ML20151C157

Insp Rept 50-271/88-03 on 880209-0321.Violations Noted. Major Areas Inspected:Actions on Previous Findings,Routine Plant Operations,Physical Security,Operational Events,Maint Activities & Licensee Response to NRC Initiatives

ML20151C157
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/05/1988
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20151C143	List: ML20151C140 ML20151C151 ML20151C157
References
50-271-88-03, 50-271-88-3, NUDOCS 8804120190
Download: ML20151C157 (22)

See also: IR 05000271/1988003

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U. S. NUCLEAR REGULATORY COMMISSION

Region I

Report No.

50-271/88-03

Docket No.

50-271

License No. DPR-28

Licensee:

Vermont Yankee Nuclear Power Corporation

RD 5. Box 169

Brattleboro, Vermont

05301

Facility:

Vermont Yankee Nuclear Power Station

Inspection At: Vernon, Vermont

Inspection Conducted:

February 9, 1988 - March 21, 1988

Inspectors:

Geoffrey E. Grant, Senior Resident Inspector

John B. Macdonald, Resident Inspector

James E. Kau her, Project Engineer

Approved By:

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N/M

Donald R.'Haverkamp,

ef "

/ 06te

Reactor Projects Sect; n No. 3C

Inspection Summary:

Inspection on February 9, 1987 - March 21, 1988

(Report No. 50-2'71/88-03)

Areas Inspected:

1.

Routine inspection on daytime and backshif ts by two resident inspectors

(210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br />) of:

actions on previous inspection findings;

routine plant

operations; physical security; licensee potentially reportable occurr-

ences; licensee event reports; operational events; maintenance activities;

cere spray weld overlays; licensee response to NRC initiatives; and

periodic reports.

2.

Safety Issue Management System (SIMS) item number 41 (MPA B-58) on scram

discharge volume capability was closed out through performance of TI

2515/90.

8804120190 880401

PDR

ADOCK 05000271

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DCD

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Inspection Summary (Continued)

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Results:

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1.

General Conclusions on Adequacy, Strength or Weakness in the Licensee's

Program

Based upon the number of long outstanding unresolved issues that were able

to be closed out during this report period, it is apparent that the licen-

see has been successful in improving tracking, resolution and documenta-

tion of these items (Section 3.0).

The licensee approach to and interim resolution of Local Power Range

Monitor (LPRM) spiking problems exhibited thoroughness and sound engineer-

ing judgement (Section 8.1).

The licensee demonstrated strong short-notice maintenance planning and

implementation during the March 18-19, 1988 power reduction (Section

9.1).

2.

Violations

One violation was identified concerning failure to meet the limiting con-

dition for operation in Technical Specifications 3.9.A.1 and Table 3.9.1

concerning requirements for obtaining 24-hour grab samples of the service

water system when the effluent radiation monitor was inoperable (Section

6.0).

3.

New Unresolved Items Identified

The licensee program for identifying, analyzing and reporting plant

occurrences is weak in some areas.

Adequate investigation to determine

reportability of plant events is sometimes lacking. Occasional lapses in

application of reporting requirements by plant personnel have led to non-

conservative, incorrect determinatior.s.

This area is unresolved pending

licensee improvements.

Related to the review of this area was an unre-

solved item addressing the reportability of operations with a service

water (SW) effluent radiation monitor having insufficient sensitivity in

some operational modes (Sections 6.0 through 6.3).

The licensee must resolve the conflict between existing TS requirements

and procedures regarding surveillance testing of reactor vessel head spray

isolation valves RHR-32 and RHR-33 (Section 6.4).

The licensee 1987 Annuai Operating Report did not include all of the

information required by 10 CFR 50.59 and is an unresolved issue pending

licensee review and corrective actions (Section 12.0).

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TABLE OF CONTENTS

PAGE

1.

Persons Contacted. . . . . .

1

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2.

S umma ry o f Fa c i l i ty Ac t i v i t i e s . . . . . . . . . . . .

1

3.

Status of Previous Findings (IP 92701)*. . . . . . . .

2

3.1 (Closed) Follow Item 83-21-03:

NRC Review of Clarification of the Licensee

Response to Item 3.0 of IEB 80-08 . . , . . . . .

2

3.2 (Closed) Unresolved Item 83-17-10:

Service Water System Safety Evaluation. .

2

....

3.3 (Closed) Follow Item 84-10-02:

Torus and CST TS Level Setpoint Changes . . . . .

2

3.4 (Closed) Follow Item 84-01-02:

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HPCI and RCIC Steamline Drain Valves. . . . . . .

2

3.5 (Closed) Unresolved Item 84-12-01:

Stack Gas Instrumentation Problems,

. . . . . .

3

3.6 (Closed) Unresolved Item 84-22-02:

Emergency Lighting for Alternate Shutdown . . . .

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3.7 (Closed) Unresolved Item 85-20-08:

Testing of the Health Physics Notification

Network (HPN) at the New Emergency Operating

Fa c i l i ty ( EO F) . . . . . . . . . . . . . . . . . .

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3.8 (Closed) Unresolved Item 85-18-05:

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Invalidated Environmental Qualification (EQ) of

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Conri Assemblies Containing Teflon in the Pottin

Compound. . . . . . . . . . . . . . . . . . . . g

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3.9 (Closed) Follow Item 85-02-01:

Review of Applicability of IEN 84-86. . . . . . .

4

3.10 (Closed) Follow Item 84-26-04:

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Review of New Revision to LP 3126, Shutdown

Using Alternate Methods . . . . . . . . . . . . .

4

3.11 (Closed) Unresolved Item 85-14-07:

Review of Nitrogen Inerting System.

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4.

Operational Safety (IP 71707, 71710, 61726). . . . . .

4

4.1 Plant Operations Review .

4

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4.2 Safety System Review .

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4.3 Feedwater Leak Detection System .

5

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4.4 Inoperable Equipment.

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4.5 Review of Lif ted Leads, Jumpers and Mechanical

Bypasses. . . . . . . . . . . . . . . . . . . . .

