ML20197C249

Insp Rept 50-423/86-09 on 860218-0314.Violation Noted: Failure to Implement Defined QC Surveillance Program

ML20197C249
Person / Time
Site:	Millstone
Issue date:	05/02/1986
From:	Eselgroth P, Prell J, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20197C228	List: ML20197C222 ML20197C243 ML20197C249
References
50-423-86-09, 50-423-86-9, NUDOCS 8605130275
Download: ML20197C249 (13)

See also: IR 05000423/1986009

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-423/86-09

Docket No. 50-423

License No. NPF-49

Licensee: Northeast Nuclear Energy Company

P. O. Box 270

Hartford, Connecticut 06141-0270 -

Facility Name: Millstone Nuclear Power Station, Unit 3

Inspection At: Waterford, Connecticut

Inspection Conducted: February 18-March 14,1986

Inspectors: .' 1h [

J rell, Reactor Engineer / daYe

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P. Wen,RactorEng/neer

Approved by: -

5/2./86

F. Ese'Tgrot , Chief, Test Programs 'ddte

Section, , DRS

Inspection Summary:

Routine Unannounced Inspection Conducted On February 18-March 14, 1986

(Report No. 50-423/86-09)

Areas Inspected: Startup program review, power ascension test procedure

review, test results review, test witnessing, and review of licensee actions

on previous findings.

Results: One violation was identified (failure to implement the defined QC

surveillance program).

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DETAILS

1. Persons Contacted

• G. Clossius, QA Supervisor, NNECO

+J. Crockett, MP-3 Unit Superintendent, NNECO

P. Finck, Startup Engineer

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E. Fries, Startup Engineer, NNECO

-*J. Jensen, QA Specialist, NNECO

r E. Laware, QA Engineer, NNECO

, C. Libby, Operations QA Supervisor, NUSCO

• +D. McDaniel, Reactor Engineer

D. Miller, Startup Manager, NNECO

D. Moore, Assistant Operations Supervisor, NNECO

M. Pearson, Assistant Operations Supervisor, NNECO

, +W.-Richter, Assistant Startup Supervisor, NNECO

U.S. Nuclear Regulatory Commission

+F. Casella, Resident Inspector

T. Rebelowski, Senior Resident Inspector

+T. Shedlosky, Senior Resident Inspector

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I + Denotes those present at nini exit meeting on February 28, 1986

• Denotes those present at the exit meeting on March 14, 1986. The

inspectors also interviewed other personnel during this inspection period.

2.0 Licensee Action of Previous Inspection Findings

(0 pen) Unresolved Item (50-423/85-34-03) pertaining to qualification of

the containment High Range Radiation Monitor, Kaman Model KDI-1000 and

< Mineral Insulation Cable Assembly.

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Qualification of Model KDI-1000 High Range Containment Area

Radiation Detector and Mineral Insulation Cable Assembly. Report

No. 46-0036-001, Revision A.

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Qualification Report for Model KDI-1000 High Range Containment Area

Radiation Detector and Mineral Insulation Cable System. Report No.

460036-002, Revision 2.

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Installed Specification for Mineral Triaxial Cable Penetration

Assembly. Report No. 460036-002, Revision A.

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Installation Specification for Mineral Insulation (MI) Cables for

High Range In-Containment Area Monitors.

In reviewing the above documents the inspector concluded that additional

test data needs to be made available for review in order to conclude that

the Containment High Range Radiation Monitor, Kaman Model KDI-1000 and MI

Cable Assembly, will function as required in a LOCA environment.

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The basis for this determination is the numerous deviations, anomalies

and failure reports (DAFR) identified in the qualification report No.

46C036-002, Revision 2 which reduced acceptance criteria for problem

areas.

For example;

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DAFR-9, -10, -13 reduced torque cycle requirements in an attempt to

correct relaxation problems of torqued connectors.

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DAFR-14 deletes the voltage withstand test of signal cables as a pre-

seismic functional test.

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DAFR-16 requires the chemical spray to be applied only during non-

superheated conditions.

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The thermal transients behavior effect inside the cable during rapid

temperature transients add a false signal to the radiation signal.

