

preoperational tests as described in the FSAR were reviewed on a

sampling basis (included TP 120/1, 2 and 3, TP 360/lA, TP 230/8, SP

102.6 and F1104/6).

The following information was obtained:

a.

Prooperational Testing Status

Tests Completed and Accepted

88%

Tests Partially Completed and Accepted

8%

Tests Not Completed

4%

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The licensee evaluated the overall results of the preopera-

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tional test program and concluded that the program was suf-

ficiently complete to support (relative to safety) initial

criticality.

Tests not to be completed prior to initial

criticality were previously reported in Report 50-289/74-23,

Detail 7.a.

Additionally, the following tests were not

completed:

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(1) TP 600/25, Hydrogen Addition Test

(2) TP 600/26, Degasification Test

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Tests (1) and (2)' above are to be deleted from the MTX

on the baeis that the test objectives were accomplished

by Met Ed Operating Procedures.

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(3)

TP 271/4, l'ain and Auxiliary System Functional Test

(4)

TP 600/11, Emergency Feed System and OTSC Level Control

Test

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Tests (3-) and (4) above have been partially ccmpleted and

accepted by the licensec.

The outstanding items relate

to the turbine driven cmcrgency feedwater pump which was

inoperable (needed new shaft).

Testing to be completed

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prior to reaching 5% of full power (Tech Specs require

this pump prior to reaching 5% of Full Power).

(5) TP 600/30, Hot Functional Test Checkpoints

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This test had been partially completed and ecepted by the

licensee.

The outstanding item related to the Ibin Steam

Isolation Valve Support System which did not have contact

on every valve because design temperature had not been

reached.

Testin6 to be completed prior to reaching 5% of

full power.

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The inspector has no further questions relative to the pre-

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operational test program at this time.

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b.

Test Results

The inspector verified that the test results for the proce-

dures listed below had been accepced by the licensee unless

otherwise indicated.

Included in the inspector's review (on

a sampling basis) was the licensee's disposition of identi-

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fied test exceptions and deficiencies and comparison of test

data with acceptance criteria.

No deficiencies were identi-

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(1) TP 151/1, Reactor Building Isolation Valve Leak Tests

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(2)

TP 600/4, Makeup and Purification System Operational Test

The inspector observed that the letdown isolation valve

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(MU-V3) had been successfully retested.

This item is

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closed.

(Report 50-289/74-08, Detail 6.a)

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(3)

F1104/15A, Auxiliary Building and' Fuel Handling Building

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Ventillation Functional Verification

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(4)

TP 360/2, Area Radiation Monitors Calibration and

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Functional Test

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(5)

TP 310/3, E.S. Detection and Actuation Test

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(6)

TP 305/2, Reactor Protection System Response Time Test

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(7)

SP 320/4, ICS - Turbine Interface Test

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(8)

TP 301/2, N.I. Detector Cabling and Response Tests

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(9)

TP 301/3D, N.I. Determination and Setting of Detector

Voltages

(10) SP-2, Filter Efficiency Tests

(11) TP 360/lB, Process Radiation Monitors Calibration and

Functional Tests - Atmospheric

(12) TP 360/1C, Procc as Radiation Monitors Calibration and

Functional Tests - Liquid

(13) TP 600/5, Nuclear Chemical Addition and Sampling System

Operational Test

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(14) TP 600/28, Controlling Procedure for Hot Functional

Testing

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(15) TP 600/13, Pressurizer Operation and Spray Flow Test

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Results cf this test had been partially accepted *.iith one

open item relative to setting the pressurizer spray bypass

valve.

This item was accomplished prior to approach to

initial criticality.

The inspector observed that the above tests had been cerducted

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and accepted in accordance with the licensee's commitments to

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RO:1.

(Met Ed letter dated April 19, 1974 to the Director,

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DRO, RO:I and Reports 50-289/74-17, Detail 5.b and 50-289/74-18,

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Details 2 and 3).