6

4.6 Review of Switching and Tagging Operations. . . .

6

4,7 Operational Safety Findings . .

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Table of Contents (Continued)

PAGE

5.

Observations of Physical Security (IP 71707) . . . . .

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6.

Reportable Occurrence Review (IP 50712, 90714) . . . .

7

6.1 Requirements. . . . . . . . . . . . .

7

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6.2 Service Water System Effluent Radiation Monitor

Inoperability - PRO 87-38 . . . . . . . . . . . .

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6.3 Post Maintenance Operability Testing of Valve

V10-39A - PRO 87-60 . . . . . . . . . . . . . . .

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6.4 Missed Surveillance - PRO 88-012. . . . . . . . .

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6.5 Summary of Findings . . . . . . . . . . . . . . .

12

7.

Licensee Event Reporting (LER) (IP 92700). . . . . . .

12

8.

Operational Event Review (IP 71707, 61726) . . . . . .

12

8.1

Local Power Range Monitor Spiking Problems. . . .

12

9.

Review of Maintenance Activities (IP 62703). . . . . .

13

9.1 Preplanned Power Reduction to Accomplish

Scheduled Maintenance . . . . . . . . . . . . . .

13

10.

Core Spray Safe End Nozzle Weld Overlays (IP 92703). .

14

11.

Review of Licensee Response to NRC Initiatives:

(IP 92703, 25590). . . . . . . .

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11.1 TI 2515/90:

Scram Discharge Volume Capability.

15

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11.2 RI TI 88-01:

Fitness for Duty Program.

17

.....

12.

Review of Periodic and Special Reports (IP 90713). . .

17

13. Management Meetings (IP 30703, 40700), . . ,

. . . .

18

• The NRC Inspection Manual inspection procedure (IP) or temporary instruction

(TI) or the Region I temporary instruction (RI TI) that was used as

inspection guidance is listed for each applicable report section.

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DETAILS

1.

Persons Contacted

Interviews and discussions were conducted with members of the licensee

staff and management during the report period to obtain information per-

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tinent to the areas inspected.

Inspection findings were discussed

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periodically with the management a-d supervisory personnel listed below.

Mr. P. Donnelly, Maintenance Superintendent

• Mr. R. Grippardi, Quality Assurance Supervisor

• Mr. S. Jefferson, Assistant to Plant Superintendent

Mr. G. Johnson, Operations Supervisor

Mr. R. Lopriore, Maintenance Supervisor

Mr. R. Pagodin, Technical Services Superintendent

• Mr. J. Pelletier, Plant Manager

• Mr. R. Wanczyk, Operations Superintendent

Mr. T. Watson, I & C Supervisor

• Attendee at post-inspection exit meeting conducted on March 29, 1988.

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2.

Summary of Facility Activities

Vermont Yankee Nuclear Power Station (VYNPS) continued full power opera-

tions during this period except for pre planned power reductions to accom-

plish required surveillances and a rod pattern exchange.

An unplanned

power reduction to 75% of rated power was conducted on March 11, 1988 in

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order to isolate a feedwater heater string to effect repairs on a leaking

flange.

The repairs were partially effective and the plant was returned

to full power. Further repairs may be necessary if the leak worsens. The

Institute for Nuclear Power Operations commenced a two-week evaluation of

VYNPC activities on March 14, 1988. A planned downpower to 55% of rated

power on March 19, 1988 included a drywell entry to inspect for the cause

of slightly elevated drywell leakage.

Other maintenance activities were

also performed.

An NRC Region I team completed a review of VYNPC compliance with the

requirements of 10 CFR Part 50, Appendix R during the period of

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February 8-11, 1988 (Inspection Report 88-04).

During the periods March 1-4 and 8-11, 1988, the Resident Inspector

participated in NRC technical training.

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3.

Status of Previous Inspection Findings

3.1 (Closed) Follow Item 83-21-03:

NRC Review of Clarification of the

Licensee Response to Item 3.0 of IEB 80-08.

Item 3.0 of IEB 80-08

stated that for plants which were committed during construction to

perform volumetric examination of certain penetrations on which

radiographic examination was not performed, to justify not performing

radiography or submit plans and schedules to perform radiography.

The licensee has responded that the containment penetrations were

fabricated and installed in accordance with the 1968 Edition of the

ASME Code,Section III. This code edition did not require volumetric

examination to be performed when radiography was not performed and

therefore the requirements of Item 3.0 of IEB 80-08 an not appli-

cable to VYNPS. The inspectors had no further questions.

This item

is closed.

3.2 (Closed)

Unresolved

Item

83-17-10:

Service Water System Safety

Evaluation.

The inspector reviewed procedure OP 2181,

"Service

Water / Alternate Cooling Operating Procedure", Revision 16, and verif-

ied that the service water system crossconnect valve to the fire

system (SW-8) is controlled closed during normal system configura-

tion. Manipulation of valve SW-8 for the purpose of crossconnection

or isolation of the service water and fire systems is directed by

Section K of OP 2181. Normal system configuration with SW-8 closed

is verified in Appendix A.

This item is closed.

3.3 (Closed)

Follow

Item

84-10-02:

Torus and CST TS Level Setpoint

Changes. On October 9,1985, the NRC issued Amendment No. 90 to the

operating license for VYNPS.

This amendment permitted, in part, the

licensee to revise the condensate storage tank (CST) low level switch

f rom "2 inches" to

"3*." in TS Table 3.2.1.

The setting change was

required as a result of the replacement of level instrumentation from

a mechanical float assembly to analog instrumentation which changed

the zero reference level.

The setpoint change from 2 inches to 3%

does not affect actual CST water level.

The torus high water level

switch for automatic HPCI suction transfer from the CST to the torus

was deleted from TS Table 3.2.1, as a result of the Browns Ferry Unit

1 Station Blackout concerns of NUREG/CR-2182, Vol. I.

This item is

closed.

3.4 (Closed) Follow Item 84-01-02: HPCI and RCIC Steamline Drain Valves.