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There are numerous references to test failures attributed to

moisture intrusion / contamination due to a relaxation of torqued

connectors whicn compromises the seal.

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The critical assembly / handling requirements for connector / cables re-

quires a clean room type facility when servicing or calibrating the

monitoring system. This is not found inside containment.

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The MI cables are subject to rupture from mishandling, surface

scratches and breaking of the sheath due to excessive bending.

The licensee has experienced considerable difficulties in installing a

functional Kaman High Range Containment Monitor due to problems in finding

an acceptable length of MI cable assembly. As a result, the licensee has

received an NRC exception to use an acceptable shortened cable assembly

with one Kaman monitor installed at a lower containment elevation. Two

other monitors (General Atomic) have been located inside containment at

the proper elevation.

This item remains open pending receipt of conclusive environmental test

data that ensures equipment availability in compliance with the require-

ments of NUREG-0737, Table II F.1-3.

3.0 Startuo Test Program

3.1 Procedure Review

Scope

The following approved power ascension test procedure was reviewed

for technical and administrative adequacy and to verify that test

planning satisfies regulatory guidance and licensee commitments:

3-INT-8000 Appendix 8032, Revision 0, Generator Trip From 100% Power.

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Discussion

The above procedure was examined for: management review and approval;

procedure format; clarity of stated test objectives; prerequisites;

environmental conditions; acceptance criteria; source of acceptance

criteria; references; initial conditions; attainment of test objec-

tives;. test performance documentation and verification; degree of

detail for test instructions; restoration of system to normal after

testing; identification of test personnel; evaluation of test data;

independent verification of critical steps or parameters and quality

control and assurance involvement.

Findings

The review indicated that the procedure was consistent with regula-

l tory requirements, guidance, and with the Licensee's commitments.

No discrepancies or unacceptable conditions were identified. The

inspector had no further questions on this procedure.

3.2 Test Witnessing

Scope

The following power ascension tests were witnessed te verify the 11-

censee's conformance to regulatory and procedural requirements, to

observe the performance of the operating staff, and to ascertain the

adequacy of test program' records, including preliminary evaluation

of test results:

3-INT-8000 Appendix 8023, Revision 0, Reactor Trip and Shutdown From

Outside the Control Room.

3-INT-8000 Appendix 8006, Revision 0, Secondary Plant Performance

Test.

Discussion

The above tests were witne:/,ed to verify that; the operating crew

was using the most current test procedure, minimum crew requirements

were met, all test prere:uisites and initial conditions were met,

required test equipment was properly calibrated, the implementing

procedures were technically adequate to perform the test, crew

actions during the test were correct and timely, a quick analysis ,

and evaluation was made to assure proper plant response to the test, i

all test data were collected for analysis by the appropriate per- l

sonnel, overall test acceptance requirements were met and adherence ,

to TS requirements was met for any LCO's affected by the test. i

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FINDINGS

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Appendix 8023, " Reactor Trip and Shutdown from Outside the Con-

trol Room" - The inspector verified that the test satisfied R.G.

1.68 Appendix A, paragraph 5.d.d and FSAR requirements. Credit

was taken for a test performed during the pre-core hot func-

tional test program which demonstrated that the plant could be

placed from a hot standby to a hot shutdown condition. There-

fore, this portion of the test was not conducted.

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[ -The inspector observed the pretest briefings of both the Test crew,

located at the Auxiliary Shutdown Panel, and the Normal crew located

in the Control Room. The briefing of the Test crew included a step

L by step review of the procedure, identification of the various con-

l trols and predicted plant indications and alerting the crew to actions

to be taken in the event unsafe conditions developed during the test.

The Normal crew was briefed on the purpose of test and the actions

they could and could not perform during the test.