3.

Initial Startuo Test Program

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Procedures

The inspector reviewed the following procedures and verified

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that previous RO:I comments had been included in accordance

with the licensee's commitments.

(Report 50-289/74-18,

Detail 7.b. Items (5), (9) and (10))

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(1)

TP 800/5, Reactivity Coefficients at Power

(2)

TP 800/32, Loss of Offsite Power Test

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(3) TP 800/34, Generator Trip Test

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b.

Test Results

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(1)

Detailed Review by RO:I

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The. inspector conducted a detailed review of the completed

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procedures (Official Field Copy) listed below.

The

licensee had accepted the results of these tests.

Test

results met the acceptance criteria, and all test require-

ments were satisfactorily performed.

(a)

TP 200/11D, Reactor Coolant Pump Flow Test - Hot,

,

With Core

The inspector observed that final balancing of the

reactor coolant pumps had been completed and that

data indicated frame and shaft vibrations met the

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accept.7nce criteria'.

This item is closed.

(Report

50-2S9/.74-08, Detail 6.f)

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(b) TP 200/12D, Reactor Coolant Flow Coastdcwn Test -

Hot, With Core

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(c) TP 330/5, CRD Trip Test

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The inspector observed that TCN-2 to this procedure

was approved by the TWG on >by 29, 1974, in accor-

dance with the requirements of Test Instruction

No. 9.

This TCN, which permitted tripping the

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fastest and slowest rods at the same time by re-

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patching (Hot, No Flow Trip Test), was made to

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provide easier control of reactor coolant system

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temperature by minimizing reactor coolant pump

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starts and steps.

These tests and the Hot, Full

Flow Trip Tests were conducted following repairs

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to the damaged CRD system equipment (A0 74-4).

(2) Verification by RO:I

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The inspector verified by review of the completed pro-

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cedure (Official Field Copy) that the results of TP 302/

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2C In-Core Detector Post - Installation Electrical Test

had been accepted by the licensee.

c.

SP 710/2 - Controlling Procedure for Post Core Load Precritical

Testing

The inspectors reviewed this procedure (Official Field Copy)

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which was being conducted and which was a prerequisite for

beginning the approach to initial criticality (TP 710/1).

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The TWG reviewed the status of this procedure prior to sign-

off of the Prerequisite List for Initial Criticality, and the

inspectors' verified that the outstanding items frca this pro-

cedure, as identified in the Prerequisite List, were completed

prior to starting TP 710/1.

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The results of heat loss test for the pressurizer were discussed

with licensee representatives who provided the following infor-

mation:

Modifications to the insulation showed no f mprovements

over previously observed conditions.

The excessive heat loss

was not considered an operating problem; however, for econcay

reasons the matter was still being reviewed.

Valve RC-V2 was

operating satisfactorily; however, a larger motor crerator is

to be installed.

The inspector has no further questions on

this matter at this time.

(Report 50-289/74-08, Detail 6.c).

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4.

Followuo Action A0 74-5

During a previc-

RO inspection (Report 50-289/74-23, Detail 2), the

inspector veri: 2d that the licensee had established a program for

inspecting bbkeup and Purification System Orifices.

The licensee

reported by telephone on May 31, 1974, and by telegram on June 3,

1974, that a pinhole sized leak had developed in the "B" Fbkeup

pump recirculation flow orifice.

This pump was in service and the

reactor coolant system was in a heat-up mode (utilizing Reactor

Coolant Pump heat) at the time the leak was discovered.

The "A"

- - Makeup Pump was put in service, the "B" Makeup Pump was secured,

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During this RO inspection the inspector reviewed the actions taken

by the licensee relative to this occurrence.