The HPCI steamline drain valves (V23-42 and V23-43) and the RCIC

steamline drain valves (V13-34 and V13-35) were incorrectly identi-

fied in FSAR TABLE 7.3.1 as containment isolation valves.

The cur-

rent FSAR revision has been corrected and these valves have been

removed from FSAR TABLE 7.3.1.

This item is closed.

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3.5 (Closed)

Unresolved

Item

84-12-01:

Stack Gas Instrumentation

Problems.

On June 17, 1984 the licensee determined that both stack

gas instrument channels were inoperable.

The licensee reported the

detector problems to the NRC in LER 84-10, dated July 16, 1984. Stack

gas monitor I had a faulty circuit board in the detector control cir-

cuitry which was repaired.

Stack gas monitor II had a gamma sensi-

tive aluminum window vice a beta sensitive mylar window.

A new

detector was installed and the channel was recalibrated. The inspec-

tors have reviewed the event and have no further questions.

This

item is closed.

3.6 (Closed) Unresolved Item 84-22-02:

Emergency _ Lighting for Alternate

Shutdown.

During a 1984 walkthrough of OP 3126, "Shutdown Using

Alternate Shutdown Methods", an NRC inspector roted several plant

locations where increased emergency de lighting was necessary.

The

licensee reviewed the finding and approved the lighting upgrades

recommended in plant alteration review (PAR) 84-08.

Specifically,

emergency lighting was enhanced in the MCC 9B electrical pressure

regulator cabinet area, the RCIC corner room, the A diesel generator

rcom, the east switchgear room and the torus area between the RCIC

corner room and the HPCI room.

The inspectors had no further ques-

tions.

This item is closed.

3.7 (Closed)

Unresolved Item 85-20-08:

Testing of the Health Physics

Notification Network (HPN) at the New Emergency _ Operations Facility

(E0F). The inspectors reviewed procedure GP 3506, "Emergency Equip-

_

ment Readiness Check," revision 20, to ensure it directs that the HPN

network is tested for operabilty monthly at the E0F and recovery

center.

This item is closed.

3.8 (Closed)

Unresolved

Item

85-18-05:

Invalidated Environmental

Qualification (EQ) of Conax Assemblies Containing Teflon in the

_On May 16, 1985 the licensee was notified by Conax

. Potting Compound.

Buffalo Corporation that some electrical penetrations supplied to

VYNPS by Conax contained teflon insulation and sealant material.

The

presence of teflon, which is highly susceptible to degradation from

radiation exposure, invalidates penetration environmental qualifica-

tion.

A licensee investigation determined that only four penetra-

tions on the outboard bulkhead of the containment personnel airlock

were affected.

The penetrations did not contain any safety class

electronic circuitry.

Failure of these outboard penetrations would

not have resulted in any additional release of radioactivity because

the containment boundary would be established by the inner personnel

airlock.

The licensee reported this material defect to the NRC, in

accordance with the requirments of 10 CFR 21, via correspondence FVY

85-57 dated June 19, 1985. The affected penetrations were replaced

during the first refueling outage following discovery of the issue.

The replacement was performed under MR 85-1363 and was completed on

January 27, 1986.

The inspectors had no further questions.

This

item is closed.

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3.9 (Closed) Follow Item 85-02-01:

Review of Applicabilty of IE Notice

(IEN) 84-86. This notice addressed inadequate isolation between pro-

tection system and non-safety system signals detected during computer

testing.

The licensee evaluation of this Notice concluded that

potential changes in instrument current loops due to computer sam-

pling of mercury-wetted relay failures does not apply at VYNPS. Mon-

itoring of the RPS is accomplished by indirect measurements provided

by auxiliary outputs from Rosemount analog trip units.

This signal

is an electrical duplication of the actual loop current output. The

inspector had no further questions.

This item is closed.

3.10(Closed) Follow Item 84-26-04:

Review of New Revision to OP 3126,

"Shutdown Using Alternate Methods".

The inspectors reviewed revision

3 to OP 3126 issued February 5,1983.

The inspectors observed that:

precautions are highlighted and detailed in that expected plant

respcnses to manual actions are described; immediate actions prior to

abandonment of the control room included manually scramming the

reactor, opening HPCI-24, closing at least one MSIV per steam line

and placing the ADS inhibit switch to bypass if possible; and, en-

hanced detailed riirections have been provided to accomplish a reactor

level trip using the Rosemont trip units.

The licensae has elected

not to direct the shif t supervisor to a specific location in the

event a remote plant shutdown must be accomolished.

The licensee

believes that, due to the varied control room inaccessability scen-

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arios that can be postulated, directing the shif t supervisor to a

specific location by procedure could potentially limit his effective-

ness.

Therefore, the shift supervisor has been provided the latitude

to assess plant conditions and then proceed to the location in which

his effectiveness would be maximized.

The inspectors had no further

question'.

This item is closed,

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3.11 (Closed)

Unresolved

Item

85-14-07:

Review of Nitrogen Inerting

System.

The inspector reviewed the licensee responses to IEB 84-01

and General Electric SIL No. 402 which address potential torus vent

header cracking due to direct impingement of cold nitrogen (less than

40 F). The VYNPS inerting system is designed such that redundant

cold temperature alarms and cutoff valves isolate the system preven-

ting the injection of cold nitrogen into the torus and drywell.

Fur-

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ther, the nitrogen supply lines are 9'6" off the torus centerline,

therefore, the ringheader and downcomers do not align vertically with

the nitrogen injection port.

The inspectors had no further ques-

tions.

This item is closed.

4.

Operational Safety

4.1 plant Operations Review

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The inspector observed plant operations during regular and backshift

tours of the following areas:

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Control Room

Cable Vault

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Reactor Building

Fence Line (Protected Area)

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Diesel Generator Rooms

Intake Structure

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Vital Switchgear Room

Turbine Building

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Control Room iastruments were observed' for correlatior. between

channels, proper functioning, and conformance with Technical Speci-

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fications.

Alarm conditions in effect and alarms received in - the

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control room were reviewed and discussed with the operators.

