ML20134B425

Notice of Receipt of Petition for Director'S Decision Under 10CFR2.206.Petition Requesting Certain Actions Associated W/Haddam Neck Plant Operated by Nu

ML20134B425
Person / Time
Site:	Millstone, Haddam Neck  File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	01/23/1997
From:	Miraglia F NRC (Affiliation Not Assigned)
To:
Shared Package
ML20134B413	List: ML20134B410 ML20134B425 ML20134B457 ML20134B470 ML20134B473
References
2.206, NUDOCS 9701300089
Download: ML20134B425 (4)

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7590-01-P U.S. NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-245, 50-336, AND 50-423 l NORTHEAST UTILITIES MILLSTONE NUCLEAR POWER STATION, UNITS 1, 2, AND 3 l

DOCKET NO. 50-213 CONNECTICUT YANKEE ATONIC POWER CONPANY HADDAN NECK PLANT RECEIPT OF PETITION FOR DIRECTOR'S DECISION UNDER 10 CFR 2.206 Notice is hereby given that on November 25, 1996, as amended on December 23, 1996, the Citizens Awareness Network (CAN) and the Nuclear Information and Resource Service (NIRS) (Petitioners) submitted a Petition pursuant to 10 CFR 2.206 requesting certain actions associated with the Haddam

! Neck plant, which the Petitioners refer to as Connecticut Yankee, and the i

three Nillstone units operated by Northeast Utilities (NU).

l Petitioners allege that NU has, over the past decade, mismanaged its nuclear facilities in Connecticut and operated them in flagrant disregard of NRC regulations; that NU has failed to fulfill its commitments to the NRC; that NU management had concrete particularized knowledge of serious on-going violations of NRC regulations culminating in material misrepresentations to the NRC; that regulatory oversight by the NRC to assure NU's compliance with NRC regulations has been a blatant and abject failure; that NU is in violation of 10 CFR Part 50, Appendix B; and that these failures have culminated in inconsistent and inaccurate updated Final Safety Analysis Reports at NU's 9701300099 970123

{DR ADOCK 05000213 PDR Enclosure 1

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nuclear facilities in Connecticut, thereby posing a significant safety concern j for either continued operation of the plants or decomissioning.

j The bases for these assertions are NU and NRC inspection findings and NU a

l documents referred to in the Petition, the amendment and a VHS videotape, Exhibit A, which accompanied the Petition. The videotape has been transcribed j and placed in the Commission's Public Document Room and local public document j rooms. Areas identified are surveillance testing, operation outside the design basis, and degraded material condition. Petitioners assert that this

information demonstrates that there are inadequate quality assurance programs i

at NU's nuclear facilities in Connecticut, that NU has made material false 4

statements regarding its Millstone units, and that safe decommissioning of the

! Haddam Neck plant is not possible given the defective nature of the design and i licensing basis for this facility. In ad(ition, in the smendment the Petitioners assert that certain nitrogen ialculations performed by NU for the f

l Haddam Neck facility may not comply with :0 CFR Part 50, Appendix B, and that

! the NRC failed to identify this problem. The videotape records an August 29, 1996, Citizens Regulatory Comission tele"ised interview of a former Millstone l

l Station employee. The interview included the former employee's views relating to NU's poor management in allowing: degradation of the material plant; poor radwaste practices resulting in potential radiation exposure to employees; and I

harassment intimidation and subsequent illegal termination of employees raising safety concerns.

Petitioners request the following actions

immediate suspension or j revocation of NU's licenses to operate its nuclear facilities in Connecticut;

investigation of possible NU material misrepresentations to the NRC; continued i

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j q shutdown of the NU facilities in Connecticut until the Department of Justice completes its investigation and the results are reviewed by the NRC and until i

j the NRC evaluates and approves NU remedial actions; continued listing of the f NU facilities on the NRC " watch list" should NU resume operation; barring any i

predecommissioning or decommissioning activity at any NU nuclear facility in j Connecticut until NU and the NRC take certain identified steps to assure that j such activities can be safely conducted; and initiation by the NRC of an investigation into how it illowed the asserted illegal situation at NU's l

] nuclear facilities in Connecticut to exist and continue for more than a

! decade. In addition, in the amendment, Petitioners request copies of 1'

Connecticut Yankee's nitrogen calculations and an imediate investigation of a

the need for enforcement action for alleged violation of 10 CFR Part 50, Appendix B.

t l The issues in the Petition, as amended, are being treated pursuant to 10 l CFR Section 2.206 of the Commission's regulations and have been referred to

! the Acting Director of the Office of Nuclear Reactor Regulation. As provided

! by 10 CFR Section 2.206, appropriate action with regard to these issues will l 1 \

be taken within a reasonable time. By letter dated January 23, 1997, the l Acting Director denied the request for immediate suspension or revocation of the operating licenses for the NU nuclear facilities in Connecticut.

1 A copy of the Petition, the amendment to the Petition and a ,

transcription of the videotape are available for inspection at the l i Commission's Public Document Room at 2120 L Street, N.W., Washington, DC 20037 j 1

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I and at the local public document rooms at (1) the Learning Resources Center, Three Rivers Community-Technical College, 574 New Lonaon Turnpike, Norwich, CT 06360, and (2) the Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, CT 06385.

Dated at Rockville, Maryland this23rd day of January 1997. l l

FOR THE NUCLEAR REGULATORY COP 911SSION A f f*

Frank J. ragIi,Jr.,ActingDirector Office of Nuclear Reactor Regulation l

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1 October 30, 1996 4 I

EA No.96-440 i

j Mr. Ted C. Feigenbaum i

Executive Vice President - Nuclear  !

I clo Mr. Terry L. Harpster j

] P.O. Box 128 ,

Waterford, Connecticut 06385 h

SUBJECT:

NRC AUGMENTED INSPECTION TEAM REVIEW OF THE UNDETECTED  :

INTRODUCTION OF NITROGEN GAS INTO THE REACTOR VESSEL DURING i i -

- PLANT SHUTDOWN REPOF.T NO. 50 213/96 80

! i Dost Mr. Feigenbaum:

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! On October 2,1996, the NRC completed an Augmented Inspection Team (AIT) at the  !

j. Connecticut Yankee Atomic Power Company. The enclosed report presents the results of i

that inspection. l j The AIT was chartered to review the events surrounding the inadvertent decrease in i reactor vessel water level during plant shutdown conditions. The team also reviewed other l j

decay heat removal system challenges and equipment failures. The team developed a  ;

a sequence of events, determined the safety significance of the events, and assessed the i

! quality of response by the plant staff and management.

i i For approximately four days, control room operators were unaware that nitrogen gas was ,

leaking into the reactor vessel and causing level to decrease. By September 1,1996, l

- reactor vessel level had decreased to approximately 3 feet below the reactor vessel flange. l The decrease in reactor vessellevel was potentially significant because a further decrease in level could have challenged the function of the operating decay heat removal system.

While there were no actual public health and safety consequences of this event and i adequate decay heat removal was maintained, the situation involving an unintended decrease in reactor water level in combination with the unavailability of decay heat removal

! equipment was safety significant.

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The team identified several areas where operations performance was inadequate. Several operations procedures failed to provide adequate details or contained incorrect information.

The absence of acceptable procedures was a contributing cause for both the nitrogen gas intrusion going undetected and for the inadvertent diversion of water from the reactor coolant system (RCS). Several of the events were exacerbated by plant operators failing to i follow plant procedures, conducting activities without procedural guidance, or making inappropriate decisions. A lack of a questioning attitude resulted in the failure to promptly identify the nitrogen gas accumulation in the reactor vessel. The failure by more senior operators to convey expectations to less experienced field operators during pre job briefings resulted in inappropriate equipment manipulation that either directly caused or contributed to these events.

4 Enclosure 2j{ .' }

U.S. NUCLEAR REGULATORY COMMISSION REGION I ,

Docket No.: 50 213

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License No.: DPR-61 Report No.: 50-213/96-80 j t

Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06141-0270 .

Facility: Haddam Neck Station Location: Haddam, Connecticut Dates: September 3,1996 - October 2,1996 Team Leader: J. Trapp, Senior Reactor Analyst Inspectors: E. Conner, Project Engineer R. DePriest, Reactor Engir.eer P. Habighorst, Resident inspector J. Munro, Sr. Reactor Engineer, NRR T. Shediosky, Senior Reactr>r Analyst i

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Approved by: James T. Wiggins, Director Division of Reactor Safety l

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~9M10***DCK POR AD O$000213 PDR Pd' S

e e EXECUTIVE

SUMMARY

Haddam Neck Station NRC Inspection Report No. 50 213/96-08 Event Summarv/ Safety Sionific.gaggi The AIT (team) conducted an independent assessment of the circumstances surrounding the (1) undetected accumulation of nitrogen gas in the reactor vessel head; (2) two inadvertent diversions of reactor coolant system (RCS) inventory; (3) failure of a residual heat removal (RHR) pump; and (4) a leak on a RHR heat exchanger inlet valve body. These events all occurred between August 22 and September 4,1996.

The team concluded that the combination of these events was safety significant. The operation of the RHR decay heat removal system is contingent upon maintaining adequate levelin the reactor vessel. The accumulation of nitrogen gas in the reactor vessel head significantly reduced the water level in the reactor vessel. The decrease in reactor vessel level went undetected by plant operators for nearly 4 days. The team determined that systems used to mitigate a loss of the RHR system were also adversely affected by the nitrogen gas intrusion. However, the team determined that there were no actual public health and safety consequences of this event.

Ooerstions:

The team identified several areas where operations performance was inadequate. Several operations procedures, used during shutdown ::onditions, failed to provide important detail or contained incorrect information. The absence of acceptable procedures was a contributing cause for both the nitrogen gae intrusion into the reactor vessel going undetected and for the inadvertent diversion of water from the RCS. The team noted j several examples where operator knowledge and performance failed to meet acceptable

! standards. Several of the events were exacerbated by plant operators failing to follow plant procedures, conducting activities without procedural guidance, or making i inappropriate decisions. A lack of a questioning attitude, in response to several indirect

. indications that the reactor vessel was not full of water, resulted in the nitrogen gas

{ accumulation going undetected. The failure by more senior operators to convey i expectations to less experienced field operators during pre-job briefings resulted in l inappropriate equipment manipulation that either directly caused or contributed to these j events. Due to the operators failure to fully appreciate the significance of these events, timely notifications to the NRC were not made.

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l Maintenance: 1 1

j The team noted areas where maintenance support was not timely or effective. The team '

I determined that plant maintenance failed to restore the failed "B" RHR pump to service in a i i timely manner. The unavailability of quality parts, vendor specifications, and repeated post l maintenance test failures resulted in operating with a single RHR pump for over 3 weeks.

l The timely restoration of the RHR pump was important because only the redundant "A" l i

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RHR pump remained available to provide the preferred decay heat removal method. The team also noted a few instances where maintenance staff failed to comply with procedures during the repair of the "B" RHR pump. The team concluded that the maintenance support for restoring the RHR pump to service was not timely.

The team also noted that the poor material condition of severalisolation valves was a contributing cause for too of the events reviewed. Leaking valves allowed nitrogen gas to enter the reactor vessel and water from the RCS to be diverted to the containment sump.

The large number of leaking valves noted during these events may be indicative of a broader problem with the material condition of velves whose primary function is for non safety-related equipment isolation. The failure of these valves to function properly resulted in several operator challenges. The team concluded that the maintenance support for maintaining isolation valves was not effective.

Enaineerina:

The support provided to the o'perators by engineering and technical support (E&TS) was not timely or effective. The condition of the temporary reactor head vent system was significantly degraded. Over the past several years, management failed to effectively respond to previous plant staff concerns and improve the vent header design. The E&TS organization also failed to establish appropriate design criteria or conduct proper system oversight to ensure that the system was functioning properly. The failure to establish a functional reactor head vent systems allowed nitrogen gas to accumulate in the reactor ,

vessel. Absence of a direct means of monitoring reactor vessellevel, for approximately

- 6 days while refueling activities were postponed, complicated the situation for the operators.