It was verified that communications were established and maintained

during the test between the Auxiliary Shutdown Panel (ASP) and the

Control Room, Control Transfer Panels (orange and purple trains),

the Auxiliary Building and the Engineer Safeguards Features Building. l

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The reactor was manually tripped from 15% power with the CRDM breaker

located in the Auxiliary Building. Emergency Operating Procedure E0P

3503 " Shutdown Outside Control Room," Revision 2 was used to shutdown

the reactor and transfer control to the ASP. It was verified that

the test operators were able to maintain Tave at 557*F for over 1/2

hour from the ASP and then transfer control of the plant back to the

Control Room. No problems were identified with the test, test results l

or performance of the operating crew. A problem was identified with i

QC involvement with this test. See paragraph 4.0 "QA/QC Interface" l

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Appendix 8006, " Secondary Plant Performance Test". This test is

performed at various power levels - 30%, 40%, 50%, 75%, 90% and

100% - in order to obtain base line data and identify any sec-

ondary plant problems. The inspector witnessed this test at the

30% power level. It was verified that the latest procedures

were being used, the test crew was knowledgeable of the program

and their responsibilities and the test data was being reviewed

and evaluated against expected results.

No problems were identified.

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3.3 Startup Test Results Evaluation

Scope

The test data results from the tests listed in Appendix A were review-

ed to verify that adequate testing had been accomplished. The results

were also reviewed to verify if regulatory guidance and licensee com-

mitments were satistied and to ascertain whether uniform criteria were

being applied in the evaluation of completed tests in order to assure

their technical and administrative adequacy.

Discussion

The inspector reviewed the test results and verified the licensee's

evaluation of test results by review of: test changes; test excep-

tions; test deficiencies, "as-run" copy of the test procedure; accep-

tance criteria; performance verification; recording of the conduct of

tests; QA inspection records; restoration of system to normal after

the test; independent verification of critical steps or parameters;

identification of personnel cenducting and evaluating test data; and

verification that the test results had been reviewed and approved by

licensee management.

FINDINGS

Turbine Overspeed Test ( Appendix 8016)

The capability of the mechanical overspeed trip device for the main

turbine generator was tested on February 15, 1986. Three (3) actual

trips occurred at 1962, 1963, 1963 rpm, respectively. This test was

accepted as satisfactory since the measured values were within test

procedure established criterion of s1998 rpm. The Backup Overspeed ,

Trip (BOST) Normal Mode is set approximately 0.5% above the mechanical 1

overspeed trip setting, to provide additional protection against tur-

bine overspeed during plant normal operation. The BOST electronic

circuit was also tested at 105% rated speed prior to performing me-

chanical overspeed trip test. The test result was also satisfactory.

Reactor Trip / Shutdown Outside Control Room (Appendix 8023)

The objectives of this test was to demonstrate the following:

a. Verification that the unit can be safely shutdown from outside

the control room using the Auxiliary Shutdown Panel,

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b. Verification that the unit can be maintained in a hot standby

condition from outside the control room.

c. Verification that the unit can be cooled down to the hot shutdown

condition from outside the control room.

Through test results review and test witnessing, the inspector noted

the licensee has demonstrated the remete shutdown capability and met

the above test objectives with the exception of item c.

The cold shutdown capab'lity (item c) was demonstrated earlier during

preoperational test, 3-INT-3000, Appendix 3014, " Remote Shutdown with

Cooldown" on November 1, 1985.

The remote shutdown capability test was accepted as satisfactory by

the JTG.

Automatic Reactor Control (Appendix 8017)

The performance of the automatic control system in raintaining reactor

coolant average temperature within programmed value was verified per

test procedure 3-INT-8000, Appendix S017 on February 17, 1986. The

inspector independently reviewed test data and noted that automatic

reactor control system responded well during the test. Tavg returned

to programmed value withia specified oscillation amplitude, and no

unstable response was observed. The corresponding system responses

such as pressurizer pressu. e, pressurizer level and steam generator

level responded well with no unusual behavior noticed.

During performance of this test, Step 5.5 required the feedwater pump l

speed control system to be set per 3-INT-3000, Appendix 3010. A test l

engineer misread Appendix 3010 to be Appendix 8010. Since Appendix

8010, Neutron Shield Tank Testing, has nothing to do with this test,

Step 5.5 was deleted during the test. The inspector informed the 11-

censee of this discrepancy. The test engineer re- eviewed the test

result and determined that this test change did not have adverse im-

pact on the test outcome, since Appendix 3010 had already been com-

pleted. The inspector concurred with the licensee's assessment and

had no further questions.