Oper-

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ator awareness and response to these conditions were-reviewed. Oper-

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ators were found cognizant of board and plant conditions.

Control

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room and shift manning were compared with Technical Specification

requirements.

Posting and control of radiation, contaminated and

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high radiation areas were inspected.

Use of and compliance with

Radiation Work Permits and use of required . personnel monitoring

devices were checked.

Plant housekeeping controls were observed

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including control of flammable and other hazardous. materials.

During

plant tours, logs and records were reviewed to ensure compliance with

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station procedures, to determine if entries were correctly made, and

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to verify correct communication of equipment status.

These records

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included various operating logs, turnover sheets, tagout and jumper

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logs, and Potential Reportable Occurence Reports. Inspections of the

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control room were performed on weeketds and backshifts including

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February 8-11, 16-19 and March. 1-3, 7-10, 14-18, 1988. Operators and

shift supervisors were alert, attentive and responded appropriately

to annunciators and plant conditions.

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4.2 Safety System Review

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The emergency diesel ~ generators, residual heat removal, core spray,

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residual heat removal service water, and high pressure coolant injec-

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unn systems, were reviewed to verify proper alignment and opera-

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tional staws in the standby mode. The review included verification

that (i) accensible major flow path valves were correctly positioned:

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(ii) power si pplies were energized, (iii) lubrication and component

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cooling was proper, and (iv) components were operable based on a

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visual inspection of equipment for leakage and general conditions.

No violations or safety concerns were identified.

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4.3 Feedwater Leak Detection System Status

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The inspector reviewed the feedwater leakage detection system and the

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monthly performance summary provided by the licensee in accordance

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with VYNPC letter FVY 82-105.

The licensee reported that, based on

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the leakage monitoring data reduced as of February 18, 1988, there

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were no deviations in excess of 0.10 from the steady state value of

normalized thermocouple readings, and no failures in the 16 thermo-

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couples installed on the four feedwater nozzles.

Point number 12,

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which previously had shown a downward (cooling) trend (See Inspection

Report No. 50-271/87-21, Section 4.2), did not continue to exhibit

increased cooling during this evaluation period.

Since the initial

cooling indicated at point #12, the temperature has stabilized and

deviations have decreased.

Because the normalized readings have

returned to near pre-outage values, the licenste has determined that

no instrumentation or leakage problems exist.

No inadequacies were

identified and the inspector had no further questions in this area.

4.4 Inoperable Equipment

Actions taken by plant personnel during periods when equipment was

inoperable were reviewed to verify:

technical specification limits

were met; alternate surveillance testing was completed satisfactor-

ily; and, equipment return to seavice upon completion of repairs was

proper.

This review was completed for the following item:

MG foam fire suppression system was inoperable from February 16

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- March 1,1988.

Deluge Valve DV-300 developed seat leakage

requiring the system to be isolated.

A replacement vahe was

not immediately available and was placed on back order. During

the period of system inoperability, compensatory firewatches

were established in accordance with TS and approved procedures.

Maintenance was performed under '4R 38-0343.

4.5 Review of Lif ted Leads, Jumpers and Mechanical Bypasses

Lif ted lead and Jumper (LL/J) requests and Mechanical Bypasses (MB)

were reviewed to verify that controls established by AP 0020 were

met, no conflicts with the technical specifications were created, the

requests were properly approved prior to installation, and a safety

evaluation in accordance with 10 CFR 50.59 was prepared if required.

Implementation of the requests was reviewed on a sampling basis.

The LL/J 88-0005 authorized February 4,1938 was issued to facilitate

installation of leads within lighting panels to improve the flexibil-

ity and reliability of receptacles in the Chemistry Laboratory in the

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event of a loss of offsite power as well as to provide a regulated

voltage circuit for the laboratory computer system.

Work was per-

formed under work request WR 87-0714.

4.6 Review of Switching & Tagging Operations

The switching and tagging log was reviewed and tagging activities

were inspected to verify plant equipment was controlled in accordance

with the requirements of AP 0140, Vermont Local Control Switching

Rules.

The following switching and tagging orders were reviewed:

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88-0221

issued and restored on March 19, 1988 to reposition

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nitrogen spa:tacle flanges in support of the deinerting of the

drywell.

88-0222 -- issued and restored on March 19, 1988 to preclude inadver-

tent nitregen injections into

the drywell

during

the drywell

inspection,

88-0223 -- issued and restored on March 19, 1988 to support mainten-

ance of the MSIV 86A test solenoid.

Maintenance was performed under

MR 88-0441.

Refer to Section 9.1 for more detail.

4.7 Operational Safety Findings

Licensee administrative control of off-normal system configurations

by the use of LL/J, mechanical bypass, and switching and tagging pro-

cedures, as reviewed in Sections 4.5 and 4.6, was in compliance with

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procedural instructions and was consistent with plant safety.

5.

Observations of Physical Security

Selected aspects of plant physical security were reviewed during regular

and backshif t hours to verify that controls were in accordance with the

security plan and approved procedures. This review included the following

security measures:

guard staffing; vital and protected area barrier

integrity; maintenance of isolation zones; and, implementation of access

controls, including authorization, badging, escorting, and searches.

No

inadequacies were identified.

6.

Reportable Occurrence Review

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6.1 Requirements

Licensee Administrative Procedure (AP) 0010 Revision 18, "Occurrence

Reports", details the necessary steps for determining the reporta-

bility of various classes of plant occurrences and includes proce-

dures for processing of the Potential Reportable Occurrence (PRO)

form. The bases for the spectrum of occurrence reporting require-

ments includes plant technical specifications,10 CFR Parts 20, 21,

50 and 73 (and others), AP 0012, "Notifications and Reports Due", AP

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0028,

"Operating

Experience

Review

and

Assessment / Commitment

Tracking", AP 0156, "Notification of Significant Events", and NUREG 1022.

The aggregate of these requirements and guidelines forms the

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basis for licensee determinations of occurrence r? portability and the

method of reporting.

