ML20246H354

Insp Rept 50-271/89-05 on 890404-07.No Violations or Deviations Noted.Major Areas Inspected:Implementation of ATWS Rule 10CFR50.62,open Items Identified in Previously Issued Insp Repts & Surveillance & Calibration Program

ML20246H354
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	05/03/1989
From:	Blumberg N, Finkel A, Oliveira W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20246H326	List: ML20246H321 ML20246H354
References
50-271-89-05, 50-271-89-5, NUDOCS 8905160073
Download: ML20246H354 (10)

See also: IR 05000271/1989005

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-271/89-05

Docket No. 50-271

- License No. DPR-28

Licensee: Vermont Yankee Nuclear Power Corporation

RD 5 Box 169, Ferry Road

Brattleboro, Vermont 05301

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Facilit'y Name: Vermont Yankee Nuclear power Station

-Inspection At: Vernon, Vermont-

Inspection Conducted: April 4-7, 1989

Inspectors: UW

A. Finkel, S'enior Reactor Engineer date'

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W. 0 iveira, Reactor Engineer d' ate '

- Approved by: d Cm [ h

7 'J

N. u'mberg, Ch'ief, Operational Programs

5ection, 08,'DRS / fate /

Inspection Summary: Inspection on April 4-7, 1989 (Inspection Report No.

50-271/89-05)

Areas. Inspected: Routine unannounced inspection consisting of the review and

implementation of the ATWS Rule 10 CFR 50.62; open items identified in pre-

viously issued NRC inspection reports; surveillance and calibration program;

and quality assurance / quality control interface with surveillance testing.

Results: Licensee is complying with the ATWS Rule 10 CFR 50.62. The quality

organization appears adequately staffed to support the present operation of

this site. ' Surveillance testing and the calibration program are following the

licensee established program plan. During this inspection unresolved items

85-26-01, 87-08-01 and 87-18-02 were closed. No violations or deviations were

identified.

8905160073 890504

PDR

0 ADOCK 05000271

PDC

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Details

1.0 Persons Contacted

Vermont Yankee Nuclear Power Corporation

.P. Donnelly, Maintenance Superintendent

• R. Grippardi, Quality Assurance Supervisor

• R. Pagondin, Technical Services Superintendent

• J. Pelletier, Plant Manager

D. Reid, Operations Support Manager

S. Skibniowsky, Chemistry Supervisor

R. Spinney, Training Manager

W. Wiltmer, Construction Superintendent

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United States Nuclear Regulatory Commission

G. Grant, Senior Resident Inspector

The inspectors also held discussions with other licensee personnel during

the course of this inspection.

• Denotes those present at the exit meeting on April 7, 1989.

2.0 Follow-up of Previous Inspection Findings (92706)

(Closed) Unresolved Item'85-26-01 - Establish qualification and

certification requirements of receiving inspector (QAT) in accordance

with ANSI N45.2.6-78. The job description for the quality assurance

technician (QAT) was revised November 3, 1985 and approved on

January 16, 1986. The revised QAT job description reflects the intent of

ANSI N45.2.6-1978, paragraph 3.5.2, Level II in the area of schooling and

experience. The inspector reviewed the present QAT personnel records and

verified that their qualifications comply with the licensee's revised job

description requirements. This item is closed.

(Closed) Unresolved Item (87-08-01) - QDR Manual Requirements were not

followed during maintenance of the RHR motors. The licensee's equipment

manual recommended that the motor oil be changed at six month intervals.

The maintenance program for the RHR motors required the oil to be

replaced during each refueling cycle. During the third quarter of 1987, ,

General Electric (GE) completed an evaluation of the electric motor a

maintenance program at Vermont Yankee. GE's recommendation to the

licensee on the subject of motor oil lubrication stated that oil

lubricated motors should be sampled every six months and the oil changed

every outage.

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The Vermont Yankee maintenance procedures require that oil be changed at i

refueling outages based on plant history records and motor test results.

The Maintenance Department does not plan to. perform cil sampling as

recommended by GE unless their test data indicates a motor problem. Based

on' their historical records and test hta results for their RHR motors,

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the inspector concurred that their prm*nt oil replacement program is 1

acceptable. This item is closed. j

(Closed) Unresolved Item (87-18-02) - Maintenance Department failed to

classify activities into major-minor inspection criteria as defined in

procedure AP-6025. The licensee specifies the inspection criteria in .