Power Coefficient (Appendix 8020)

The doppler-only power coefficient measurements are to be obtained at

30, 50, 75, 90 and 100 percent power. The measurement at 30% power

level was performed on February 17, 1986.

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Since it is difficult to directly measure the reactivity change due

to fuel temperature change during a power change, an indirect measure-

ment technique was utilized. This technique involved measuring the

l primary side responses such as the change in core average temperature

(ATAVG) and the corresponding change in RCS loop parameters with re-

spect to a small change in turbine load. The doppler-only power co-

efficient was then calculated from this set of data and isothermal

i temperature coefficient information as previously derived from Zero

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Power Physics Test.

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All test results, including five sets of power swing data, met the

test acceptance criteria.

The inspector noted that an assumption of dTm=dff (Where T,is

moderator temperature and T f

is average fuel temperature) was used

in the licensee's measuring methodology for the doppler-only power l

coefficient derivation. The validity of this assumption was discussed '

with licensee reactor engineer and consultation provided by a NRR

Core Physics specialist. This assumption is valid when there is no

power change. From the test results and information provided by the l

fuel nndor (Westinghouse), it appeared that this assumption is also  !

valid under test cordition with minimum power change.

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The inspector had no further questions. i

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hatural Circulation Test (Appendix 7006)

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The test witness and preliminary test result review was documented in

the Inspection Report 86-07. Although the licensee completed this ,

test and demonstrated that plant core heat can be removed satisfac- l'

totily by using natural circulation, the detailed test results had

not been thoroughly analyzed. From the test result review, the in-

spector noted that PORV lifted 3 times during the initial stage of

the natural circulation test. Plant behavior and lessons learned

from this test will be evaluated by the licensee and incorporated in

the licensee operator training program and possible procedure enhance- ,

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The licensee management agreed that this detailed evaluation and  !

incorporation into operator's training will be completed by May 1,  !

1986. This is an unresolved item (50-423/86-09-01).

RCS Flow Coastdown (Appendix 5017)

The purpose of this test was to verify that the RCS responded as de-

signed to a partial and complete loss of forced reactor coolant flow.

With four reactor coolant pumps (PCPs) operating, partial loss of ,

forced reactor coolant flow was established by tripping one RCP. Com- l

plete loss of reactor coolant flow was established by simultaneously

tripping all four RCPs. (

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The inspector verified that low flow alarm times, control rod drop

times,-core flows and low flow alarm values met FSAR Chapter 15 Table

15.3-1 requirements. This was determined by reviewing the strip

chart recordings and the sequence of events computer printouts for

reactor coolant loop flow, RCP alarm, reactor trip times, and low

flow' alarms. It was also verified that the licensee had correctly

translated the information from the strip chart recorders and computer

print-out to the data sheets and, by performing independent calcula-

tions, that the final results were correct.

Post Core Hot Functional Tests UNSATS

A review of the disposition of all UNSATs foentified during the Post

Core Hot Functional Test Program was made. These UNSAT's, identified

in Appendix B, were reviewed for conformance with administrative re-

quirements and proper disposition of problems. No problems were

identified.

4.0 QA/QC INTERFACE

While observing the performance of 3-INT-8000 Appendix 8023, Reactor Trip /

Shutdown Outside the Control Room, the inspector questioned a QC inspector

also witnessing the test as to his involvement with the test. From these

discussions the following information was determined:

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Although the QC inspector stated that he was there to verify operator

compliance to E0P 3503, he had not yet reviewed E0P3503 or Appendix

8023 nor did he have a copy of the procedures with him.

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The QC inspector had not been briefed by his management on what to

look for or expect during the test. This was later verified by his

immediate supervisor.

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The QC inspector had not reviewed the FSAR or RG 1.68 requirements

pertaining to this test although he was aware he should do so.

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The QC inspector was not aware of what was taking place as evidenced

by the fact he did not know from what power level the reactor had

been tripped.

The NUSCO QA and NNECO QC departments share responsibility for performing

surveillances of startup tests. This responsibility is loosely coordin-

ated with NUSCO QA having primary responsibility. During this particular

test, representatives from both organizations were present. The auditor

from the QA section appeared knowledgeable of the test procedure and what

was taking place.