Licensee procedures AP 0010, AP 0012, AP 0028

and AP 0156 generally contain sufficient detail to aid staff person-

nel in the determination of occurrence reportability, method of

reporting, and internal administrative tracking and processing of

reports.

The inspector reviewed the licensee reporting system, and

screened several PRO forms to determine the effectiveness of occur-

rence review and categorization by the licensee.

Two PRO forms had

deficiencies in event classification and are described below.

6.2 Service Water System Effluent Radiation Monitor Inoperability - PRO

87-38

The subject PRO covered an occurrence where the service water (SW)

system effluent radiation monitor was discovered to have no flow

through it and was therefore inoperable.

Discovery was made on

September 3, 1987 by instrumen+.ation and control (I&C) technicians

who were performing a routine calibration cf the SW effluent radia-

tion monitor flow switch.

Initial investigation found the monitor

system suction lined up to the discharge block (normal lineup) but

the monitor system pump was not running.

In accordance with TS 3.9 A.1, during release through the SW system, the effluent radiation

monitor shall be operable in accordance with TS Table 3.9.1.

Per

Table

3.9.1,

if the instrument is inoperable, then 24-hour grab

samples are required to be obtained and analyzed.

In compliance with

this requirement, the plant chemistry department was notifed and

24-hour grab sampling was commenced.

Further investigation could not

determine the reason for Icck of monitor flow and maintenance request

(MR) 87-2094 was initiated to correct the problem. Additionally, I&C

technicians found the monitor system low flow switch plugged with

silt. The switch was cleaned and the calibration completed success-

fully by establishing monitor system flow via an alternate suction

i

flowpath (SW return line to the cooling tower deep basin).

Upon

restoration of the flow switch to service, the SW effluent radiation

monitor indicated downscale (an indication of no flow through the

monitor). Actuation of the low flow switch causes deenergization of

the radiation monitor resulting in a downscale reading as an indica-

tion that the monitor system is inoperable.

This condition existed

from September 4-17 when the system was returned to operable status.

Required grab samples were taken during this period.

'

Based on the above, PRO 87-38 was initiated on September 3, 1987.

Review of PRO 87-38 by the engineering support department (ES0)

determined that the occurrence was not reportable.

This finding was

based upon the premise that systems are declared inoperable when they

are "found or made inoperable"; that the condition was found on

Sr ptember 3; and, that the TS limiting condition for operation (LCO)

.

.

9

in Table 3.9.1 was met (24-hour grab samples).

Reviews by plant

management concurred with this determination.

The Plant Manager

assigned an AP 0028 Category A action item to the operations depart-

,

ment to determine the cause of loss of monitor system flow.

This

action item was completed on October 13 and was approved by the

Operations Superintendent on November 19. The finding of this review

,

indicated that flow through the monitor system was decreased when the

SW system was placed in an abnormal lineup to support outage related

activities.

This resulted in insufficient monitor system suction

pressure and probable tripping of the monitor system punp due to

thermal overload. Subsequent investigation revealed that this situa-

tion existed for twelve days from August 22, 1987 to September 3, 1987

during which time required 24-hour grab samples were not taken.

!

>

Various recommendations to prevent recurrence were included in the

response to the action item.

The results of this operations depart-

ment review were not forwarded to the engineering support department

for further review of reportability requirements.

Inspector review of PRO 87-38 indicated that the licensee was most

probably unaware that a TS requirement had been missed (24-hour grab

samples per Table 3.9.1) and was reportable under 10 CFR 50.73. Dis-

cussions with the Engineering Support Supervisor on February 10, 1988

precipitated licensee review of PRO 87-38.

This review established

the technical aspects and sequence of events described above and

determined that the occurrence was reportable under 10 CFR 50.73.

i

Licensee event report (LER)

88-01

was

subsequently

issued

on

March 11, 1988, over six months after the occurrence.

1

The enhanced licensee review that was performed between February 10

and March 11, 1983 uncovered a related condition that is potentially

reportable as well.

This involves a lack of sufficient sensitivity

of the SW ef fluent radiation monitor, during periods when the circu-

lating water (CW) system is shutdown, to ensure river release limits

specified in TS 3.8.A.1 are met.

Initial indications are that, dur-

ing periods when the CW system is secured (and possibly when in

closed cycle operation) and the dilution effect of this flow is lost,

the SW effluent radiation monitor does not exhibit sufficient sensi-

tivity to ensure release limits are being met and therefore 24-hour

grab samples are required.

This item remains unresolved pending

further licensee review and determination of reportability (50-271/

88-03-01).

.

Licensee performance during the initial review of PRO 87-38 raised

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several concerns with the inspector.

First was that an abnormal

system lineup of SW caused an unanticipated and unrecognized loss of

the SW effluent radiation monitor that went undetected and uncompen-

sated for twelve days.

Second was the lack of aggressive investiga-

tion to determine reportability of PRO 87-38.

Third was the failure

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of the operations department, upon completion of the action item

review, to recognize the reportabilty of the occurrence.

Fourth was

the programmatic disconnect betveen completion of the AP 0028 action

item and forwarding pertinent results to the engineering support

department.

The first concern represents a violation of TS Table 3.9.1 require-

ments (50-271/88-03-02). Corrective actions contained in LER 88-01

appear to be responsive.

The remaining concerns are unresolved

pending licensee review of the occurrence investigation, reporting

and tracking programs (50-271/88-03-03).

6.3 Post-Maintenance Operability Testing of Valve V10-39A - PRO 87-60

On September 29, 1987 work was performed on valve V10-39A (RHR-39A)

per maintenance request (MR) 87-2550 to repair a packing leak.

Valve

RHR-39A is the loop

"A"

torus spray / test isolation valve and is a

containment isolation valve.

Required post-maintenance operational

testing was to be stroke and time testing per MR 87-2550. Documen-

tation of this testing did not occur until October 12, 1987,

In the

interim, the reactor was taken critical on October 2 and 5,1987.

The PRO 87-60 was initiated on October 13, 1987 to review the report-

abilty of critical operatiens with RHR-39A in an apparently untested

condition.