Manaoement Oversicht and Controls:

The failure by plant management and staff to fully appreciate the significance of this event resulted in a poor event response and in a delay in initiating an integrated event recovery plan. The team determined that the actions taken during and following this event to determine actual reactor vessel level were not timely. Then actions included taking local reactor vessel levelindication system (RVLIS) and core exit thermocouple (CET) readings i and completing the special test to verify level. Delays wee encountered in reinstalling control room RVLIS and CET indications and aligning a rtactor coolant pump (RCP) for standby service. The actions implemented to monitor the operating RHR pump, following the "B" RHR pump failure, were also not comprehmsive or timely. Delays were also experienced in initiating and staffing the independent review team that investigated these events.

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, TABLE OF CONTENTS

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l 1.0 INSPECTION OBJECTIVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i

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2.0 BACKGROUND

AND EVENT

SUMMARY

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3.0 S AFETY SIGNIFICANCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 a

4.0

SUMMARY

OF CONTRIBUTING CAUSES FOR EVENTS . . . . . . . . . . . . . . s . . . 5 s

5.0 EVE NT D ETAI LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4

. Detailed Sequence of Events ........................................ 12 1

1. Oper ations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 03 Operations Procedures and Documentation ......................... 31 03.1 Reactor Coolant System Draindown ......................... 31 03.2 Reactor Coolant System Vent Header . . . . . . . . . . . . . . . . . . . . . . . . 32

-03.3 Fill and Vent of the Charging System ........................ 33 03.4 Placing Residual Heat Removal Purification System in Service . . . . . . . 34 03.5 Residuel Heat Removal System Operation . . . . . . . . . . . . . . . . . . . . . 35 03.6 Boration injection Flowpath Alignment ....................... 37 03.7 Loss of Residual Heat Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 04 Operator Knowledge and Performance . . . . . . . . . . .................. 39 04.1 Reactor Coolant System Draindown ......................... 39 04.2 Return of the Reactor Coolant Locps to Service . . . . . . . . . . . . . . . . . 40 04.3 Operator implementation of Abno' mal Operating Procedure for Reactor Coolant System Leakage ndication ................... 42 04.4 Placing Residual Heat Removal Pu ification System in Service . . . . . . . 44 04.5 Improper Alignment of Boration Fl >wpath . . . . . . . . . . . . . . . . . . . . . 45 04.6 Failure to Control Maintenance Testing on Containment Spra y Valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 04.7 Operator Response To Excessive Liquid Nitrogen Use . . . . . . . . . . . . . 47 04.8 Event Notifications ..................................... 48 05 Operator Training and Qualifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 03.1 Level Indication and Vent Header System Training During Mode 5 Shutdown Conditions ........................ 49 lI . M aint e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 M1 Conduct of Maintenance ....... ............................ 50 M1.1 Timeliness of RHR Pump Replacement . . . . . . . . . . . . . . . . . . . . . . . . 50 M1.2 Removal and Replacement of the "B" RHR pump . . . . . . . . . . . . . . . . 51 iv

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TABLE OF CONTENTS (CONT'D)

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i M2 Material Condition of Facilities and Equipment ....................... 52 l l

. M2.1 Material Condition of the RCS Vent Header & Isolation Valves . . . . . . . 52 e

l l Ill. Engineering .................................................. 53-E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 53

. E2.1 - Technical Response to the Accumulation of Nitrogen Gas in the Reactor Vessel ........................................ 53 E2.2 Technical Response to the RHR Pump Failure . . . . . . . . . . . . . . . . . . . 54 E2.3 Technical Response to the RHR Heat Exchanger Valve Leak . . . . . . . . 55 E2.4 Review of Generic Information ............................. 56 i E2.5 Outage Planning and Scheduling . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 )

E2.6 Reactor Coolant Vent System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 '

E2.7 Root Cause Determination of the Failure of the "B" RHR Pump ...... 59 I E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . . . . . 60 E7.1 Independent Event Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 E7.2 Management Oversight Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 Figure 1 - The Undetected Nitrogen Gas introduction into the Reactor Vessel from the Charging Line Figure 2 - Chemical & Volume Control System Figure 3 - Residual Heat Removal System (RHR)

Attachment 1 - AIT Charter Attachment 2 - AIT Exit Meeting Slides l

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4 Reoort Details 1.0 INSPEC TION. OBJECTIVES The objective of this inspection was to conduct an independent assessment of the circumstances surrounding the (1) undetected accumulation of nitrogen gas in the reactor vessel head; (2) inadvertent diversions of reactor coolant system (RCS) inventory; (3) failure of decay heat removal system components. Specific inspection objectives ~were provided to the team, by the Regional Administrator, in the AIT Charter (Attachment 1 of

, this report). The inspectors used NRC Inspection Manual Chapter 93800 as guidance for j conducting this inspection activity.

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2.0 BACKGROUND

AND EVENT

SUMMARY

l j Undetected Accumulation of Nitroaen Gas in the Reactor Vessel

! A reactor shutdown was completed on July 22,1996, in accordance with technical specifications (TS), to address design issues with the containment air recirculation fans.

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Following the shutdown, plant management decided to remain shutdown and to start the planned refueling outage.

l The plant was in cold shutdown during this event. The residual heat removal (RHR)

system was in service to remove decay heat. The "A" RHR pump was in service and the

, "B" RHR pump was in standby (later determined to have been inoperable). The reactor l vessel head was installed and the RCS loop stop valves were closed. The reactor vessel I

levelir:dication system (RVLIS) and core exit tt ermocouples (CETs) had been removed from service in accordance with refueling proce dures. The integrity of the RCS and containment boundaries was not established; I owever, controls were in place to  !

reestablish these barriers, if necessary, i

! The chemical and volume control system (CVCS) was aligned to provide the required boric

scid flow path to the RCS. The charging pumps were not operating. The volume control

! tank (VCT) was isolated with a nitrogen cover Jas at a pressure of 30 psig. The VCT was

! pressurized to provide an additional source of snakeup to the RCS.

The RCS was depressurized and the vent head 3r system was in service. The vent header i system was connected to the reactor vessel head using a temporarily installed hose. The i vent header system maintains a small vacuum to remove gasses from the RCS (See Figure 1 1 of this report).

i The available RCS level indications were the hot and cold calibrated pressurizer level

. indications, a temporary standpipe on the pressurizer, and the cavity level instrument. The cavity levelinstrument and temporary standpipe are placed in service during shutdown conditions. The range of the cavity levelinstrument is from the bottom of the RCS loop to

, the top of the refueling cavity. The cavity levelinstrument does not provide a direct

indication of reactor vessellevel. The cavity levelinstrument willindicate the highest

. elevation of water in the RCS. When water is in the pressurizer, the cavity level instrument willindicate pressurizer level.

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, On August 28,1996, nitrogen gas from the VCT began to enter the reactor vessel (See

( Figure 2 of this report). A Nuclear Systems Operator (NSO) was in the process of j realigning certain CVCS valves to establish a new boric acid flow path. During this

evolution, the NSO inadvertently opened the wrong manual valve (BA-V-355). This error l caused water and nitrogen gas from the CVCS system to enter the RCS. Control room operators noticed an increase in RCS level and promptly instructed the NSO to reclose the

valve. Closing the valve should hne stopped the nitrogen gas from entering the RCS; however, the valve failed to fully close and nitrogen gas continued to leak into the RCS.

Nitrogen gas continued to flow through the idle charging pumps and severalleaking valves j into the reactor vessel. The nitrogen gas accumulated in the top of the reactor vessel

head. The rate of addition of nitrogen gas exceeded the removal capacity of the vent l header system. Therefore, a nitrogen gas bubble begsn to grow in the top of the reactor i vessel. The water displaced by the gas bubble was forced out of the reactor vessel and l into the pressurizer. The transfer of water inventory from the reactor vessel to the

, pressurizer resulted in a decrease in reactor vessel level.

, The rate of addition of nitrogen gas to the RCS decreased as the size and pressure of the i nitrogen gas bubble increased. As the differential pressure between the VCT and RCS l decreased, the leakage of nitrogen gas into the RCS also decreased. Likewise, the j increase in RCS pressure caused the removal of nitrogen out of the vent header to j increase. The rate of accumulation of nitrogen in the reactor vessel was decreasing prior j to the termination of this event on September 1,1996.

4 On August 29,1996, in preparation for reacter vessel head removal maintenance activities, operators drained approximately 500) gallons of water from the RCS. Later that day, the decision was made to postpone the re'ueling activities and 1000 gallons of water were added back into the RCS. The pressurizer level and cavity level indications following this addition were approximately equal to the irdications before the 5000 gallons of water were removed. The 4000 gallon difference is indicative of the size of the nitrogen gas bubble in the reactor vessel. Indicated cold caJbrated pressurizer level continued to increase (~8%) over the next 3 days, as nitrogsn gas continued to displace water from the reactor vessel into the pressurizer.

On' September 1,1996, the nitrogen gas supply to the VCT was isolated in an attempt to identify the source of the relatively high nitrogen gas usage. The isolation of nitrogen gas to the VCT stopped the nitrogen leakage into the RCS. The vent header system now had the capacity to remove the nitrogen gas which had accumulated in the reactor vessel. The volume of nitrogen being removed from the reactor vessel was replaced by the water from the pressurizer. The pressurizer level rapidly decreased until the levelindication was no longer on scale. Six RCS makeups were required, totalling approximately 5000 gallons, to stabilize pressurizer level. Stopping the nitrogen leakage and refilling the reactor vessel terminated this event.

Inadvertent Diversions of RCS Inventory The team reviewed two instances where water was inadvertently diverted from the RCS.

The first event occurred on August 22,1996, and the second event occurred on September 4,1996.

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l On August 22,1996, a NSO was conducting valve manipulations to place the RHR I i purification system in service. The RHR purification system is used to filter the water in l 1 the RCS. The RHR system was inservice providing shutdown cooling. During the valve 1 l alig.1 ment, the NSO failed to complete a procedure step that required closing the l l purification pump suction isolation valve from the refueling water storage tank (RWST).

j The subsequent procedure step required opening the purification pump suction valve from the operating RHR system. When this valve was opened, approximately 500 gallons of 1 4 RCS water were diverted from the RHR system to the RWST. The team concluded that l j this failure was caused by inadequate procedures, the absence of a pre-job briefing, and j the failure of the NSO to properly follow the procedure.

i j On September 4,1996, an event occurred where RCS water was inadvertently diverted j from the RCS to the containment sump. Maintenance personnel were starting a preventive l j maintenance (PM) activity on two motor-operated containment spray valves (RH-MOV-34 and RH-MOV-23). The PM required the valves to be cracked off their closed seats. The i valves are used to isolate the RHR system from the containment spray header. A l l downstream manual valve (RH V-23A) was closed to prevent the RHR system flow from i j being diverted to the spray header. Approximately 30-40 minutes after cracking open the 4

valves, a control room operator noticed a decrease in pressurizer level and an increase in f containment sump level. The water was being diverted from the RHR system to the j containment sump. The leakage path was through the cracked open valves (RH MOV-34

& 23), through the closed manual isolation valve (RH V-23A), and through an open drain i line from the spray header to the containment sump. A total of 200-300 go!!ons of RCS l inventory was inadvertently diverted to the containment sump. The operators closed j motor-operated valves (RH MOV-34 & 23) to stop the leakage and refilled the RCS.

l Prior to the September 4,1996 event, the licensee stopped all work that could adversely i

affect the function of the RHR system. The valve PM work was inappropriately performed l despite the stop work order. The team concluded that the isilure to properly implement i

the stop work order and the leaking isolation valve were the causes of this event.