Included in the in-

spector's review were the following:

written maintenance procedure

(WA 563) including the retest requirements (operational hydro per

.,

USAS B31.1.0), weld history records, welder qualification records,

~^".NDT examination records, X-rays of other orifices in the system,

~ visual inspection of the repairs with the system in operation, and

the licensee's plans relative to the other orifices in the system.

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No deficiencies were identified.

The following actions had been

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taken or were planned:

A replacement orifice (from Unit 2) was installed and the "B"

a.

Makeup Pump returned to service.

b.

X-rays were taken of the recirculation orifices for the "A" and

"C" Makeup Pumps and the letdown block orifice.

X-rays indicated

severe hydraulic erosion had occurred to the orifice for the "A"

Makeup pump; no evidence of erosion to the orifice for the "C"

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, Makeup pump, and slight erosion to the letdown block orifice.

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Run times for the Makeup pumps (from electrical switchgear in-

c.

stalled instrumentation) were as follows:

"A" - 1684 hours0.0195 days <br />0.468 hours <br />0.00278 weeks <br />6.40762e-4 months <br />,

"B" - 1141 hours0.0132 days <br />0.317 hours <br />0.00189 weeks <br />4.341505e-4 months <br />, and "C" - 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />.

d.

As a result of Items b and c above, the "A" Fbkeup pump was

valved out, the "C" tbkeup pump was running, and the "B" Make-

up was lined up but not running.

The "C" and "B" Makeup

pumps were selected for E.S. mode per Technical Specification 3.3.1.1.b.

However, during this R0 inspection the "C" Makeup

pump had to be removed from service (refer to Detail 6), and

"A" and "B" }bkeup pumps were put in the E.S. mode with the

"B" Makeup pump operating.

Additionally, operating instruc-

tions containing appropriate precautions relative to the

"A"

pump were issued.

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The UT inspection program which was being implemented per B&W's

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recommendations did not detect the leak (actually discovered

by an operator) because the leak occurred in an area -(thru the

wall at tie last orifice plate) not being examined.

The area

bedng examined was upstream of the orifice assembly rather

than on ene downstream side which was the location of'the

a

leak previously discovered at A50-1.

The licensee intends to

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institute a weekly radiography examination of the recircula-

tion orifices for the

"A",

"B", and "C" Makeup pumps and the

letdown block orifice.

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Orifices of a new design had been ordered, and two of these

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(for the Makeup pumps) were expected to arrive on site by

June 10, 1974.

klans were to install these orifices on a

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priority basis on the "A" and "C" Makeup pumps.

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5.

Followup Action AO 74-6

The licensee reported on June 3,1974, during this RO Inspection

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and by telegram on June 4, 1974, that a leak was discovered in a

socket weld on the discharge header of the reactor coolant system

makeup pumps.

The leaking weld joint was located in a 1" 304

stainless steel l'ine which supplies makeup water to the core flood

tanks.

The reactor coolant system was at 350cF and 1000 psig when

the Icak was discovered.

The secondary system was bottled up,

the reactor coolant pumps and the operating makeup pump stopped,

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and the leaking weld isolated to make repairs that were completed

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on June 4, 1974.

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The inspector reviewed the written maintenance procedure (WA 570)

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including the retest requirements (operational hydro per USAS

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B31.1.0) and Welder Qualification Records.

The repair consisted of

grinding out the defect, laying a fillet weld in the ground out

area until the area was filled up, and making a cover pass over the

entire weld to provide additional strength.

The inspector made a

visual inspection of the weld following repairs and while the

system was in operation.

Additionally, the inspector had dis-

cussions with GPUSC - QA representatives concerning this f ailure,

the repairs and PT examination.

The following information was

obtained:

This was the first socket weld failure in a stainless

,

steci joint.

The weld records for the original weld were reviewed

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and no abnormalities were identified.

The makeup and purification

system had been successfully hydroed to 4575 psig per USAS B31.1.0

(1.5 times design pressure).

The cause of the failure was unknown,

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but was surmised to be possibly insufficient weld metal.