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their individual department procedures and uses the peer inspection .

process to perform first level inspection of the task. Procedure AP-6025 {

was revised on April 25,.1988 to delete the requirement for using the

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major-minor method of classifying inspection criteria. The surveillance

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and audit inspections were performed by the quality organization as

defined in their quality control program for this site.

To ensure that the'Ap-6025 revision removing the requirement for major-minor

classification did not present a safety concern, the inspector reviewed

the QA Surveillance Reports88-130 on the Instrumentation and Control Depart-

ment-and 88-113 on the Mairitenance Department, which evaluated the effective-

ness of the revised AP-6025. These surveillance reports and the inspector's

evaluation of sampled surveillance performed during 1988-1989 indicated <

that the departments were complying with the criteria defined in Ap-6025

without an apparent lessening of work quality or safety. This item is

closed.

3.0 Anticipated Transient Without Scram (ATWS) Rule,10 CFR 50.62 (.2500/20)

3.1 Scope:

The scope of the inspection was to determine that the Vermont Yankee's

(VY) present system design complies with the requirements of 10 CFR

50.62, paragraphs C (3), (4), (5), (d), and that their test programs

l are designed to ensure that the Alternate Rod Injection (ARI), Standby

Liquid Control System (SLCS), and the recirculation pumps will function

as designed under conditions indicative of an ATWS. The documentation

that was reviewed by the inspectors and used in the validation of the

design and test data reviews are listed in Attachment A of this report.

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3.2 Alternate Rod Injection (ARI) System and Recirculation Pumps

3.2.1 Description of Modification:

To comply with the independence criteria of 10 CFR 50.62,

paragraph C, the ARI system is designed so that upon a high

pressure or low water level sigual the system relays, which

are in a de-energized state, will energize and trip the

reactor. The Reactor Prctection System (RPS) upon a high

pressure or low water level signal will de-energize causing a

reactor trip.

The VY ARI has a 10 second time delay added to provide for longer

core flow during level decrease transients, to assure adequate

core cooling, before the recirculation pumps are shut down. The

high reactor vessel pressure trip signals opens the GE AKF-2-25

motor generator (MG) field breakers which supply the voltage for

the recirculation pumps thus causing the pumps to stop and coast

down. Both the RPS and the ARI systems remove the air supply

from their scram valves, causing these solenoids valves to open

thus providing a reactor scram signal. The RPS and the ARI systems

are redundant from the sensors to the activating devices, however,

the level transmitters and the logic cards are the same manufacture

(Rosemount 1152T0280's), but completely redundant.

3.2.2 Finding:

To ensure that the ARI installed system complied with the

system design criteria and that the test program demonstrated

the operation of the system, the inspector verified tne

installed equipment of one train of the ARI system from the

sensor to the tripping relay devices. The inspector also

reviewed the system surveillance test data from 1988 through the

first quarter of 1989; verified that the ARI train was installed

per the design drawings; and verified that the test data

determined that the ARI system was functional. This item is

closed.

3.3 Standby Licuid Control System (SLCS)

3.3.1 Description of Modification:

The ATWS rule, 10 CFR 50.62C, (4), requires all BWR's to modify 4

their SLCS to be equivalent in control capacity to an 86 gpm

injection rate of 13% sodium pentaborate solution. The 86 grita

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and 13% values stated in the ATWS rule were based on a reactor

vessel size of 251-inch diameter. Vermont Yankee's reactor

vissel size is a 205-inch diameter vessel and based on this -

' their safety analysis establish different rates. VY is also

using a higher inventory of Boron-10. The use of Boron-10

isotope provides the neutron absorption (negative reactivity)

required to shutdown the reactor if the RPS and ARI- systems do

not. Yankee Atomic Electric Company calculation (VY C-572)

determined the minimum Boron-10 aton' percent to be 43, based on

a 41 gpm injection rate, to meet the ATWS 10 CFR 50.62C, '4)

rule for SLC equivalency.

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3.3.2 Fitidings:

The inspector reviewed the pump test and chemistry reports for

1988 through the first quarter of 1989 to ensure that the system

design and the chemistry concentration of the SLCS complied with

the requirements of calculation VY C-572. The pump tests _and

the corrected chemistry data results were within the VY C-572

calculation requirements for this system design.

During this inspection, the refueling SLCS serve 111ance test per

Procedure No. OP4114 was performed. The data had not yet been

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approved through the licensee's data review system during this

inspection period; however, test data reviewed by the. inspector-

appeared to be within the requirements of the test procedure.

The test data supported the VY C-5'/2 calculations for the pump

injection rate and the chemical concentration of the SLCS tank.