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The above occurrence of an inspector being assigned an inspection task

for which that inspector had not been trained or briefed constitutes a

violation (50-423/86-09-02).

At the exit meeting, the licensee QA representative agreed that:

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More specific guidance will be provided to QC inspectors who will

cover the remaining startup tests, as to which areas / criteria to

look into.

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NUSCO QA will continue to provide surveillances during the startup

test program.

The following NUSCO Operations QA surveillance reports were reviewed to

determine the adequacy of 00A's involvement with the Startup Test Program:

NUSCO 00A SURVEILLANCE NO. TITLE

TC 3950 Digital Rod Position Indication

TC 3960 Boron Endpoint Measurements

TC 3960A Boron Endpoint Measurements

TC 3961 Initial Criticality

TC 3961A Initial Criticality

TC 3968 Natural Circulation

TC 3972 Preparations for Power Ascension

Testing

TC 3976 Power Ascension Test

TC 3986 Integrated Plant Testing

TC 3986A Integrated Plant Testing

TC 39868 Integrated Plant Testing

The surveillance appeared to be thorough with good follow-up of identified

concerns. No problems were identified.

5.0 Independent Calculations

3-INT-5000 Appendix 5010, "RTO Bypass Loop Verification" obtains data

which is used to calculate the hot leg and cold leg flow and transport

times through the RTO Bypass Lines. Using the formulas:

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H,= Ft Fc' = Ft-Fgg

(1+ Fc) ,

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Fc' = Calculated cold leg flow

FH ,= Calculated hot leg flow

Fg= Measured Total flow

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Fc = Measured cold leg flow

FH = Measured hot leg flow

The inspector verified the flows and transport time for the hot and cold

leg RTO bypass lines.

Power coefficient measurement requires lengthy data reduction. The

inspector independently verified that the predicted reactor physics

parameters were correctly taken from the nuclear design reference. The

inspector also performed an independent calculation and confirmed that  ;

the first power swing case data were being correctly reduced.

6.0 Plant Tours

The inspector made several tours of the facility durtag the course of the

inspection. This included tours of the control building and control room.

A review of the work in progress, security, cleanliness and housekeeping

was made.

7.0 Exit Meeting

An exit meeting was held on March 14, 1986 to discuss the inspection

scope and findings, as detailed in this report (see paragraph 1.0 for

attendees).

At no time was written material given to the licensee. The inspector

determined that no proprietary information was utilized during this

inspection.

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APPEN0IX A

TEST DATA REVIEWED

TEST NUMBER TITLE

3-INT-5000 Appendix 5001 Shutdown Margin

3-INT-5000 Appendix 5010 RTD Bypass Loop Verification

3-INT-5000 Appendix 5015 Oigital Rod Position Indication

Operation Test

3-INT-5000 Appendix 5016 Loose Parts Monitoring

3-INT-5000 Appendix 5017 RCS Flow Coastdown

3-INT-5000 Appendix 5031 Chemical and Volume Control System

3-INT-6000 Initial Criticality

3-INT-7000 Appendix 7006 Natural Circulation

3-INT-8000 Appendix 8023 Reactor Trip / Shutdown Outside

Control Room

3-INT-8000 Appendix 8016 Turbine Overspeed Test

3-INT-8000 Appendix 8017 Automatic Reactor Control

3-INT-8000 Appendix 8020 Power Coefficient

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APPENDIX B

POST CORE HOT FUNCTIONAL TEST UNSATS REVIEWED

TEST PROCEDURE UNSAT #'s

3-INT-5000, Anpendix 5001 7497

3-INT-5000, Appendix 5002 7471

3-INT-5000, Appendix 5004 7341, 7342

3-INT-5000, Appendix 5006 7495, 7492, 7493

3-INT-5000, Appendix 5007 7485, 7486, 7489, 7496

3-INT-5000, Appendix 5009 7466

3-INT-5000, Appendix 5015 7487

3-INT-5000, Appendix 5016 7475, 7479

3-INT-5000, Appendix 5017 7504, 7510

3-INT-5000, Appendix 5031 7472, 7473, 7474, 7476

7477, 7478, 7484, 7488

7490, 7491, 7499

3-INT-5000, Appendix 5033 7378, 7417, 7420

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