Failure of Decav Heat Removal Comoonents i The RHR system is used to remove decay heat from the fuel when the reactor is shutdown. The system is designed such that a single failure, of certain components, will i not prevent the system from performing its decay heat removal function. At the same time j there was a nitrogen gas bubble in the reactor vessel, one redundant RHR pump and heat

exchanger were not available for service (See Figure 3 of this report).

On August 31, a NSO identified a small amount of leakage from the "A" RHR heat

, exchanger inlet valve (RH-V 791 A) body. The valve was closed to isolate the leakage.

Closing the valve isolated the heat exchanger and removed it from service. The RHR

! system continued to remove decay heat using the redundant "B" RHR heat exchanger.

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4 On September 1,1996, the "B" RHR pump failed to operate after two start attempts After each start attempt, control room operators manually secured the pum.n. ahen the expected motor ampere indications were not achieved. Plant operators were unsuccessful in there attempt to rotate the pump shaft manually. The operators concluded that the pump had become seized and was not operable. The RHR system remained inservice, removing decay heat, using the "A" RHR pump and the "B" RHR heat exchanger.

3.0 SAFETY SIGNIFICANCE ,

The team concluded that the combination of these e94ts was safety-significant.

However, there were no actual consequer.cos < 'hne events on the health and safety of the public and plant staff. This section of tha aspection report provides a summary of the safety-significant aspects of these events.

Plant operators were unaware of the accumulation of nitrogen gas in the reactor vessel or the decrease in reactor vessel level for approximately 4 days. The inadvertent diversion of RCS inventory and the RHR system equipment failures further increased the significance of this event. The team concluded that it was fortuitous that the nitrogen bubble did not reach a size where RHR pump cavitation would occur. Had the nitrogen addition rate been increased or the vent header removal rate been decreased, the nitrogen bubble could have continued to expand in size.

The operation of the RHR decay heat removal system is contingent upon maintaining ,

adequate levelin the reactor vessel. Had the nitrogen gas bubble continued to expand in i size, the nitrogen could have been entrained in the RHR pump suction, causing the operating "A" RHR pump to cavitate. The control room indications for a cavitating RHR pump would be fluctuations in motor ampere readings and a low RHR flow alarm. If the operators properly diagnosed gas entrained in the pump suction, then the correct response would be to secure and vent the pump. However, plant operators would have been  :

presented with conflicting indications regarding pump cavitation during this event. The l pressurizer level, on both the cavity and pressurizer level indications, erroneously indicated that the reactor vessel was full of water. Therefore, plant operators would not have expected that low reactor vessel was causing pump cavitation. If prompt operator actions were not taken, the only operable RHR pump could be damaged by the cavitation. If the RHR system were lost during this event, the time for the RCS to heat-up to 200 'F was approximately 52 minutes.

j. The abnormal operating procedure for a loss of RHR requires that the cavitating pump be secured and vented. However, the location of the RHR pump vents was not optimal and i significant difficulty was encountered during venting the "B" RHR pump following i maintenance to address its seizure. Therefore, an effective venting of a RHR pump may l not have been easy to achieve. Throughout this event, the "B" RHR pump would not have

. been available for service. The operators were unaware that the "B" RHR had seized following its last operation on August 19,1996. The failure of the "B" RHR pump added to the significance of this event. If the "A" pump was damaged, the RHR system would not be available to remove decay heat. All the RHR support system equipment, such as, 3 emergency power, component cooling, and service water were available throughout the duration of this event.

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5 if the RHR system were available to remove decay heat, plant procedures require pressurizing the RCS to facilitate natural circulation cooling using a steam generator as a heat sink. However, a prerequisite for this evolution requires filling the pressurizer using the charging pumps. During this event, nitrogen gas had displaced water in the charging pumps and the pumps would not have operated until the CVCS was vented and refilled.

An additional potential conseque.nce of the nitrogen gas in the RCS was that the gas could have entered the steam generators and obstructed natural circulation. The reactor coolant pumps had been removed from service and would not have been readily available to remove the nitrogen gas from the steam generators. If the steam generators are not effective in removing decay heat, plant procedures direct the use the low pressure safety injection pump (LPSI) to provide decay heat removal.

A decay heat removal path using the LPSI pumps and the RHR heat exchangers would have remained available. This flow path provides water pumped by a LPSI pump from the RWST to the RCS. Reactor coolant is letdown from the hot leg, through the RHR/RCS isolation valves, through the idle RHR pump and heat exchanger, and back to the RWST.

The team concluded that this would have been an :.vailable decay heat removal method.

The inadvertent draindowns of the RCS inventory also contributed to the significance of

! this event. For the specific draindown events that occurred, plant operators stopped the

! diversions before a significant amount of water was removed from the RCS. However, had i the diversions not been stopped, these events could also have caused RHR pump cavitation.

4.0

SUMMARY

OF CONTRIBUTING CAUSES FOR EVENTS The team identified several contributing causes for these events. This section of the report j provides a brief summary of the contributing causes. The contributing causes are

, described in detailin subsequent sections of this inspection report. The section of this

! report that provide supporting information for the assessments are noted in parenthesis.

l 1. Operations l

Operations Procedures and Documentation 1

Inadeouate Procedures l I

The team concluded that poor quality operating procedures were a contributing cause for several operator errors. Poor procedure quality was a contributing cause for the operators' failure to detect the nitrogen gas accumulation in the reactor vessel in a timely manner.

The lack of adequate procedures was the root cause for one of the two inadvertent RCS i

draindown events. The team concluded that the large number of procedure deficiencies may be indicative of a broader problem with procedure quality. The inadequate procedures were allinfrequently used procedures for plant shutdown and draindown evolutions.

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l The operators compensated for a lack of detailed procedure guidance by writing instructions in accordance with administrative control procedure (ACP) 1.2-5.3, Evaluation l l

of Activities \ Evolutions Not Controlled by Procedure. The instructions written in 1

accordance with ACP_1.2 5.3 do not require the same level of review and approval that other plant procedures receive. Operators wrote instructions for the draindown of the RCS l j and the venting of the charging pumps using the guidance in ACP 1.2-5.3. The team 1

j concluded that the guidance written for the RCS draindown did not provide adequate t detail. Specifically, the guidance did not require RCS inventory balances, specify reference levels, or verify proper head vent operation. The team concluded that using ACP 1.2-5.3 to develop guidance for these activities was not consistent with regulatory requirements. I The failure to provide detailed procedures contributed to the failure by plant operators to j {

identify the accumulation of nitrogen gas in the reactor vessel. (Sections 03.1 and 03.3) j The team determined that the normal operating procedures for the vent header system, RHR system, and RHR purification system provided insufficient detail or incorrect

_information. The vent header operating procedure specified using equipment that was not

] installed in the plant, provided inaccurate system sketches, did not specify design requirements for temporary equipment, and did not require periodic monitoring of system j

performance. The RHR system operating procedure failed to provide specific guidance for shifting RHR pumps, establishing maximum allowable flow rates, removing a heat exchanger from service, and venting the pump. The lack of necessary step completion signature locations in the RHR purification procedure contributed to the August 22,1996, RCS flow diversion. (Sections 03.2, 03.4, and 03.5)

Operator Knowledge and Performance Failure to Imolement Procedures The team noted instances where activities were not properly conducted in accoidance with procedures. In other cases, activities were conducted without having a procedure.

On August 28,1996, the failure by a NSO to properly align a boric acid flow path, resulted in the inadvertent injection of water and nitrogen into the reactor vessel. The boric acid flow path alignment was not conducted in accordance with the procedure. This error was '

significant because the valve continued to leak by and allowed nitrogen gas to accumulate in the reactor vessel. (Section 04.5)

A plant procedure required measuring boron concentration in the RCS loops prior to l opening the loop stop isolation valves. The failure to follow the procedure for collecting the boron samples resulted in inaccurate boron measurements. The boron samples are required to ensure that the RCS boron concentration is not inappropriately reduced when the loop stop valves are opened. (Section 04.2)

The team noted that nitrogen gas was isolated to the VCT and the RCS loop stop relief valves were isolated without using procedures. Closing the loop stop relief isolation valves was an inappropriate action because it isolated the necessary pressure relief capability for isolated sections of the RCS. (Section 04.1)

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j 7 Lack of a Questionina Attitude -

! The team concluded that control room indications were available to identify the j accumulation of the nitrogen gas in the reactor vessel. The failure to properly diagnose

! abnormal control room indications was a contributing cause for this event.

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j Plant operators failed to properly diagnose the cause of RCS levelindication anomalies.

1 Pressurizer level indication continued to increase for 4 days without any corresponding additions of water to the RCS. The operators erroneously concluded that the level increase was the result of leaking loop stop isolation valves and manually tightened down on these valves. The operators failed to monitor plant indications, following the completion of this j action, to verify that the desired result of stopping the in-leakage was achieved. (Section j 04.1)

On August 29,1996, approximately 4000 gallons of water were removed from the RCS 1 without a significant change in RCS indicated level. The operators failed to identify or

evaluate this anomaly. The team determined that this anomaly offered another opportunity to have identified the existence of the nitrogen gas in the reactor vessel. (Section 04.1)

The source of the relatively high nitrogen gas usage was not investigated in a timely l manner. High nitrogen gas usage began on August 28,1966, and continued until the source of the leakage was identified and isolated on September 1,1996. Tt.ere was not

an aggressive effort to identify the source of the high nitrogen usage before September 1, l 1996. (Section 04.7)

Inanoropriate D scision Makina The team noted examples where the plant sta ff failed to make conservative decisions.

l These failures to make conservative decisions resulted in an inadvertent diversion of RCS l inventory and improper isolation of overpress are protection for portions of the RCS.

On September 2,1996, the Unit Director plac ed a stop work order on any work that could adversely affect the RHR system's decay hea: removal function. Contrary to the stop

work order, plant work control supervisors ard control room operators incorrectly allowed 1

maintenance activities to be performed on the; RHR system containment spray motor-operated valves. This maintenance activity resulted in the diversion of RCS inventory to 4

the containment sump. (Section 04.6)

A second example of inappropriate decision making occurred on August 29,1996, when operators isolated the loop stop pressure relief valves. These valves were closed in an attempt to stop a perceived in leakage from the isolated RCS loops. The plant operators failed to properly evaluate the consequences of isolating the pressure relief valve for a i water solid system. (Section 04.1) i 4

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8 Inadeauste Pre Job Briefinas The failure to conduct pre-job briefings adversely impacted job performance. The improperly performed evolutions were both being performed by the NSOs for the first time.

On August 22,1996, a NSO failed to properly align valves while placing the RHR purification system inservice. The NSO was unaware of the proper procedural steps that needed to be performed. The lack of a pre-job briefing allowed the NSO to proceed with the valve alignment without the benefit of a thorough understanding of the activity to be performed. (Section 04.4)

The second instance where the failure to conduct a pre-job briefing resulted in poor performance occurred during a boric acid flow path valve alignment on August 28,1996.

During this evolution, the NSO manipulated the wrong valve, that resulted in the injection of water and nitrogen gas into the reactor vessel. By not conducting a pre-job briefing, shift supervision failed to provide expectations and guidance to the NSO. (Section 04.5)

Failure to Reoort Event The team concluded that plant operators failed to make the appropriate prompt report to the NRC following the identification of the nitrogen gas bubble in the reactor vessel on September 1,1996.10 CFR 50.72 (b)(2)(iii)(B) requires a 4-hour event report be made for l "Any event or condition that alone could have prevented the fulfillment of the safety f'metion or structures or systems that are needed to: (8) Remove residual heat." The team de.ormined that the nitrogen gas in the re setor vessel could have prevented the RHR system from removing residual heat. Therefore, this event was reportable. The Connecticut Yankee (CY) licensing staff determined that this event was reportable and a 50.72 report was made on September 11,1996,10 days following the event. (Section 04.8) l Operator Training and Qualifications Failure to Conduct Planned Trainina l

} The team .;oncluded that the failure to complete planned operator requalification training i

may have contributed to the operators inability to properly diagnose the indications of i nitrogen gas in the reactor vessel.