The licensee determined that, since the valve tested

operable on October 12, and adjustments to packing nuts do not affect

valve seat leakage, the occurrence was not reportable. The inspector

noted during the review of PRO 87-60 that a component undergoing

maintenance must be considered inoperable until the required post-

maintenance operability testing has been satisfactorily completed.

Licensee procedures AP 0021, "Maintenance Requests" and AP 0025,

"Plant Equipment Control" support this tenet, and control the pro-

cessing of safety-related maintenance.

The inspector furthe

noted

'

that per TS 3.7,0.1, RHR-39A is required to be operable during reac-

tor power operation except as specified in TS 3.7.D.2.

This TS would

allow RHR-39A to be inoperable during reactor power cperation if at

least one valve in the associated line was in a mode corresponding to

the isolated condition (closed in the case of RHR-39A). This was the

case as RHR-34A was closed during this period.

However, TS 4.7.2

requires in this condition that the position of RHR-34A be logged

daily.

This

requirement

was

not

met

during

the

period

of

October 2-12, 1987, and apparently constituted a missed TS required

surveillance requiring reporting under 10 CFR 50.73.

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11

Discussions with the Engineering Support Supervisor on February 10,

1988 did not yield a satisfactory conclusion.

The licensee main-

tained this occurrence was not reportable.

Subsequent discussions

with licensee management have resulted in further review of PRO 87-60

to determine reportability.

The results of this review determined

that stroke and time testing of RHR-39A was in fact accomplished on

October 1,1987 but due to communication and/or documentation errors

this information was not available.

Therefore, RHR-39A was actually

operable af ter October 1 and invoking the requirements of TS 4.7.2

was not necessary.

However, based upon the original information

contained in PRO 87-60, the licensee should have initially determined

the item to be reportable.

6.4 Missed Surveillance - PRO 88-012

In December 1981 the licensee implemented plant design change request

(PDCR) 81-12 which terminated and blank flanged the reactor vessel

head spray system (RVHS) inside primary containment.

The RVHS sys-

tem, which is a subsystem of RHR, was deactivated to reduce reactor

vessel penetrations and because the system was identified as being

susceptible to intergranular stress corrossion cracking. The system

was terminated downstream of the two containment isolation valves,

l

RHR-32 and RHR-33.

The blank flange was seismically qualified and

was leak rate tested satisfactorily after installation.

On February 16, 1988 the licensee generated PRO 88-012 concerning

missed TS-required surveillance testing of CIVs RHR-32 and RHR-33.

These valves are required to be tested at least once per operating

cycle to ensure primary containment isolation capability in accord-

ance with TS 4.7 D.1.a.

The surveillance test had not been accom-

plished for these valves since the system was deactivated.

The

licensee deemed this PRO to be not reportable. The isolation valves

RHR-32 and RHR-33 are exempt from 10 CFR 50 Appendix J Type C leakage

rate testing per TS Table 4.7.2.6.

The blank flange installed meets

the containment integrity requirements of TS 4.7.3.

,

The inspectors expressed the concern to the licensee that, although

containment integrity has been maintained and the RVHS system has

been permanently isolated, a TS requirement exists to test the re-

sponse of RHR-32 and RHR-33 to an isolation signal.

Therefore, the

.

licensee should readdress the ef fect of the PDCR on the ability to

'

perform and meaningfulness of impacted TS requirements and request TS

amendments and revise procedures as appropriate. The inspectors will

follow the resolution of this issue as an unresolved item (50-271/

88-03-04).

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6.5 Summary of Findings

Although the licensee appears to have adequate procedures to cover

occurrence reporting, in at least one case a programmatic weakness

prevented significant information from being disseminated to appro-

priate organizational levels.

The examples above indicate an occas-

ional lack of aggressive investigation of events surrounding a PRO.

Also, occasionally the licensee uses faulty logic in determining

reportability that appears to result in incorrect /non-conservative

findings.

Licensee attention in this area is required. Corrective

actions will be reviewed under the unresolved item in Section 6.2.

7.

Licensee Event Reporting (LER)

,

The inspector reviewed Licensee Event Report (LER) 88-01, Plant Service

Water Effluent Stream not Monitored Due to Procedure Deficiency. This LER

was submitted as a result of a failure to meet TS limiting conditions for

operations as described in Section 6.2 of this inspection report.

The inspector determined that with respect to the general aspects of the

event: (1) the report was not submitted in a timely manner due to poor

analysis by the licensee as described in Section 6.2; (2) description of

the event presented was accurate; (3) root cause analysis was performed;

(4) safety implic4tions were considered; and (5) corrective actions imple-

mented or planned were sufficient to preclude near-term recurrence of a

,

similar event. Violations and reporting concerns, and the need to address

t

in greater scope and specificity related long-term operational corrective

actions to preclude recurrence, were identified and are detailed in

Section 6.2.

8.

Operational Event Review

8.1 Local Power Range Monitor Spiking Problems

,

Recently, VYNPS has experienced an increased incidence of high output

signal spiking in Local Power Range Monitors (LPRMs). Spurious spik-

ing in LPRMs is a historical problem that has existed for years in

boiling water reactors (BWRs).

The recent increase at VYNPS is at-

tributed to new model NA300 detectors installed during the 1987 out-

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age. Of the eight new installed detectors (two strings of four de-

tectors each) five have experienced cne or more random high spikes.

The NA300 LPRMs are an improved de'ign that, among other attributes,

are intended to reduce the frequency of random spiking observed in

earlier models (NA200 and NA1CO).

Licenses communications with the

vendor have indicated that e9aluation and analysis of the phenomenon

are in progress and potential remedies are being formulated.

Interim

'

resolution is expected in the next few months.

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Normally, LPRM spiking is considered a nuisance and does not create

any significant operational problem or concern.

However, due to the

design of the VYNPS average power range monitor (APRM) system and

reactor protection system (RPS), a single randomly spiking LPRM of

sufficient magnitude has the potential of causing a reactor scram.