Following

repairs the PT cxamination revealed no indications, as determined

by a Level III inspector.

6.

Loss of t'ne "C" lbkeuo Pump

At approximately 1:29 AM on June 5,1974, the "C" Makeup Pump which

was operating, was manually stopped and the "B" Makeup Pump was

started.

The plant was being maintained at 5320F + 20F with re-

actor coolant pump heat.

Control rod groups 1 thru 4 (Safety Rods)

were fully withdrawn with the remaining control rods (groups 5 thru

....

- .- M S-8) fully inserted, and reactor coolant system boron concentration

"2E~was about 2070 ppm.

Preparations were underway for the initial

critical approach with systems / components being aligned in accor-

dance with an approved prerequisite checkoff list.

The control

room operator received a " Low Lube Oil Pressure" alarm on the "C"

Makeup Pump.

The same pump was " punched up" on the computer con-

sole and a high bearing temperature was indicated by the computer.

The operator manually started the "B" Makeup Pump and secured the

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~<:--

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"C" pump.

Subsequent checkout of the computer memory indicated

that the total clapsed time from oil pressure alarm to securing the

pump was nineteen (19) seconds.

Flow to the reactor coolant pump

seals was maintained at all times.

In order to maintain the elec-

.

trical separation criteria.and the.needs of the emergency makeup

capabilities, the "B" pump was later switched to the alternate

power supply and valves were aligned.

The elapsed time frca oil

- pressure alarm until all electrical / valve ~ lineups were completed

was three (3) hours.

The inspector accompanied the operator out-

side the control room and followed most of the valve lineups and

electrical switching.

In all cases, the operator appeared to

-

follow all applicable safety regulations and approved valve lineup /

+

electrical switching procedures.

Severe damage occurred to the "C"

pump, and the pump was being replaced with an identical pump from

Unit 2 when this L3 inspection ended.

In the Unit 1 design there

are 3 Makeup Pumps; however, only 2 of these pumps are required to

be' operable in the E.S. mode powered from independent separate

busses to make the reactor critical.

This condition was satisfied.

7.

Sodium Thiosulfate Tank Ooerational Status

During a previous RO inspection (Report 50-289/74-23, Detail 3),

this matter was reviewed.

During this RO inspc ction, the inspector

observed that the installatten of the tank insulation was essentially

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complete.

The inspector also determined that installation and test-

ing of the heat tracing (lower section) had been completed.

The

upper section of the heat tracing has been installed; however, cir-

cuitry design changes are to be made to this section to permit its

actuation as determined by tank temperature.

According to licensee

representatives, the lower section is sufficient to maintain a tank

temperature of >75 F, and the upper section would be actuated on, as

required during severe cold weather conditiens.

'

.,_ At the time of this inspection the tank contents were out of specs

Prf as a result of dilution by Sodium Hydroxide.

This occurred as a

'

biEiresult of leakage thru a normally closed valve (BS-VS1) during re-

circulation of the Sodium Hydroxide Tank.

This condition was identi-

fied as a result of sampling both tanks.

As an interim measure,

operating procedures have been revised to require removal of the

spool piece (between the normally closed valve and associated tank)

in the supply lino to the tank not being circulated.

The licensee

-- is evaluating a sy stem modification, that consists of installing

-

a second isolation valve in the supply line to each tank, to correct

this apparent desige deficiency.

The licensee plans to drain the

tank, and add a new batch of Sodium Ti.losulfate to bring the tank

contents back into specs; a licenser representative stated that

RO:I would be notified when this m olution was completed.

RO:I was notified by the licensee by telephone on June 14, 1974,

that the contents of the Sodium Thiosulf ate Tank met the require-

ments of Technical Specification 3. 3.1. 3 (e. g. , 35,800 pounds of

sodium thiosulf ate at 29.5 weight percent solution with 1.5 weight

percent boric acid - sodium hydroxide buffer at a pH of 9.2).