The planned operator requalification training was postponed due to the unanticipated early start of the outage. The training materialincluded a discussion of NRC Information Notice 94 36, Undetected Accumulation of Gas in Reactor Coolant System. This information notice provided information on expected plant indications with nitrogen gas in the reactor

vessel during shutdown conditions. (Section 05.1) i I

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11. Maintenance 4 Conduct of Maintenance

4 l Avoidable Delavs Durina the RHR Pumo Reoair The "B" RHR pump was not returned to operable status until September 25,1996,25 i days following the identification of its failure. The team noted several avoidable delays were encountered during the "B" RHR pump repair. Some of the delays included a lack of quality replacement parts, inadequate vendar supplied information, lack of technical eyaluations for floor block removal, and the absence of appropriate vent locations. Further

]

delays were encountered following pump reassembly when the motor end bell was found 4 cracked. A thorough inspection of the motor in parallel with the pump maintenance could have prevented this delay. Problems encountered as a result of an improperly installed bearing oil seal, also caused further delays. (Section M1.1)

Material Condition of Facilities and F.quipment

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Poor Plant Eauioment Condition 4 The team concluded that the failure of valves to isolate flow and the poor condition of the

temporary reactor head vent system were root causes for these events.

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The team noted several valves in the CVCS and RHR that failed to perform their isolation I

! function. The leaking CVCS valves allowsd nitrogen gas to enter the RCS. Leaking RHR j valves allowed water from the RCS to be diverta d into the containment sump.

i The team also identified significant inadequacies in the design and operation of the

! temporary reactor head vent system. There we e no documented design specifications for

the system. The team identified several design deficiencies including using an inappropriate diameter temporary hose, using diet tape to prevent tygon tubing from kinking, operating the system with known loop seals, and not having the capability to automatically drain the system. The team also identified several deficiencies with the operation of the system including condensate ir loop seals, lack of appropriate levelin moisture dropout tanks, and kinked tygor' tubing. The team determined that the overall absence of quality in the installation and operation of the reactor head vent system I seriously degraded its ability to remove gas from the reactor vessel. A properly designed, j maintained, and operated reactor head vent system may have removed the nitrogen gas from the reactor vessel and prevented this event. (Section M2.1) 4 1

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! 10 i 111. Enaineerina i

Engineering Support of Facilities and Equipment Untimelv Technical Response

The team concluded that certain aspects of the technical support response to these events

! were not timely. The following are examples where the team concluded that a more timely .

• response would have been appropriate. (Sections E2.1, E2.2, and E2.3) -

l Local RVLIS and CET readings were not taken until September 3,1996, two days following the identification of nitrogen gas in the reactor vessel. The local RVLIS and CET i- readings were taken with available test equipment and existing procedures. The team

i. noted that confusion among plant staff on the ability to collect this information contributed to this delay.

i 1 A reactor coolant pump (RCP) could be used to provide forced circulation of the RCS in the *

, event that the RHR system became unavailable. The RCPs could also be used to fill the i steam generator U tubes if nitrogen gas accumulates in the steam generators. A RCP and '

l associated support systems were not aligned for service for 2 days following the i identification of nitrogen gas in the reactor vessel. The alignment of the RCP was not a

, simple evolution and required considerable time. The team determined that the technical i response to this event was not proactive in aligning contingency equipment to respond to a i potential loss of the operable RHR pump.

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l The technical support organization developed tvio special test procedures to determine 1 reactor vessel level and to verify that the reactor vent was not blocked. These tests were not completed in a timely manner. Tha tests wore not completed until September 5,1996, four days following the event.

j The team concluded that the technical support 'or developing additional monitoring of the only operable RHR pump was inadequate. The :echnical staff provided no j recommendations for additional monitoring of the operating RHR pump for 6 days following j the identification of the failure of the "B" RHR pump. The additional monitoring that was s implemented was to have the NSO observe pump operation 4 rather than the normal 3

! times per day. The team concluded that additional non-intrusive pump performance monitoring could have provided additional assurances of proper RHR pump operation.

L Poor imolementation of Generic Information

The United States Nuclear Regulatory Commission (NRC) issued information Notice 94 36

! describing a 1993 event at another nuclear power plant where nitrogen gas from the VCT accumulated in the reactor vessel. In response to the notice, the technical support staff at i CY conducted a detailed evaluation of the information notice and provided several

recommendations in 1994. The recommendations included maintaining at least 1 channel

j. of RVLIS available when the VCT cannot be maintained isolated. The evaluation also recommended that " Operations Department should review the applicable shutdown i

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11 procedures to ensure that guidance is given to operators warning them of the potential for I gas void formation in certain plant conditions and provide steps to ensure that monitoring for a void occurs." These recommendations were not adequately implemented prior to this evm.t. The team concluded that implementing these actions may have provided operator-the information needed to have identified and mitigated this event. (Section E2.4) l Weak Enoineerina/ Operations interface The team identified several occasione when operators fa!!ed to solicit assistance from the E&TS organization. In other cases, E&TS failed to be proactive in providing operator j support. The team determined that soliciting E&TS assistance would have been appropriate for identifying the source of the excessive nitrogen gas usage, evaluating the root cause for the increase in RCS leul, and developing procedures for draining the RCS ,

I and venting the charging pumps. The E&TS organization failed to provide operators with en adequate reactor head vent system design or direct indication of reactor vessel level. l Both of these deficiencies were previously identified in employee suggestions and generic information evaluations. In addition, a proactive E&TS organization, that was aware of l system configuration changes, could have prevented the inappropriate tnaintenance activity on the RHR spray valves and the inappropriate closure of the RCS loop stop relief valves.

The absence of an effective operations and E&TS interface was a contributing cause for these events. (Sections 5,04.1, and 04.3)

Inaooropriate Outaae Schedulino Decisions The team concluded that the decision to postpone refueling activities over the long weekend did not adequately include shutdown risk considerations. The outage schedule was delayed at a time when reactor vessel indications were disconnected and the steam

! generators were isolated. The team concluded that while the licensee had completed u i detailed outage risk evaluation, that these insights were not properly considered when the i decision was made to postpone the outage. (Section E2.5) l i Lack of Direct Reactor VesselIndication

! The absence of direct reactor vessel monitoring instrumentation had an adverse affect on j the operatnrs ability to monitor reactor vessel conditions. The cavity and pressurizer level ,

j instruments do not provide a direct indication of reactor vessel level. The RVLIS, which -

l does provide a direct reactor vessel level indication, was disconnected in preparation for )

l refueling maintenance activities. The CETs are another direct indication of reactor vessel j conditions. The CETs were also disconnected during the duration of this event.

The RVLIS indicates reactor vessel level at discrete elevations. During this event, the RVLIS would have indicated abnormal reactor vessel level when the actual level dropped

! below the reactor vessel flange. Following this event, local RVLIS readings were collected and a temporary jumper was purchased and installed to provide RVLIS indications in the j control room. A second jumper was installed to provide CET indication in the control l room. The team concluded that the failure to provide operators with direct indications of l reactor vessel conditions was a contributing cause for this event. (Section 5, E2.1) l n r- *

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,2 l Quality Assurance in Engineering Activities Slow initiation of Event Resoonse Team a

{ The evaluation of these events by the CY staff was delayed. The initial response by licensee management was to conduct the event review using plant technical support staff.

This team was later disbanded and an independent Review Team (IRT) was chartered.

This team was not fully staffed and functioning until September 4,1996. The team

] concluded that the initiation of the IRT should have been more timely. The !RT had not 1 completed its evaluation prior to the conclusion of the AIT onsite inspection activities on September 16,1996. Therefore, the AIT did not assess the quality of the IRT efforts.

l (Section E7.1) l 5.0 EVENT DETAILS i

Detailed Seouence of Events i The team developed the following event sequence based on interviews, review of plant logs and operational data, and information from the licensee's IRT. Team comments and j assessments regarding the events are provided in the bold and italicized text.

UNDETECTED ACCUMULATION OF NITROGEN GAS IN THE REACTOR

, VESSEL i

! Julv 22.' 1996 The reactor is shutdown to address design issues with the containment air recirculation fans. Following the shutdown, plant management decides to start the planned retueling outage early. (The endy start to the outage resultedin delaying

, requaHfication traldng. The team noted that the planned requsHHcation traidng l hchded a discussion of the indications of dtrogen gas in the mactor vessel. The

team concluded that the delay k conduedng tNs training was a contdbuting cause for tNs e ent.]

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July 27.1996 l

The primary vent header is placed in service per normal operating procedure (NOP)

2.9 6, " Primary Vent Header Operation." /The vent header system is used to remove radioecdve gases imm the RCS dudng shutdown conditions. A temporedly instonedhose k used to connect the reactor vesselhead to the vent header. The l reactor head vent hose is only instaned when the RCS is depressurked. TNs system k not a safety related system. TNs system was not intended to remove the

large quantity of gas that was Jenking into the RCS during the strogen intrusion

) event. The inaNNty of tNs system to remove all the dtrogen gas from the reactor vessel, resultedin the accumulation of dtrogen in the reactor vessel. Several design and operationaldeHclencies were identified with this system by the AIT.]

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l 13 Auaust 12.1996 j 1241 The cavity levei indicating system (CLIS) is placed in service. /The CLISprovides

controlmom levelindication for the refuenny cavity when the cavity is Mooded. It

, also provides RCS levelindicadon when the nsselheadis kstaMed dudng l shutafown conditions. The CLIS variable leg is attached to a drain connection for

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the Loop #1 het leg (LT-151A). The miennce seg is attached to the reactor nssel head (LT-151C2) and compensates for vacuum in the wat heeder. The variebte and

- mierence ing transmitters are electrordceMy subtracted to produce the cavity lent signal. The cavity involindicadon readout is in inches elevation abow sea level. A change of about 3.9 inches in cavity levelis equivalent to 1% change in pmssurizer lent. The instmment scale is 170 to 600 inches elevation.]

i 1638 The RCS loop stop valves are closed in accordance with plant procedures. / Closing the loop stop valves isolates the steam generators and reactor coolant pumps from the mactor vessel. During tNs event, the cloxdloop stop valves prevented nitrogen gas from entering the steam generators.]

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Auoust 26.1996 0000 The liquid nitrogen storage tank level decrease is 2.5 inches during the previous 24

! hours. /TNs data establishes the basefine rdtrogen usage in tNs made of operation.

The nitrogen usage efter 8/28/96 increases substantisNy.]

i Auaust 27.1996 1

0034 The RVLIS and the CETs are disconnected. /TMs was a scheduledmaintenance .

i activity in preparation for the removal of the reactor vesselhead. The RVLIS i system is the only diact reactor vessellevelindication. The team concluded that l the lack of a dinct reactor vessellevelindication was a contributing cause for this

} event.]

I Auaust 28,1996 (Wednesdav) i

The Technical Specification required boration flow path is from the boric acid mete. ring tank (BAMT), through the CVCS metering pump, and into the RCS. /The licensee is required to maintain one boric acid flow path available during this mode of operation. It was a misalignment of valves, while establisNng a new boric acid

! Row path, wNch caused tNs event.] '

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j 0628 The cavity level indication reading is 318 inches elevation. /At tNs cavity level

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kdcation, the indicatedpossurizerlevel would be approxknately (~) 4% and the l water Arvelin the reactor wssel wouldbe ~43 inches below the top of vesseland

~48 inches aban the mactor wesel flange.]