The system of 80 LPRM detectors (20 strings of four detectors each)

is arranged such that some LPRMs are shared by two APRMs (A and 0, or

C and F) in opposite RPS trip channels (A and B).

In such a condi-

tion, a spike of sufficier.t magnitude in one of these shared LPRMs

-

has the potential for tripping both of the associated APRMs satisfy-

ing the RPS logic for a full reactor scram.

Four of the new NA300

LPRM detectors are assigned to shared APRMs, and three of these

detectors have experienced spiking problems.

In recognition of the

potential for an unnecessary plant trip due to this condition, the

licensee placed two of the shared APRMs (C and 0) in Bypass on

February 25, 1933. This allowed configuration would prevent a single

random

spike

from causing a

reactor

scram.

Subsequently, on

'

March 8, 1988, the licensee individually bypassed the four shared

LPRM detectors and returned APRMs C and 0 to normal operation.

The

four remair.ing NA300 detectors (which input to non-shared APRMs) will

remai, in operation and have their performance monitored for the

remainder of the cycle or until interim correction is implemented.

,

,

The licensee approach to this problerr, demonstrated thoroughness and

sound engineering judgement.

Communications with the vendor were

timely and appropriate and adequately documented. Licensee permanent

I

corrective actions are contingent upon final vendor analysis and will

be reviewed by the inspector when available.

No inadequacies were

identified.

9.

Review of Maintenance Act'"ities

9.1 Preplanned Power Reduction to Accomplish Scheduled Maintenance

During the weekend of March 19-20, 1983 the licensee performed a

preplanned power reduction to 55*4 power to facilitate various

maintenance activities.

The containment was deinerted to allow

a drywell entry to identify potential primary leakage paths.

Power was reduced to 55'4 at 7:40

a.m.

on March 19, 1988,

Scheduled maintenance .as completed and the unit returned to

i

100*4

power

after

c:mpleting

PCIOMR

at

10:35

a.m.

on

March 21, 1983.

9

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Significant maintenance accomplished included:

f

Drywell inspection - minor packing leakage was observed

.

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from valve RHR-88.

'

MSIV 86A test solenoid was repaired to correct slow test

,

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response times.

Slow closure and fast closure tests were

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performed satisfactorily.

The IB high pressure heater feed recirculating line flange

--

leak was repaired by use of Fermanite sealing compound.

The 2A high pressure heater drain valve operator was

--

replaced.

4

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The B feedwater regulating valve control air tubing was

--

replaced.

,

The main condenser tubes were chill shocked and cleaned.

--

All maintenance activities scheduled to be performed during the

'

power reduction were completed.

The licensee exhibited strong

preplanning and interdepartmental scheduling in accomplishing

<

these objectives.

The inspectors had no further questions.

10. Core Speay Safe End Nozzle Weld Overlays

By letter dated March 1,

1988 (FVY 88-19), VYNPC requested NRC:NRR to

approve a plan for long-term operations with the current weld overlays on

the core spray nozzles.

This issue was most recently addressed by the

,

inspector in Inspection Report 50-271/87-21 (Detail 3.1) with a review of

the VYNPC post-outage report of weld overlay inspections (letter datd

October 20, 1987 [FVY 87-100]). In that letter, the licensee committed to

notify NRC:NRR of future plans regarding replacement of the safe ends upon

i

completion of an evaluation of long-term operation with weld overlays.

1

This submittal fulfills the commitment and includes a technical report,

"Vermont Yankee Nuclear Power Station, Justification for Long Term Opera-

tion for Vermont Yankee Core Spray Nozzle Weld Overlays". Based on this

report, VYNPC determined that continued operation with weld overlays on

,

l

the core spray nozzles is acceptable beyond Cycle 13 (current cycle)

operation.

The licensee further committed to periodically confirm the

evaluation results and the acceptability of continued operation with the

weld overlays through performance of periodic ultrasonic examination of

j

the overlays per NUREG-0313, Revision 2 and the inspection program

,

developed for the past outage. Concurrence with VYNPC plans for continued

operation with the weld overlays was requested of NRC:NRR with a target

review completion date of June 1,1988 to suppen planning for the 1989

outage. The inspector reviewed the current submittal and will follow the

.

results of the NRC:NRR review,

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11.

Review of Licensee Response to NRC Initiatives

11.1 Scram Discharge Volume Capability (TI 2515/90)

Temporary Instruction (TI) 2515/90 provides guidance for performing

inspection followup of boiling water reactor (BWR) licensee activ-

ities to ensure scram discharge volume (SDV) capability in accordance

with long-term commitments concerning Multiplant Action (MPA) Item

B-58. As required by TI 2515/90, the inspector conducted an inspec-

tion of VYNPC design / procedures regarding SOV capability. Each item

of TI 2515/90 and the results of the inspection are delineated below:

Item 4.01 - Scram Discharge Header Size:

Generic safety evaluation

report (SER) for all BWR licensees, dated December 9,1980 shows the

status of this item as acceptable for Vermont Yankee.

Item 4.02 - Automatic Scram on High SDV Level: Four trip units from

each instrument volume (two from each instrument volume for each

reactor protection system trip system) initiate a scram to shut down

the reactor while sufficient free volume is still present to receive

the scram discharge.

Item 4.03 - Instrument Taps Not on Connected Piping:

The inspector

performed a field inspection of the instrument volume (IV) and found

that each IV has four separate upper and lower taps located on the IV

proper.

The taps are located 90 degrees apart and are adequately

supported.

Item 4.04 - Detection of Water in the IV: The level sensing instru-

ments on the IV use separate taps (total of four instruments per IV)

and are powered by separate power supplies and therefore meet the

redundancy requirement.

The diversity issue was addressed by NRC in

a letter dated September 10, 1985.

The safety evaluation report

attached to the letter concluded that the Vermont Yankee design

satisfied the diversity requirement for the IV.

Item 4.05 - Vent and Drain Valves System Interfaces:

The vents and

drain are dedicated systems.

They are not intertied with any other

system that could have an adverse ef fect on proper functioning.