In addition, the refueling surveillance test, OP 4114, also met

the VY C-572 flow criteria. The above data reviewed by the

inspector complied with the VY Technical Specification (TS)

section 3.4 Limiting Condition for Operation and 4.4

Surveillance Requirements. .This item is closed.

4.0 Complex Surveillance (61701) and Testing / Calibration (61725)

4.1 Scope:

The inspector observed selected tests to determine if the tests were:

(1) performed in accordance with approved Instrument and Control

(I&C) and Maintenance surveillance / test procedures; (2) properly

administrative 1y controlled, and (3) were performed by trained and

qualified personnel.

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4.2 Findirns:

4.2.1 Surveillance / Test Program:

The surveillance / test program was well documented and

administrative 1y controlled by the Engineering Support

Department (ESD), the Maintenance Department, and the I&C

Department. The key procedures for administering the

surveillance program are Administrative Procedures (AP) 0200

(Maintenance), AP 0310 (I&C) and AP 4000 (ESD). Departmental

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procedures (DPs) provided the details. for implementing the

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surveillance program. These procedures and other documents

referenced in the inspection report are listed in Attachment A.

4.2.2 Surveillance Program Activities:

The performance of two I&C surveillance tests and one

Maintenance test listed below were observed to assess the

surveillance and test / calibration effort for: administrative

controls) test procedure technical content; planning and timely.

performance of the tests; QA coverage; corrective actions;

reporting and analyzing the results (data); and post test

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restoration.

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Vibration tests by the Maintenance Department of the core

spray pump in accordance with OP 4123.

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Reactor core isolation cooling (RCIC) system A/B isolation

logic test in accordance with OP 4368.

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Automatic initiation test of the Primary Containment Isolation

System (PCIS) valves in accordance with OP 4334.

The above tests were satisfactorily performed. However, while

observing performance of OP 4334 the inspector noted that step

57 required the actuation of PCIS valves A-20, A-22A and A-228

in the Control Room. The arrangement of Control Room Panel 9-3

was, however, changed by EDCR No.86-411 and these valves were

redesignated NI-20, NI-22A and NI-22B, respectively. AP 6004

and YNSD WE-100 requires procedures and drawings affected by an

EDCR to be corrected. A Plant Operations Review Committee

(PORC) approved EDCR 86-411 without verifying that the "no

procedural changes required," statement of the data package was

correct. Documents AP 0155 and control room drawing 191175,

which designate valve numbers, were not changed per the approved

EDCR 86-411. The licensee acknowledged the error and initiated

actions to correct the error and determine the root cause of why

the EDCR process did not identify these documents. The

inspector had no further questions.

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4.2.3 QA/QC Interfaces with Surveillance Testing:

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Five annual QA audit reports and six QA surveillance reports

listed in attachment A sections 2.3 and 2.4 were reviewed j

by the inspector. The reports were comprehensive and the i

corrective action resp 0nses were complete and timely.

A Quality Service Department (, tor was observed

performing surveillance inspection of the PCIS valves

automatic initiation test of OP'4334. The QSD inspector

documented the NRC inspectors concern regarding the PCIS

valves discussed in paragraph 4.2.2 and placed in the

licensee's corrective action system.  ;

4.3 Conclusion: j

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The surveillance / test and calibration effort is documented and  !

1mplemented satisfactorily by qualified personnel. The licensee

has initiated action to determine the root cause regarding the

proper implementation of the EDCR process discussed in paragraphs i

4.2.2 and 4.2.3.  !

5.0 Exit Interview ,

Licensee management was informed of the scope and purpose of the

inspection at the entrance interview on April 4,1989. At the conclusion I

of the site inspection on April 7, 1989, an exit itterview was conducted

with the licensee's senior site representatives (denoted in section 1).

The fir. dings were identified and the status of previous inspection

findings were discussed.

At no tiCE during this inspection Was written material provided to the

licensee by the inspector. Based on the NRC Region I review of this

report and discussions held with licensee representatives during this

inspection, it was determined that this report does not contain

ir. formation subject to 10~CFR 2.790 restrictions.