0630 The Unit Supervisor (US) briefs the operators and indicates that preparations are to

be made to run the 2B emergency diesel generator (EDG). Procedure SUR 5.1-159B 1 is used to align the koration flow path. [A new bodc acid flow path was requimd,

) because the metedng pump wm not be operable during the EDG survenence. The i

Mcensee considers the EDG inoperable during the surveMance test. The ernergency

! power for the metering pump is suppued by the 28 EDG. Themfom, the metering pwnp is aAno considendinoperable.]

j 0800 A NSO is assigned the task of aligning the boric acid flow path. A formal pre job briefing is not pertormed. (Thw Udt Supervisor (US) beHond that the NSO was i ordy to wrify that the existing valve Alneup was cormet. The NSO beNend he had been given the authodty to man (pulate valves as necessary. The team concluded i that the faMure to conduct a formalpre-}ch brieSng was a contnibuting cause for the 1 error made by the NSO dudng the valve aHganent. TMs was the first time that the NSO hadperformed tNs particular veh.w aHgnment.]

l j The NSO simultaneously opens both valves BA-V 354 (Blended makeup to the VCT) j and BA V 355 (Blended makeup to the charging pump suction). /Thepmcedum

does not aHow thsse valves to be openedat the same time. The errorin aMgdag thesa valves slowedidtrogen gas to Josk into the RCS. The strogen gas pressure l k the VCT viss ~30 psidgreater than the RCS pressum. TNs pressure diffemntial

provided :ne motive force to dscharge water and rdtrogen gas into ths RCS. The l_ waterin the CVCS charging header was dischargedinto the RCS. After sH the i water was discharged from the CVCS system p4dng, nitmgen gas began flowing

' hto the RCS. The hot caHbratedpressudserinvalindication increased from 0 to 10%. The team concluded that tNs was an example of an operator fsMing to follow

! procedums. A procedum requirement to contact the controlroom before

! manipulating valves was also not knplemented.]

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i The HIGH CAVITY LEVEL Alarm annunciates. The reactor operator, responds by

! increasing the alarm set point. The alarm came in a second time and the reactor operator contacts the NSO performing the boration flow path verification. Valve

BA V 355 is ci> sed with the assistance of a valve wrench. [Following tNs event,

, valve BA V 355 continues to leak continuing the addition of rdtrogen gas kto the l reactor vessel.]

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! 0845 The operators develop a procedure to vent the charging header and "B" charging pump using ACP 1.2 5.3, Evaluation of Activities / Evolutions not Controlled by Procedures. (The charging headerrequired venting because retrogen gas from the VCT had displaced the water in the charging header. The team concluded that it was inappropdate to use ACP 1.2-5.3 to develop this procedum. Procedures developed in this marmer do not receive the same level of review and approval as .

Other station procedures.]

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0900 An adverse condition report (ACR) 96-0946 is prepared to document the valve manipulation errors made during the boration flow path alignment. //t was noted that wide sNfting baration flow path, VCTgas mond charging header waterinto the RCS. Pressuriterlentincreased about 2% and VCTpressure decreased about 2 psi with no Aent decrease. The teem concluded that the operating cmw shouW i han soGcited appropriate techdcolsupport to fumy understand the cause and knpact of tNs event.]

The unit supervisor has a NSO verify that the condensate storage tank rupture disc or its supply is not leaking nitrogen gas. Nitrogen gas use increases to 9 inches in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. /TNs is the first documentedindication that the operators wem investigating the high dtrogen gas usage. The next documentedindication occurs on September 1,1996, when the VCTdtrogen supply was isolated. The team conchded that the plant operations stafflacked an appropriate questoodng attitude and fsBed to so6 cit techdcol support in resolving the abnormaHy high dtrogen gas usage.]

1333 The NSO completes venting the charging pump suction header. /The header required unting because dtrogen gas had entered the charging pumps when the inadvertent vain madpulation (event time 8/28 at 0800) nHowed the nitrogen gas from the VCT to enter the charging pump header.]

1422 The levelin the pressurizer is increased approximately 2% with the metering pump to ensure the charging header was full af water. /The.veteringpump was run to transfer any left over dtrogen gas in th ' charging headerinto the RCS. This was performed to ensure the charging hende r was fiWed with water.]

1602 The boration flow path alignment is co npleted. /The sligrwnent was made using a different valve Kneup independent of valves BA V-354 and BA V-355. The new

! boration Mow path included the use of Itw "B" charging pump. However, nitrogen was continuing to enter this charging p imp suction due to leakage through BA V-l 355. The team concluded that the creditsd techdcolspecification required boric i acM tiow path was not operable at this time. The flow path remainedinoperable

, untR the strogen was isolated from the VCTand the CVCS system vented on

, September 1,1996.]

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1805 The cavity levelindication rea: ting is 323 inches elevation. /This is a Sinch elevation increase since the 0628 cavity levelreading. It was expected that cavity level wouWincrease slightly as a result of the water added to the RCS during the valve aHgnment enor and the CVCS refHl.]

! 1858 The 28 diesel generator surveillance test is performed. /This was a routine surveHlance test that required the &nge in the boric acM flow path alignment.]

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l Ayoust 29.1996 (Thursdav) 4 j 0600 The liquid nitrogen storage tank level decrease is 9 inches over the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (TNs reposented 3 times the nonnelusage recorded on 8/25. The team

} concluded that the piant staff fared to adequately huestigate the source of the Ngh l

Ntragen usage.]

0646 The cavity level indication reading is 337 inches elevation. /TMs represents a 14

inch (~3% pressurizerlevel) kcrease in cavity level skce the 8/28 at 1805 modng.

l This increase was caused by the accumulation of the Ntrogen gas in the reactor wssel. The water from the reactor vessel was dsplacedinto the pressurizer. The

, incmase in waterin the pressudrer caused the pressurizer level to heresse. The l team concluded that the plant operations staff did not adequately evasunto the

} cause of tNs involincrease.]

l 1025 Control room operators began intentionally draining water from the RCS The water j was being removed to reduce the water level in the reactor vessel to approximately the reactor vessel flange. The reduction in level was necessary to support the conoseal removal maintenance activity. A step list was developed by operators to drain the RCS. The step list was developed in accordance with ACP 1.2-5.3.

Pressurizer level was decreased from approximately 10% to 2%. /The team l determined that the Mcensee faned to have the requiredprocedum for draining the ,

i RCS. The step Hst developedin accordance with ACP 1.2-6.3, didnot require the i j same level of review and approval a hat a normal operating procedum receives. The team concluded that the lack of deraRk the step Mst was a contdbuting cause of l tNs event. The step Mst did not provide inventory belances, specify niismnce invols, or vedfy the operation of the vent header. A detanedprocedure for draining i

the RCS couM have resultedin a more tiirnely Mentification of the nitrogen gas in the

, reactor vessel. The team concluded that this was an exampie whom operations

faned to soWelt assistance from E&TS.)

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1120 Draining of the RCS is stopped at a cavity levelindication of 299 inches elevation.

j The Shift Manager stops the draindown to confirm agreement between various level indbators. [At ~301 inch nievation thepmssurizerlevelindication wouMgo off scair low. The reactor vessel flange elevation is - 270 inches.]

1400 The cavity level indication is ~ 321 inches (from ~ 2% pressurizer level to ~4%

pressurizer level). (TNs increase k cavity level from -299 inches at 1120 was caused by dtrogen gas entedng the RCS and displacing RCS waterinto the

! pressudier. The operators wem unaware that the strogen gas in the reactor was l the cause of the levelincrease.]

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! 17 1427 The loop stop overpressure relief valves are closed and the stop valves are manually

tightened to prevent inleakage from the loops. /These veins were closed without proceduralguldence and no enginnedng evaluation was performed. The team i

conckded that tNs was also an examide of poor deci:lon making. The loop entpressure reWef valves were pmviding the overpressum protection for the l\ leolatedportions of the RCS.]

The loop stop veins were known to leak dudng previous outages. The operators fanied to adequately review the msults of the valve closum to verify that the

, in-leakage had stopped. If the operators hadmulewed these actions they would i have concluded that loop stop valve leakage was not the source of the level kcmere. The team concluded that tNs was a missed opportudty to how Mentified

the strogen leakage. Operations shouW han soWcited E& TS assistance to evaluate l the increase in pressudzerlevel.]

1455 The RCS draindown continues to the target value of approximately the reactor

vessel flange. /The flange elevation is approxhnstely 270 inches. The indicated

pressurizer invol wouW be off scale low at tNs elevation. Water wouM remain in l the bottom of the pressudzer and the surge Nne wouW remain full.]

j 1609 The RCS draindown is stopped. The cavity level is ~265 inch elevation. /The team ndfied the licenste estknates that a totalof ~5000 gaMons of RCSinventory was removed during the draindown evolution, based on the aerated drains tank level i increase.]

1800 The operators noted that cavity sevelir:lication is increasing and pressurizer levelis

{i coming on scale. During this abnormalty, an NSO verified no water in the head i vent, no changes in VCT pressure or temperature, and no change in RWST level.

i The operations crew incorrectly attribu ed the inventory increase to leakage from the loop isolation valves and the overp. essure relief valves. /The team concludedet tNs point the operators should han sorcited techedcol support assistance. With indicated cavity level at 282 inches, pressudzer level shouW han been off scale bow. The cavity levelindication should not han been increasing.]

l 2000 Cavity level indication reading is ~ 299 inches elevation. No water was added to j the RCS and cavity level is back to the elevation it was at before the second i draindown, /The team concluded that the controlroom operators failed to adequately evaluate this increase in Indicated cavity level. The increase was being

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c0used by the Strogen gas displacing vesselinventory into the pressurker.]

Plant management makes the decision to suspend refueling activities over the weekend. The operators were directed to refill the RCS. /The team concluded that i' the decision to delay refueling activities did not property include risk considerations.

A previous analysis of outage activities indicated that this was a relatively Ngh risk

. period during the outage. The team concluded that to discontinue outage activities at tNs time was inappropriate.]

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18 i 2030 Operators add ~ 1000 gallons of water to the RCS in three separate makeups.

After the makeups, the cavity levelindication is at ~316 inches elevation. Before i' the RCS was drained the cavity level indication was at ~324 inches elevation.

(Ginn the decision to delay afueHng activities, the decision to restore RCS level

one the weekend was appropdate. Howent, a 4000 gaNon disempancy existed between the 5000 geRans removed and the 1000 gaMons added to the MCS. A i alght kch diffennce in cavity level (dfierence between before and after ktdown and makest.)hdication would on!y account for ~ 500 gaMon kwntory dfference.

The team concluded that the lack of an innntory balance in the draindown step Est

and the faMure of a questioning attitude by the plant operations staff eMowed tNs 1 dsompancy to go undetected. The 4000 gaMons dscrepancypmvfdes an Indication i of the volume of the dtmgen gas bubble k the reactor vessel.]

l Auoust 30.1996 (Fridav) 0000 The liquid nitrogen storage tank level decrease is 18 inches over the previous 24

. hours. (The strogen gas is used for many purposes throughout the plant. The

\ daHy change in Ntrogen gas use appests to be the result of strogen being suppHed i to various other equipment. However, the majodty of the nitrogen gas was leaking l klo the RCS.]

0416 The four loop overpressure isolation va..ss are opened. (This reestablished the i normal valve Kneup. These valves had hoen closed at 1427 on September 29,

during the attempt to stop perceivedleshage from the steam generators. These valves were inappropriately left closed for ~ 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.]

0719 The cavity level indication reading is 320 inches elevation. /This is a 4 inch increase in cavity icvelreading since 8/23 at 2030. The cause of this increase was the increase in nitrogen gas accumulatfor in the reactor vessel.]

l 1800 The cavity levelindication reading is 33C inches elevation. (The cav/tylevel kdication continued to increase as nitrog en ga; continued to accumulate in the reactor vessel.]