The

--

licensee was informed that R0:I had no further questions on the

operability of this tank and that R0:I would reconmend to DL that

License Condition 2.c. (1) be cleared.

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8.

Followup Action - A0 74-04

The licensee's ten day letter for this occurrence had not been re-

ceived at the time of this inspection; however, the corrective

action to restore the Control Rod Drive System to operational

status following the fire in a CRD cabiaet was reviewed.

This

consisted of a review of selected maintenance procedures used to

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repair the damage and a review of rod scram time test dat- obtained

after the repairs.

R0:1 has no f ttrther questions in this area at

this time.

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9.

Followup Action - A0 74-03

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The licensee stated in a letter dated May 30, 1974, that tha

followup action to correct the condition of the 13 Emergency Diesel

'

Generat' r failure to run was to generate an approved procedure to

o

repair, test and restore to service and brief personnel inv'olved in

the need to use procedures.

The inspector reviewed the approved

' procedure used to restore the 1B Diesel and also the procedure used

to modify the 1A Diesel Generator.

The inspector also discussed

with a licensee representative the briefing of the maintenance

3;gpersonnel and verified such training was conducted.

RO:I has no

'

this-time.

(9g: further questions in this area at

'

The design change evaluation (GPU Startup Problem Report 975 com-

,

pleted on June 3, 1974) was reviewed by the inspector.

This report

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reevaluated the design adequacy of the circuit changes based on

quescions raised during a previous RO inspection.

The inspector

has no further questions on this matter, and the item is resolved.

~

(Report 50-289/74-18, Detail 13.m)

10.

Reactor Building Sumo

_

The inspector physically inspected the Reactor Building sump and

several feeder drains.

These appeared to be clean and free of

debris.

This item.is closed.

(Report 50-289/74-18, Detail 12)

11.

Instrument Status Indication

'

A deficiency in the licensee's program used to indicate the opera-

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1<:. bility and/or calibration status of meters, gauges and instrunents

~

was identified in Report 50-289/74-21, Detail 3.

The licensee has

written and implemented a procedure and tagging system whereby out

of calibration meters, gauges and instruments are identified and

marked.

The inspector revieced two (2) different instrucents which

had been marked in accordance with this procedure.

Temporary tags

are being used, but permanent black and white stickers have been

ordered according to a licensee representative.

RO:I has no further

questions in this area at this time.

12.

Burned Terminal on Control Rod Drive Power Transformer

on June 3, 1974, the licensee reported an observed imbalance ia

the phase currents of the CRD system.

The problem was traced

to a burned terminal on the primary of one of two main transfor-

mers for the system.

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The inspector observed the work in progress and observed that a pro-

cedure was being used and Q.C. coverage existed.

The licensee's

representative stated all similar terminals were checked and found

to be tight.

Additionally, he stated no safety issue was involved

in this occurrence.

The inspector observed the return to service

of this equipment prior to initial criticality and has no furtner

questions on this matter.

13.

Inspectors' Witness of Initial Criticality

Certain activities, procedures, and documents of the licensee were

-

EjyLireviewedbytheinspectorstoverifycompletenessofrequirements

u.qrfor initial cciticality.

The following is a list of specific topics

reviewed and the inspector's comments relating to. these items:

a.

Criticality Procedure: Official copy was current and had all

revisions incorporated.

.

.b.

Primary Chemistry: Most recent samples indicated chemistry

,

within Technical Specification limits.

.

c.

Gaseous and Liquid Effluents Monitoring:

The installed effluent

monitoring system was observed to be operating and within the

current calibration cycle.

d.

Chemical Factor Analysis:

The inspector witnessed a Met Ed

Radiation Chemistry Technician perform the determination of

the primary coolant boron concentration.

This activity was

accomplished using the correct procedure.

A one time sample,

.

specially diluted by the Unit 2 Station Engineer was passed

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to the Radiation Chemistry Technician as a normal sample.