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Attachtnent A

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1.0 Anticipated Transient Without Scram (ATWS) 1

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1.1 Alternate Rod Injection (ARI) Documentation l

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RPS Instrumentation and Control Drawings 850-851 and 855-856

• AR1 Instrumentation and Control Drawings 860-867

• RPS Level Drawing - Rosemont 1152T0280

• ARI Level Drawing - Rosemont 1152T0280

• RPS Logic Trip Drawings 803-828

• ARI Logic Trip Drawings 862

• RPS/ARI Scram Valve Drawing G;91170

• Operations Procedure (0P) 4337 - Reactor Water Level ECCS

Initiation, Isol, Funct/Cali-

bration, Revision 21,

March 31,1989

• Operations Procedure (OP) 4342 - Rx Pressure RPT/ARI/Eccs

Func/ Calibration, Revision 8,

January 25, 1989

• Operations Procedure (OP) 4369 - RPT/ARI Operational Test,

Revision 3, January 5, 1989

1.2 Standby Liquid Control System (SLCS)

Calculation VYC-572 - Yankee Atomic Electric Company SLCS Design

Calculation

• Operations Procedure (OP) 4611 - Sampling and Treatment of the

Standby Liquid Control System

• Operations Procedure (OP) 4611.01 - Standby Liquid Control

Sy5. tem Treatment Request

• Operations Procedure (OP) 4611.02 - Analysis of Enriched Boric

Acid, Enriched Sodium

Pentaborate Decahydrate,

Borax and Sodium Pentaborate

Solution for % B-10

Composition

• Operations Procedure (0P) 0630 - Water Chemistry

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Attachment A 2

• Operations Procedure (0P) 2114 - Operation of the Standby

Liquid Control System

Operations Procedure (OP) 4114 - Standby Liquid Control System

Surveillance

Department' Procedure (DP) 0641 - Procedure for Logging Results

of Chemical Analysis i

Drawing 191171 - Standby Liquid Cantrol Flow Diagram

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SLCS Surveillance Test Reports 1981 - First Quarter 1989 -

1.3 Reactor Recirculating Pumps .

B-19130 (SH702) - Control Wiring Diagram - Reactor Recirculating  !

Motor Generator Set MGI-1A Field Breaker

• B-191301 (SH803) - Reactor Protection System - Reacter Scram .

Sensors - Channel Al  ;

2.0 Complex Surveillance and Testing / Calibration.

2.1 Administrative Procedures

• Administrative Procedure (AP) 0155 - Current System Valve and

Breaker Lineup and Identification, Revision 11, March 2, 1988

• AP 0200 - Maintenance Program, Revision 11, July 1, 1987

• AP 0310, Surveillance Preventative and Corrective Maintenance

Program, Revisian 3, August 12, 1988

AP 4000 - Surveillance Testing Control, Revision 10,

July 20,1988

AP 6004 - Engineering Design Change Requests, Revision 12,

November 3, 1988

• Yankee Nuclear Services Division Engineering Instruction

WE-190, Engineering Design Change Request, Revision 17,

February 15, 1989

e Quality Assurance Department Procedure 0QA-X-1, Quality

Assurance Surve111ances, Revision 12, December 29, 1987

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Attachment A- 3 )

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Instrument and Control (I&C) Department Procedure (DP) 0301,

Calibration and Control of Measuring ar.d Test Equipment (M/TE),

Revision 11, September 23, 1987

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2.2 Test Procedures

• Operations Procedure (0P) 4123 - Core Spray System

Surveillance, Revision 18,

February 27, 1989

OP 4334 - Automatic Initiation Test of PCIS Valves, Revision j

11, March 23, 1989

• OP 4368 - Reactor Core Isolation Cooling (RCIC) System A/B

Isolution Logic Test, Revision 18, February 17, 1989

2. 3 Annual QA Audit Reports I

• VY-86-11 - Measuring and Test Equipment, September 19, 1986

• VY-87-11 - M a suring and Test Equipment, June 11, 1987 {

• VY-87-15 - Technical Specifications, July 29, 1987

• VY-88-15 - Technical Specifications, March 11, 1988

• VY-88-11 - Measuring and Test Equipment, September 16, 1988

2.4 QA Surveillance (SURVs)

• SURV 88-0S - Core Spray System Surveillance, January 8, 1987

• SURV 88-36 - Surveillance of Scram Discharge Instrument, Volume

High Water Functional Test, March 4, 1988

• SURV 89-21 - Shutdown to Low Power Standy, February 11, 1989

= SURV 89-22 - Reactor Core Isolation System Injection to Reactor ,

Vessel, February 13, 1989  !

• SURV 89-24 - Battery Performance Test / Preventive Maintenance, ,

February 22, 1989

• SURV 89-45 - Assembly of Reactor and Drywell System for

Operation, March 23, 1989

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