, Auoust 31.1996 (Saturdav) 4 0000 The liquid nitrogen storage tank level decreases 19 inches over the previous 24

hours. (The operators knew that nitrogen usage hadincreased without explanation, and that cavity level was trending in an unexpected manner. This information was

< not appropriately acted upon.]

1 0800 The cavity levelindication reading is 338 inches elevation. /The cavitylevel indication cantinued to increase as nitrogen gas continued to accumulate in the reactor vessel.]

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19 i 1600 The cavity level indication reading is 344.8 inches elevation. /The cavitylevel l indication continued to kcrease as dtrogen gas continued to accumulate in the l m ector w ssel.]

4 l September 1.1996 (Sunday) 2 0000 The liquid nitrogen storage tank level decrease is 9 inches over the previous 24

} hours. [The me}ority of the Stragen gas was bekg suppnied to the VCT. The Stragen from the VCTseekedkre the reactor wssel. The wnt hendersystem was .

mmoving most of the dtrogen gas from the mactor vessel.]

l 0600 The cavity level indication reading is 34d.1 inches elevation. /The dtrogen gas i bubble in the reactor wsselhead appears to be maching an equniflviurn condition i when the dtragen go'sintenhage equals the venting capacity. The cavitylevel indicadon has only incmased ~ 2 kches k the poi: 14 frours.

. The cavity levelindication hadincmased from 316 inches after the last makeup on l 8/30. TNs is equivalent to en - 7.6% dse in pmssudzerinvel. At tNs point the l mactor wssellevelmachedits minimum sevel of appmxknately 39 inches below the

reactor vessel 91ange. TNs is ~21/2 feet abow the top of the RCSloops and ~ 7 l feet above the top of the fuelassembnies.]

! A NSO informs the US of the abnormally high use of nitrogen gas. Plant operators i

systematically isolate and unisolate nitrogen gas supplies to various equipment. /A l procedure was not used for Mentifying orisolating dtrogen to various components.

l The system engineer was not hvotnd in this activity. The team concluded that

! tNs was an example of operators conducting acdvities and evolutions without i procedums. In addition, tNs is also en exemper where operators failed to soNcit appropdate E&TS assistance.]

l 0900 The nitrogen gas supply to the VCT is isolated in response to a NSO report of high

nitrogen flow noises through the VCT nitrogen regulator. When the nitrogen gas is i isolated, the cavity level indication decreases and pressurizer level decreases off l

scale low. (The pressurizer level decmases because the nitrogen gas in the reactor vesselis being displaced by water from the pressurizer. When the dtrogen gas entedng the mactor wssel was stopped, the vent header removed the Strogen gas that had accumulatedin the mactor vessel.]

The control room operators directed the NSO to reopen the nitrogen supply regulator to the VCT slowly. This evolution is completed in -20 minutes. (The

, operators stated that they wanted to retum the system configuration to that which i existed before RCSleveldecrease.]

I- 1010 The operators commence adding water to the RCS using the RHR suction from the l RWST (MOV-21) to refill RCS. (A total of six makeups were needed during the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to stabalze pmssurizer level. Each RCS make-up filled the pressurizer to i approximately 10% After the make-up was stopped, the pressudzerlevel l

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20 decreased off scale low and another refMI was requimd. The sixth refm resulted in a stable 10% pressudrerlevel. The levelin the pressurizer decreased as nitrogen gas in the mector vessel was bekg removed by the vent header. Water from the pressuriter was replechg the dtragen gas in the reactor vessel.]

1026 The operators believe that a RCS leak has occurred and implement abnormal operating procedure (AOP) 3.2 31 A, Reactor Coolant System / Refueling Cavity Leak (Modes 5 and 6). A NSO is dispatched to inspect various systems for potential RCS leakage.

1029 A NSO drains a small amount of water from the reactor head vent hose. /This kdicated to the operator that the reactor wssel was fuM. TNs water was Mkely condensation ham the vent heeder and not RCSinventory. At tNs point the reactor vesselhond was not completely fuM. At 10% pressuriterlevel, the headis not expected to be fed.]

l 1030 Valve BA-V-354 (Blended makeup to the VCT) is closed to isolate suspected  !

nitrogen flow through valve BA-V 355 (Blended makeup to the chargir'g pump i suction) (TNs provided a two valve isolation of the nitrogen leak path. In addition, the VCTdtrogen pressure was reduced to about 4 psl.]

1057 The operators comolete the leak investigation using AOP 3.2-31 A, Modes 5 and 6 Reactor Coolant System Leak. No leakage is identified from the RCS. Operators direct attention to gas flowing into the RCS. (At tNs point, the operators meMzed that nitrogen gas in the mactor vessel ned caussd the decrease in pressurizer level.]

1255 Operators complete the makeup to the RCS. The estimated amount of water added is -5,000 galIons from the RWST. [TNs inventory addidon replaced the 4000 gaMons lost durkg the draldag evolition on 8/29. The additional 1000 gaMons raised the pressuriterlevelabove that which existed before the event began on 8/28.]

1353 The operators document the receipt of the cavity low level alarm and use of AOP 3.2-31 A in ACR 96-0966. [The operators fsMed to appreciate the importance of these events. No effort was made to contact techdcolsupport for assistance. The team concluded that tNs was an example where operations failed to soucit the proper management and techdcol support.]

4 1800 The cavity levelindication is 348.8 inches elevation.17his level wouldbe equivalent to a pressuriterlevel of ~ 12% and a vented reactor vessel would be

nearfy fuM.]

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l 21 EVENT RESPONSE Seotember 1,1996 i l

1900 A conference call between licensee and NRC management is conducted t'o discuss f the nitrogen gas event and other equipment failures.  ;

I Seotember 2,1996 (Monday Labor Dav) l 1200 A conference call between licensee and NHC management is conducted to discuss compensatory actions taken in response to the inoperable RHR pump and heat exchanger.

1400 The NRC AIT members arrive onsite.

The Technical Support Manager establishes an initial root cause team to review the nitrogen inleaksge event and calls in the team members.

September 3,1996 (Tuesdavl The Re0i on i Regional Administrator establishes the AIT charter to formally

investigate these events.

1000 A walkdown of the reactor vessel head vent header system by the NRC team

! identifies a kinked tygon tube, lack of adequate levelin the loop seal of dropout

! tank TK 861 A; and water in the vent line at the manometer in containment. The i

manometer in the Primary Auxiliary Building (PAB) indicates a vacuum of 14 inches which ir, normal. }The capabnifty of the wnt header system to mmon the nitrogen

from the reactorhead was sedously degradedby these conditions. The team i

concluded that the faMure to propedy design, insteN and maintain the wat header system was a contdbuting cause of this ennt.]

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1200 Instrument and Control (l&C) technicians locally collect RVLIS readings using available test equipment and procedures. All 6 level indicators in the plenum region

. indicate full, the lower head probe (#2 at flange level) indicates full and the upper

} most probe (#1 Approx. 8" from top head) shows "some" isvel indication (train A was uncovered and train B was covered). /The team concluded that the failure to conect this hrfonnation sooner was an example of the lack of a comprehensin and deely technicalresponse to this ennt. The Mcensee appropriately continue to -

pedodicaMy take RWIS readings.]

, Executive Vice President - Nuclear directs the Nuclear Safety and Oversight organization to establish an IRT. The IRT purpose and scope were established.

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22 l i Seotember 4,1996 l j The IRT membership is finalized, a kickoff meeting held, and data-gathering I j commenced.

C532 Three CETs are placed in service. The CETs provide control room indication of i temperatures above the fuel. lin the event the RHR system was lost, the CETs l wonidproulde operators knportant khwmation on the k-nssel condtlans. The \

team conckded that the technical support orpardzation shouid have been more

&nely k pmmiding CETindcation.]

1600 The #3 RCP is placed in standby. lin t/m event the RNR systern becomes

unavaHeble, the RCP could be piecedin service to promide forced ckculation. The i RCP used k cor$ unction with the steem generator cou6d be used to mmove decay i

heat. A second decay hent removalmethod would be the use of the LPSIpumps and RNR heat exchangers. A closed decay heat removal cycle can be estabWshed with these systems and the RWST. The team detemdned that placing an RCPin \

standby should have been completed at an eerfler time foWowing the event.]

l September 5 19.9.6

A number of reactor head vent system fittings are replaced to improve system l 1 perfa

mance. (The team noted that this change was performed without any

technical mview or official documentation.]

! The IRT is fully statfed. /The Noensee staffed the IRT with personnel without direct l responsibiWty for normalplant functions.]

I l 1500 A specin1 test is performed to verify reactor vessellevel. This was performed using l procedure ST 11.7-197, Determination of Reactor Vessel Level. (The reactor required about 2236 gal (based on metering pump run times and flow rate),1074

gal (based on increase in Li-402, pressuriser cold caNbrated indication),1,000 to l 2,000 gal (based on RWSTlevel decrease) before water started coming out of the

! vont hose. Cavity level system indicated an increase inun 341 to 390 inches. The i team conckded that these tests should have been completedin a more timely

! manner.]

2101 After ST 11.7-197 was completed, the licensee drained about 725 gallons to the

! ADT. Pressurizer level decreased from 22% to 15% and cavity level changed from j 390 to 360 inches.

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September 6,1996 A calculation to determine the minimum reactor vessel water level during this event is completed by operations. The calculation makes two initial condition assumptions that (1) the reactor level starts at 49% and (2) the head is full. For the first case, the vessel level was calculated to drop to 39 inches below the reactor vessel flange. Using the assumption that the vessel is full results in a vessel level decrease to 20.5 inches below the vessel flange. (The fesa concduded that i i

the fkst case with reactor kvalstarting at 49% was the more appropriate assanption. A kval of 39 kches below the reactor nssel Range is :f'at to

' the RCS kvel bekg ~21/2 feet abon the RCS kops and ~ 7 feet abon the top of the fuelassembWes.]

i 1330 l&C reports RVLIS "A" and "B" channels both indicate 100% vessel level. /TNs )

kdicates that the reactor nssel was fuM on September 5,1996.]

September 10,1996 The IRT charter is approved. The data-gathering phase of the investigation is nearly completed and a time-line of significant events is being developed.

September 11,1996 1504 The licensee makes a 10 CFR 50.72 event notification to the NRC on the introduction of nitrogen into the reactc r vessel displacing inventory into the pressuriter. (The team concluded that tNs not/Heation shouWhan been made on September 1,1996.]

2018 The necessary electrical cable was pur::hased and Train "B" of RVLIS was placed in service. The RVLIS indicates 100%. The #W.lS system indication provides the operators direct nsselhuMcation. The instsNetion of tNs}umper wouWhan provided the plant operators kformation that couW han been used to Mendty ano prevent tNs event. A }umper of tNs type was discussedin the Noensee's evaluation of NRC Information Notice 94-36. The team concluded that the lack of direct nsselindication was a contributing cause for tNs ennt.] l lNADVERTENT DIVERSION OF RCS INVENTORY tAunust 22.19991 Auoust 22,1996 ,

0740 The RHR purification system is shutdown to support a local leak rate test (LLRT).

(The RHR purf6 cation system is used to fMter the RCS when the RHR system is l kservice.]

24 2100 The LLRT is completed and operators prepare to restore the RHR purification system to service. (A pm}ob bntefkg of the NSO for the aMgnment of the RHRpurification system vehes was not performed. The NSO was performing tNs task for the first time. The team determined that the fanun to conduct a pow}ob brieRng was a conenibuting cause for tNs event.]