It was correctly analyzed and its off-nominal concentration

. correctly identified.

.

e.

Procedure Prerequisites: A sample of Surveillance Test Pro-

cedures and system Operating Procedures, which were required

to be completed by the Initial Criticality Procedure (TP 710/

1), were reviewed.

No deficiencies in the results of the

Surveillance Tests were noted.* The following deficiencies

were observed during the review of the operating procedures:

• Deficiencies in the licensee's Surveillanca Test Program were pre-

viously identified in Report 50-289/74-23, Detail 21.

The licensee

had conducted a comprehensive review 3 and the inspectors noted the

identified deficiencies had been corrected.

RO:1 het no further

questions in this area at this time.

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(1)

The Makeup and Purification System was being lined during

preparations for initial criticality in accordance with

OP 1104-2.

The Control Room Working Copy used to acccm-

plish this activity was observed to be Revision 1 dated

February 1, 1974.

The Control Room File Copy was ob-

served to be Ravision 2 dated April '- 1974.

Subsequent

review of the Procedure Index indicated that Revision 3

to this procedure had been distributed on Fby 24, 1974.

(2)

The Decay Heat Removal System was lined up during pre-

parations for initial criticality in accordance with OP

r,m_.

~{'(.} ~

1104-4.

The valves in this system wers positioned per

Enclosure 1 to OP 1104-4 Revision 0 on May 30,1974.

It

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was observed that the Control Room Working and File

copies of this procedure, including Enclosure 1 were

Revision 2 approved April 27, 1974 and placed in the

Control Room files on May 16, 1974.

The above are in violation of the FSAR 'Section lA, Opera-

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ting Quality Assurance Plan,Section VI, Document Control,

which states in part:

". . .The Generation Division docu-

ment control procedure further requires that each >bnager

---

and Station Superintendent provide in their procedures

measurec: ... to ensure that approved changes be pronptly

transmitted for incorporation into documents; and to

ensure that obsolete or superseded documents are elimin-

ated from the system and not used..."

(3)

The Waste Gas Disposal System was lined up during pre-

>

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parations for initial criticality in accordance with

OP 1140-27.

It was observed that the Control Room Werk-

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ing Copy which had been used to accomplish this setup

had pen and ink changes to certain valve, level indicator

and electrical panel designators on pages 5 and 6.

There

was no indication that.these changes were approved by any-

one.

It was also noted that these changes did not appear

in the Control Room File Copy of OP 1104-27.

This copy

was also observed to be missing pages 22, 23, 24, 25, 26

and 29.

The above is in violation of the FSAR Section lA, Opera-

ting Quality Assurance Plan,Section VI, Document Control,

which states in part:

"The Superintendent of the TMI

Station Unit 1 will ensure that no changes are made to

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site instructions, procedures, and drawings unless such

changes are approved by the appropriate approving organi-

zation ..."

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These items were brought to the attention of the Plant Superinten-

'

dent who directed a review of the revisions of the procedures used

to align all safety related systems during preparations for initial

criticality.

As a result of this review the licensee stated all

j

valves affected by differences in the current and superseded check-

lists were correctly positioned prior to beginning the approach to

,

._,;,, initial criticality.

The Plant Operations Review Committee (PORC)

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'0, , s evaluated this occurrence as an abnormal occurrence, and it was

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reported as A0 74-7.

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f.

Required Plant Systems in Service:

The systems required by

the prerequisite lists were reviewed on a sampling basis and

were found to bc operational or noted exceptions had been

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reviewed and determined to have no impact on criticality.

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g.

Calibration Checks of Instrumentation:

A sample of safety

related system surveillance tests were reviewed and found

_.

to be current and complete.

h.

Source Checks c f the Source Range Instrumentation:

The

response of the S.R. Channels to the installed neutron

sources was observed.

i.