The

HVAC ducts have fixed lower openings nearby so that, in effect, the

vents are to atmosphere.

Similarly, the drain pipe terminates open

ended inside a floor drain pipe collar and is not interconnected with

anything else.

This information was provided by the licensee in an

August 11, 1980 response to IE Bulletin 80-17, Supplement 1, and was

accepted by NRC in its generic SER dated December 9, 1980.

Item 4.06 - Vent and Drain Valves _.Close on Loss of Air: The vent and

drain valves at Vermont Yankee are designed such that they are held

i

open with air against spring pressure. Consequently, they will close

'

on loss of air pressure,

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Item 4.07

Operator Aidi

Four differential pressure type level

-

transmitters on each of the two IV's provide signals to analog-to-

digital trip units. One of the trip units on each instrument volume

initiates an alarm for operator action prior to the IV high level

scram setpoint being reached.

-

Item 4.08 - Active Failure in Vent and Drain Lines: There are two

air operated-to-open, spring pressure-to-close, valves in each drain

.

line.

This configuration meets the single active failure criteria.

The vent lines are configured with one air operated-to-open, spring

pressure-to-close valve and one check valve in each line. This con-

figuration was evaluated as acceptable in the NRC generic SER dated

December 9, 1980.

Item 4.09 - Periodic Testing of Vent and Orain Valves: The SDV vent

'

and drain valves are operability tested monthV per surveillance test

OP 4111.

During the conduct of the operability test the valve

closure time is recorded, and is required to be less than 30 seconds.

Item 4.10 - Periodic Testing of Level Detection Instrumentation:

A

functional test of the Whigh water level instruments is performe 1

i

every one-to-three months per surveillance test procedure OP 4310 and

TS Tables 4.1.1.

Calibration of the IV level instrurent loops is

performed once each operating cycle. The functional part of the pro-

cedure tests the IV high level alarm, rod block and reactor scram set

points, and includes restoration of the system configuration.

The

functional test is perfomred using the electronic calibration unit.

The loop calibration 15 performed using an external input to the

transmitter from a manometer.

l

Item 4.11 - periodic Testing Operability of the Entire System:

The

_

technical specifications at Vermont Yankee contain no requirernent to

perform an integrated operability test of the entire system. A one-

time special system functionality test was performed in 1981.

No

commitment was made by the lice' >ee to modify the TS to require this

testing, and no requirement was imposed by the NRC.

However, the VY

i

TS's were amended (Amendment 73) to conform to the NRR model TS

l

regarding this issue,

i

The inspector determined that the licensee has met its long-term

,

commitments to up grade SDV capabilty, and no violations or devia-

,

tions were identified. TI 2515/90 is closed.

)

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11.2 Region I, TI 88-01:

Fitness for Duty Program

'

Region I Temporary Instruction 88-01 (RI TI 88-01), "Fitness for Duty

(Drug Testing) Information and Reporting" was performed to collect

additional data for review of the experience associated with drug

testing as part of the licensee fitness for duty program.

Informa-

t

tion was obtained to support evaluation of program effectiveness and

responsiveness to the NRC policy statement on fitness for duty

(Federal Register 51 FR 27921 August 4, 1986).

This information was provided to the Region I Division of Reactor

Projects Fitness for Duty Coordinator.

Additionally, reportability

under 10 CFR 73.71 of drug related events was discussed with the

licensee.

NRC NUREG 1304, which summarizes NRC guidance in this

area, is currently under licensee review for impact on the program

and associated procedures.

The licensee is striving to balance

reporting guidelines with the need for confidentiality.

No inade-

.

quacies were identified during this review of the licensee fitness

i

for duty program.

T

]

12.

Review of Periodic and Special Reports

)

Upon receipt, the inspector reviewed periodic and special reports submit-

,

ted pursuant to Technical

Specifications.

This review verified, as

'

a

applicable: (1) that the reported information was valid and included the

NRC-required data; (2) that test results and supporting information were

consistent with design predictions and performance specification; and (3)

that planned corrective actions were adequate for resolution of the

t

4

problem. The inspector also ascert.ined whether any reported information

!

should be classified as an abnormal occurrence.

The following reports

(

,

were raviewed:

4

l

Monthly Statistical Report for plant operations for the month of

--

3

4

February 1988.

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Effluent and Waste Disposal Semi-annual Report for Third and Fourth

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Quarters,1987 Including Annual Radiological In' pact on Man for 1987

[

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Vermont Yankee 1937 Annual Operating Report

i

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Review of the VYNPS semi-annual effluent release report included the

following areas:

Liquid and gaseous radioactive effluents and solid waste

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Radiological dose comnitments associated with any ef fluent releases

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Meteorological data

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Changes to the Off-Site Dose Calculation Manual

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The report concluded that since there were no routine or accidental liquid

releases from VYNPS, no doses were attributed to individuals in unre-

stricted areas.

Because noble gas release was below the lower limit of

detectability, no air dose was attributed to this potential release path.

Dose assessments due to iodines and particulates in gaseous effluents were

well below 10 CFR Part 50 Appendix ! and TS 3.8.G.1 criteria. No inade-

quacies were identified.

The VYNPS Annual Operating Report is required to be submitted in accord-

ance with 10 CFR Part 50.59 (b)(2), which states that the report should

contain brief descriptions of facility changes, tests and experiments per-

formed without prior NRC approval per 10 CFR 50.59 (a)(1), as well as a

summary of the safety evaluation associated with each.

Contrary to the

above, the 1987 VYNPS Annual Operating Report did not include summary

safety evaluations for facility changes, tests or experiments.

When

informed by the inspector of this deficiency, the licensee committed to

review and revise as appropriate the report format.

This item is unre-

solved pending documentation of appropriate licensee review and action on

this issue (50-271/83-03-05).

13. Management Meetings

At periodic intervals during this inspection, meetings were held with

senior plant management to discuss the findings.

A summary of findings

for the report period was also discussed after the conclusion of the

inspection and prior to report issuance. No proprietary information was

identified as being included in the report.

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