2110 The NSO opens the RHR purification pump suction valve from the RHR system (RH-V 874A) in accordance with NOP 2.7-4, Attachment 4, step 1.5. The NSO does not observe the expected purification system pressure increase and continues to open valve RH-V 874A one full turn. (The NSO was unewsm that step 1.4 of NOP 2.7 4, wNeh closes the RHitpuri6cadon pump suction vatw ham the RWST(PU-V-261A), had not been completed. The fahne to close vehe PU-V 261A, before opening vahe RH V-874A, usulted k a drect Row path from the operating RHR system to the RWST. Approxhnstely 500 ganons of RCS water were hadvertently dverted to the RWST. The team concluded that the fanure of the NSO to crmtly knpiement tMs procedure was a contnibuting cause of tMs event.]

2115 A CAVITY LOW LEVEL alarm annunciates in the control room. The control room operator observes that pressurizer levelis decreasing. The control room operator immediately directs the NSO to close valve RH V-874A. Closing valve RH V-874A terminates the diversion event. (The controlmom operators prompdyident/6ed and termkated tMs event. The faHure to take prompt operator action couW have resultedin RHR pump cavitation.]

! 2130 The NSO returns to the control roam and discusses the event with the US. The

! operators conclude that the NSO isiled to complete the proper procedure step i sequence. (The team concluded that the poor quauty of the NOP was a contdbuting l causes for tMs operator error. NOP 2.7 4 dd not provilde step completion signatum locations for steps 1.3 or 1.4. AM the otherprocedum steps were provided with a l' signatum location to indicate the step was complete. The absence of the step completion signatum contributed to the NSO entering the procedum et the wrong step.]

! 2222 The valve alignment for RHR purification system is completed correctly and the s*, stem is returned to service.

i 2234 An ACR 96-0926 is written to document this event.

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l 25 INAD_VERTENT DfVERSION OF RCS INVENTORY ISeptember 4,1996)

Aucust 18,1996 [

l A surveillance test is completed that strokes open and closed the RHR containment spray valves (RH-MOV-23 and RH-MOV-34). A downstream manual isolation valve (RH V 23A) is closed to prevent water from the RHR system from entering the containment spray header. There is no leakage through valve RH-V-23A identified during this test. (TNs k s&nl6 cant because the manualisoladon RH-V-23A k the make dket Aneks by and causes the RCS Row dfuersian that occurs on September 4,1996.] >

! Auoust 20,1996 1

2 A clearance is written, but not applied, for conducting a preventive maintenance activity on valves RH-MOV-23 and RH-MOV-34.

Aunust 29,1996 l 1035 The work orders to disassemble valves RH-MOV-23 and RH MOV-34 are released j ay operations to electrical maintenance.

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i Seotember P 1996 I The Unit Director / Operation Manager ist Jed a partial stop work order on all work i that could affect the RHR system or RCE inventory. The communications of the i stop work order are done mostly by phole. Work control generates a list of work orders that needed to be called back to implement the stop work. The work orders

} for valves RH-MOV 23 and RH-MOV-34 sre not called back. /This partis /stop work order was kaued to protect the RH t systems abHfty to remove decay heat.

At ins tbne the "B" RHR pump and the A' RHR heet exchanger wem known to be '

uneveneMe for service. The fanure to ceMback these work orders was

}: b ,: , .; te.]

i Sectember 4.1996 0700 The Electrical Supervisor issues the work orders for valves RH MOV 34 and RH-

, MOV-23 making them available for work. [Since the work orders were notpuMed back, the Bectrical Supervisor assumed that work couldproceed on the valves.]

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1000 The lead electrician discusses the work with Administrative US. The  !

l' Administrative US discusses the work with other control room operators. The Shift I Manager was not informed. (The Administratin US and controlroom operators . l i

shouWhave noted that work on these veins was not consistent with the stop work i order and stopped tNs work actMty.] j i l i 1050 The control room operator cracks open valves RH-MOV-23 and RH MOV 34. The )

! lead electrician removes power to the valves at the breaker cabinet prior to entering containment to work on the valves. (The valves am used to isclate the operating l RHR system from the containment spray header. A closed downstream manual l valve (RH V-23Al, was to prennt the RHR system flow from being divertad to the \

i spray header. The team concluded that the operator should han been aware of the l condition of the RHR system and stop work order and should han not opened i these values.] '

l 1135 The Shift Manager notices pressurizer level decreasing and informs the US. The decrease in pressurizer level is confirmed by checking the cavity level (decreasing) and the containment sump (increasing). The lead electrician is contacted to get  !

i permission to reclose the valves. (The RCS was being dinrted through the cracked '

open valves (RH-MOVs 23 and -34), leaking through the closed manualisolation vain, andinto an open drain Mne off the spray header to the containment sump.)

1140 Electrical power is restored to the valves RH-MOV-23 and RH-MOV-34 and the

! valves are closed. Closing these valve stops the leakage and terminates this

, event. (The plant recorder charts indicits a ~ 1.5% loss of pressurizerlevel and a

~ 6 inch loss in cavity lentindication. De totalinventory diverted from the RCS to the containment sump is estimated to be ~200 to 300 gallons.]

i 1145 A NSO is dispatched to verify that valve RH V 23A is completely closed. The NSO

, manages to close the valve, by hand, another 118 of a turn. Later in the afternoon,

another attempt to close the valve further was made using a valve wrench, another

% turn closed was made. (The failure tv completely close this valve causedit to l inak by. The team concluded that poor rosterial condition, resulting in operators

! being unab6e to adequately close severalmanual valves andinsking manualisolation

valves, was a root cause for several of tne events reviewed.]

1241 The containment sump is pumped and ~380 gallons of water are removed.

1320 An ACR 96-0978 is written that documents the RCS inventory diversion event.

(The team concludes that, in addition to the failed work stoppage, the common manual RHR to containment sprey isolation valve, RH V 23A, leaking by its seat was a contributing cause of this ennt.]

i 1530 A complete stop work order is issued from the Unit Director to all employees with a written explanation.

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27 FAILURE OF DECAY HEAT REMOVAL COMPONENTS July 29,1996 1010 An inservice test is conducted on the RHR pumps.

(The "B"RHRpump was placedinto the " Alert" range due to Ngh vibration. An

,;..,:.- :2., evaluation of the vibration date was performed and the test frequency was changed ham quarterly to montNy.]

Auaust 19,1996 0730 The "B" RHR pump was started and run in parallel with the "A" RHR pump. /The pump was run to cooldown the RCSin preparadon ihr the LPSIsurveWance test.

During the test the RHR pumps em secumd. It appears that the final damage to the "B" RHR pump occurred during tNs pump operation.]

0824 The "B" RHR pump is secured. (TNsis theinst time the "B"RHRpump was successfuWy operated before it was found seked on September 1,1896. The aperators stated that the pump ddnot dspiny any abnormalindications during tNs aperation. Itis presumed that the pump was seked andincapable of operating from tNs time and remained unevaHable untM the pump was ident/ Red as inHed on September 1,1996.]

Auaust 31,1996 0900 A NSO, while conducting routine rounds in the RHR pump area, identifies a small amount of water leakage from the "A" RHR heat exchanger inlet valve (RH V-791 A) l body.

l 1100 ACR 96-0968 is written to document the leakage from valve RH V-791 A.

1352 The pin hole leak on the "A" RHR heat exchanger inlet valve RH-V-791 A is reported under 10 CFR 50.72(b)(2)(i). ITS Lhn/r/np Condition for Operation 3.4.1.4.2 l entered because ofinoperable RHR loop.] \

1440 Operators reduced RHR flow to 2190 gallons per minute (gpm) in preparation of

isolating valve RH-V 791 A. (The RHR NOp faHed to provide maximum flow rates

with a single heat exchanger.]

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' 1 1532 Valves RH-V-791 A and RH-V-794A are closed to isolate the "A" RHR heat l exchanger. (The "A " and "B" RHR pumps feed a common header, so any comNnation of pump / heat exchanger may be used to remove decay heat.]

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i i 28 September 1.1996

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1419 The decision is made to start the "B" RHR pump, to allow the "B" RHR pump to j operata with the "B" RHR heat exchanger, j The "B" RHR pump fails to operate on the first start attempt. A second st' art is

! attempted within a few minutes of the first and the second attempt also fails.

j During both attempted starts motor ampere indication remained off scale high

(>400 amps) for no more than 5 seconds. An attempt to manually rotate the pump j shaft identifies that the shaft would not turn. The "B" RHR pump is declared inoperabie. [Tids mduces the conning capacity of RHR to one pump ("A') and one heat exchanger ("B").]

l 1448 ACR 96-0964 is written to document the failure of the "B" RHR pump.

! 1752 The RCS loops #3 and #4 cold leg stop valves are opened. /This was to pmWde a l backe coonng methodin accontence with procedures.]

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2124 RCS Loops #1 and #2 are unisoleted (cold leg and hot leg isolation valves opened).

ITids step was one mouined forplacing the steem generators k service as backup to i the RHR system for wsselconning. The team noted the faBure to fonowpmcedwes resultedin the conection ofinaccurate bomn samples of the steem generator l hunntary. The steem generators am sampled to prennt dituting the water k the reactor.]

j September 2,1996 l Maintenance activities to repair the "B" RHR pump begin.

I l September 3,1996 j The spare RHR pump assembly is moved from the warehouse to the maintenance

shop. (The apare replacement pump had ccncentricity andperpendicularity i afe6clencies between the bearing and head casings. The pump was shpped out to l the under for repaks. The pump imponer requked additionalmachining and j belancing. The team determined that the uneveHabWity of quaHty spare parts j contributed to the siged6 cant delays the RHR pump replacement.

4 l The nndor specl6 cations of the RHRpump were not readHy avaHable. The team concAuded that the teck of vendor information on the pump resultedin senrel evoldeble aRelays duning the pump repair.]

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29 Seotember 4,1996 1330 The floor blocks are removed to access the RHR pump for repair. The pump is removed from the casing. A visual examination identifies some wear on the circumference of the pump wear rings. /The reem noted that the lack of an avaMahan techedcol evaluation for removing the RHR pump floor blocks resulted in a delay in the pump repairs and contributed to the length of thne required for the pump repair.]

Seotember 7.1996 The licensee issues TPC 96-570, Nuclear System Operator Primary Side Shutdown Logs, to require increased surveillance frequency (every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) of the "A" RHR pump oil levels, pit ambient temperature, and pit area status. /The team concluded that tMs actior: was appropriate but not thr,cly. The team also beHeved additional non-intrusive survemance measurements of the "A" RHRpump wouldhave been appropr!ste.]

Seotember 9,1996 Maintenance activities to machine the casing bore for the wear rings continued.

Other ongoing activities include, making a new wear ring, and machining the inside diameter of the impeller, i

l Scotember 11-13,1996 i

9/11-13 The pump maintenance is delayed while waiting for vent valves for casing j and bearing cooler line, wear ring, and other parts.

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I September 13.1996

An ultrasonic non destructive test (UT) is performed on the leaking RHR heat

exchanger inlet valve (RH V-791 A). The casting wall thickness did not indicate any substantial wallloss.

Seotember 14.1996 The repairs and installation of the "B" RHR pump are completed, i

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30 i 1945 A post maintenance test is performed on the "B" RHR pump. The pump failed to l paes the post maintenance test. The pump is stopped after approximately 10 seconds, when the indicated motor amperage is less than expected. /The technical staff stated that the cause for the pump test faRure may how been hadequate venting. The pump hadbeen notedk accordance with the RHR system normal opsrathy procedure. Howewr, the pump unts em notlocatedat the high points and thn tecludcolstaff beGewed thatpoor ventlocation was the cause of the hedequate venthg.]