Signal to Noise Checks:

The licensee took an exception to

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the requirement in TP 710/1 to verify the source range signal

to noise ratio greater than 2.

This exception was based on

a preoperational test accomplished prior to fuel loading.

j.

Minimum Count Rate on Each Channel:

The inspector observed

an indicated count rate in excess of the minimum required

(>2 ' CPS) at the tite Control Rod withdrawal ccamenced.

k.

Reactor Trip Settings Lowered: The inspector reviewed the

most current surveillance test of Channel C of the Power

Range Instrumentation and observed that the trip setting

was lowered in accordance with the Test Change Notice to

the Power Range Surveillance Test.

1.

Increased Channel Sensitivity of Instrumentation:

The

inspector reviewed the most current Surveillance Test of

Channel C Power Range Instrumentation and observed that

the gain had been increased in accordance with the Test

Cha'nge Notice to the Power Range Surveillance Test.

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Special Instrumentation In Service:

The inspector observed

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m.

the counter-scalers and stop watches required by the pro-

}

cedure appeared operabic.

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Data Properly Recorded:

The inspector observed that data

required by the procedure being entered into the data

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sheets,

Data Properly Presented for Analysis:

The inspector observed

o.

boron concentration data and count rate data being presented

to the data takers and analysts for evaluation.

~c~..

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Analysis Performed:

The inspector observed the generation

of the 1/M plots and verified several points independently.

Additionally, the dilution rate was checked using stop

watch and the installed flow meter.

q.

Adequate Staff: The inspector reviewed a Test Manning memo

prepared by the Plant Superintendent and found it met the

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requiremants of the Technical Specifications for this test.

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SRO's and R0 at the Required Location: The inspector observed

-

the number of licensed operators required by the Test Manning

Memo were on watch during the procedure.

s.

Maximum Crew Working Time:

A representative of the licensee

informed the inspector that the maximum shif t being worked

j

by the personnel directly involved with the initial criticality

.

was twelve (12) hours.

i

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Shift Turnover Performed in an Orderly Manner:

The inspector

attended several crew turnovers and test briefings.

These

were conducted in an orderly mann;r.

Orderly and Systematic Use of Procedure: The insp.: tor wit-

u.

nessed the initial criticality evolution and observed that

the evolution was conducted in accordance with the procedure

in an orderly manner.

The following is a brief summary of the sequence of events observed

during the actual evolution.

All times listed were on June 5, 1974.

The procedure commended at 2:30 pm.

Plant conditions were as

follows.

Temperature 532 120F, primary pressure 2155 1 25

psi, Control Rod Assemblics (CRA) Croups 1-4 fully withdrawn,

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4 Reactor Coolant Pumps running, primary Boron Concentration

was 2070 ppm.

At this time Groups 5, 6, 8 were fully with-

drawn and Group 7 pulled to 75%.

Rod withdrawal was com-

pleted at 3:35 pm and boron dilution of the primary commenced

at 4:56 pm at a rate of 50 1 10 GPM.

Dilution was secured at

10:27 pm, and the letdown flow was increased to about 140 GPM.

The reactor was declared critical at 10:36 pm with a boron

concentration of 1623 ppm compared to the acceptance "alue

of 1626 150 ppm.

The power level was stabilized with rod

motion and necessary data was collected.

_Eh.fy.

Procedural Signoff Completed:

The inspector observed the sign-

[

off of the prerequisites for the procedure and the appropriate

r

checklists.

w.

Data Analysis Performed in a Timely Manner:

The maintenance

of the 1/M Plots and estimated critical condition updates were

observed to be performed in accordance with the procedure which

provided adequate time to ensure control of the evolution.

'- -

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Criticality Achieved Near Predicted Value: The licensee's data

-

indicated the critical concentration of boron was 1623 ppm com-

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pared to a prediction of 1626 150 ppm, with all other specified

reactivity sources as required by the procedure.

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