September 15.1996 A procedure is written to allow the pump to operate for 2 minutes at the reduced motor amperage. This is to allow time for the pump to discharge any gas that may be present in the pump! The pump is instrumented with local flow and pressure indications. The pump is started and again indicates low motor amperage, no flow,  ;

and low discharge pressure for the first 90 seconds of operation. At 90 seconds, l

the pump appears to start pumping and normal flow, pressure and motor amperage readings are observed. The pump operated for a short time before the operator )

noticed low oil level in the thrust bearing oiler and the test was stopped. The post maintenance test is not successfully completed. /The team concluded that theisch of appropriate pump vent locations contributed to the alged6cance of these events.

Had the nitmgen bubbse gmwn to a point whom altragen gas entered the RHR pumps, then procedures require securing and venthg the pumps. The inck of adequate vent locations could have adversely knpacted the abWty to retum a gas bound RHR pump to service.]

September 16.1996 The "B" RHR pump post maintenance test is performed. Test data indicated a problem with the pump motor. A visual inspection of the motor identified a crack in the motor endbell. The post maintenance test results show a test failure.

The AIT completes onsite inspection activities. / Work on the "B"RHRpump continued for another 9 days after the team concluded the onsite inspection activities. The work perfonned kciuded the replacement of bearing oW seals.]

September 25.1996 2100 The "B" RHR pump is declared operable. The pump was declared inoperable on September 1,1996. /The team noted that the pump maintenance was conducted k a methodicaland conservative manner. However, several avoidable delays contributed to the extended repair duration.]

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1. Ooerations j 03 Operations Procederes and Documentation ,

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03.1 Reactor Coolant System Draindown l

j a. Inspection Scone On August 29,1996, the on shift operators conducted two RCS draining evolutions ,

i in preparation for reactor vessel head removal. The team reviewed the procedure j used to drain the RCS written by shift management in accordance with ACP 1.2-j 5.3, Evaluation of Activities / Evolutions Not Controlled by Procedure, and discussed

its performance with the cognizant operations Unit Supervisor (US) and the Reactor l Operator IRO). Refer to section 04.1 of this report for NRC findings related to

operator knowledge and performance for this activity.

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b. Observations and Findinas j

Shift management used the guidance of ACP 1.2 5.3 and developed a seven step '

t procedure to control the RCS draindown in preparation for reactor head removal.

j The procedure was developed by the US since no valve alignments were specified i in the normal operating procedure (NOP) 2.3-4, Shutdown from Hot Standby to Cold

, Shutdown, to drain the RCS. The required actions included verification of RHR l purification system operation, placing :avity level and local standpipe level indication in-service, stationing operat es at each indicator, adjustment of the cavity low level alarm to 268 inches, and a c sindown flow path through valve WD-V-210 l to the aerated drains tank (ADT).

l The team noted that the procedure fai ed to indicate a specific vessel reference level

! value for the draindown, to provide fo reconciliation of inventory drained against an j expected level indication change, and to require verification of reactor vessel vent i

header system operation prior to the F CS draindown. Furthermore, the procedure

provided no information to control systems necessary for maintaining RCS inventory l while at the refueling reference level. NOP 2.3-4 cautions prior to RCS draindown to not change any valve line-ups that . night adversely affect the RHR system or RCS enventory while at the refueling reference level.

1 l Technical Specification (TS) 6.8.2 requires that each procedure described in TS

' 6.8.1, including the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978, and changes thereto, shall be

reviewed by the Plant Operations Review Committee (PORC) and approved by the

' Vice-President - Haddam Neck prior to implementation. Appendix A of Regulatory Guide 1.33 specifically requires written procedures for draining the RCS. Contrary l to TS 6.8.2, ACP 1.2-5.3 allows etivities that require PORC review to be

performed at the discretion of the Shift Manager (SM). .

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32 ACP 1.2 5.3 states that "This procedure is not intended to supplant existing l procedures ... " and "... applies to activities that are currently outside established  !

procedures." However, team discussions with operations personnel confirmed that l NOP 2.3 4, step 6.4.16, provided the only existing guidance for draindown of the -

RCS.

l The team made the following observations regarding the licensee's use of and  :

performance of an ACP 1.2 5.3 procedure for RCS draindown:

e The licensee's failure to have the procedure reviewed by the PORC and approved by the Vice-President prior to implementation was in violation of TS 6.8.2.;

e The worksheet written instructions were in violation of ACP 1.2 5.3 since draindown of the RCS is addressed by an existing procedure - NOP 2.3 4; and,  !

e The worksheet written instructions were inadequate in not providing for reconciliation of inventory drained against an expected level indication change, verification of reactor vessel vent header system operation prior to RCS draindown, or control of systems necessary for maintaining RCS inventory while at the refueling reference level.

c. Conclusions The team concluded that the ACP 1.2-5.3 written procedure guidance was inadequate for control of a RCS draindown evolution. The licensee failed to ensure the procedure guidance was reviewed and approved in accordance with TS 6.8.2, and the operators failed to question the lack of this review. The team concluded that the failure to have a detailed RCS draindown procedure was a contributing cause for this event.

03.2 Reactor Coolant System Vent Header

a. Ineuection Scone The inspection scope was to review the existing procedural controls of NOP 2.9-6, Primary Vent Header Operation, and evaluate system operation in accordance with NOP 2.9-6 during a system walkdown by the team on September 3,1996.

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b. Observations /Findinas The team noted that no vacuum pump installation connections are avai.sle to i

support venting of the RCS loops, and the NOP 2.9-6 system alignment drawing did not match the field installation for the reactor vessel head connection. The team learned that the operation of RCS vent and the reactor head vent are not periodically verified, and only checked for moisture or loop seals if deviaGns occur between cavity level and pressurizer level indications. The above deliciencies were presented to the licensee during the inspection.

On September 14,1996, the licensee revised NOP 2.9-6. The major changes to the NOP included: the establishment of a loop seelin containment vent header moisture drain tank; specific system valve alignments and guidance for connection to the loops, pressurizer, and pressurizer rol of tank; independent verification requirements during the installation of the resctor head vent; installation of a vacuum compensator; and additional requirements for daily checks of the digital manometer, daily walkdown of the system hosing, twice weekly drainage of moisture from the vessel head vent, and a check for level in the moisture drain tank.

The provision for a vacuum pump instead of the air eductor was removed from the procedure. The revised procedure approved by the PORC also included a technical evaluation for the changes and addressed the expected air removal capability of the system. The team verified that the system improvements were installed on September 14,1996,

c. f.;onclusions ,

The RCS vent installation lacked adequa e procedural guidance and controls. No procedurally required system checks occurred to verify acceptable operation prior to September 14,1996. The team concluced that the failure to have a detailed

procedure for the installation and operat.on of the vent header system resulted in the system being incapable of adequatel removing all the nitrogen gas from the reactor vessel.

l 1 03.3 Fill and Vent of the Charoitio System l

j a. Insoection Scone l On August 28 and September 1,1996, the operators conducted a fill and vent i procedure for the charging suction header and the "B" charging pump after an inadvertent injection of nitrogen into the charging system. The team reviewed the fill and vent procedure written by shift management in accordance with ACP 1.2-5.3 and discussed its performance with the cognizant operations US.

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b. Observations and Findinas 4

l Shift management used the guidance of ACP 1.2 5.3 and developed a ten step l procedure to fill and vent the charging suction header and the "B" charging pump.

+

The team discussed the content of this procedure with the US and reviewed piping

! and instrument drawings. The licensee's IRT identified several questions regarding l the adequacy of this procedure that required additional evaluation.

The developed procedure to fill and vent the charging system was not in accordance with TS 6.8.2.

c. Conclusions j The licensee failed to ensure the procedure guidance was reviewed and approved in j accordance with TS 6.8.2, and the operators failed to question the lack of this a review.

03.4 Placina Residual Heat Removal Purification System in Service I l

i i s. Insoection Scone, 1

On August 22,1996, the cavity levelindicator experienced an unexpected drop 4

from approximately 325 inches to 273 inches. The cause of the event was the l f ailure of a non-licensed operator to preterly perform the RHR purification system i valve alignment to restore the system firm a temporary shutdown condition. The I

team reviewed Adverse Condition Repor'. (ACR) No. 96-0926 and the resulting i temporary procedure change (TPC), and :liscussed this review with the cognizant

. operations US. Refer to section 04.4 of this report for NRC findings related to I operator knowledge and performance for this activity.

{ b. Observations and Findinas i The purification system is designed to renove impurities from the RCS during plant  !

! shutdown and depressurization. On Aui;ust 22,1996, the system was placed in a )

temporary shutdown status in accordance with NOP 2.7-4, RHR Purification System Operation, to allow for testing of safety injection (SI) system check valves per surveillance procedure (SUR) 5.7-66. The RHR purification effluent discharges

through the high pressure safety injection (HPSI) loop stop valves and associated check valves. Attachment 4 of NOP 2.7-4 provides guidance for both temporary i system shutdown (steps 1.1 through 1.3) and subsequent restoration (steps 1.4 and 1.5). The system shutdown status checklist (step 1.3) isolates the purification system from the HPSI discharge line and opens the purification pump suction valve

, (PU V-261 A) from the RWST. This valve is reclosed in step 1.4, during performance of the purification valve alignment prior to opening the purification pump suction isolation valve, (RH-V-874A), from the RHR system in step 1.5 to j reestablish system flow.

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' 35 ACP 1.2-6.5, Station Procedures, advocates the use of personnel sign-off blocks for  !

actions that are likely to be overlooked and includes actions that "make alignments" l for consideration in the use of personnel sign-offs. NOP 2.7-4 did not require ,

personnel sign offs for completion of the valve alignments directed by action steps 1.3 and 1.4. ,

The licensee initiated an ACR and issued a TPC which added personnel sign-offs for i steps 1.3 and 1.4, and clarified the purpose of step 1.4 with regard to system restoration.

c. Conclusions The inspectors concluded that NOP 2.7-4 was inadequate although all procedural steps were found to be correct. The licensee failed to ensure the action steps required for system restoration were clearly identified and did not designate all  ;

system alignment steps for personnel sign-off as recommended by procedure ACP 1,2-6.5.

03.5 Residual Heat Removal System Ooeration

a. Inspection Scope l

The inspection scope was to evaluate the quality of operating procedures for the RHR system. In particular, the review evaluated the procedural controls and instructions for shifting RHR heat exchangers and pumps during a cold shutdown ,

condition. The inspection consisted of interviews with operators and system i engineering personnel.

b. Observations /Findinas l 1

The team learned that on August 31,1996, the licensee isolated the "A" RHR heat exchanger from service due to a valve body leak on the manual inlet valve RH-V-791 A. Later on September 1, the operators attempted to shift RHR pumps and identified the failure of the "B" RHR pump to start.

The procedural guidance in NOP 2.3-4 requires when RCS temperature is less than 120 degrees fahrenheit (*F) that RHR flow be reduced to 2000 2200 gallons per minute (gpm) and one of the two running RHR pumps be secured. NOP 2.9-1 initially establishes RHR flow for hot shutdown with two RHR pumps and two RHR heat exchangers operating.

On August 31,1996, procedural guidance did not exist for shifting the RHR pumps once aligned in accordance with NOP 2.3-4, or for isolating one of the two RHR heat exchangers. Additionally, NOP 2.9-1 did not specify any limits for maximum RHR flow rates through the heat exchangers. The team reviewed the Updated Final Safety Analysis Report (UFSAR) and concluded that no design limitations were documented for maximum RHR flow through the heat exchangers. Prior to the operators isolating the "A" RHR heat exchanger, the RHR system engineer specified

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maximum allowable RHR flow limits of 2200 gpm through any one heat exchanger (based on the design limits supplied by the vendor) and 3000 gpm per RHR pump to prevent pump runout. The operators isolated the "A" RHR heat exchanger fiom j service, using general knowledge of heat exchanger operation and isolation, by i shutting the RHR inlet valve and maintaining the RHR outlet valve open while ,

continuing to supply cooling water flow through the heat exchanger. l The team also learned that NOP 2.9-1 did not p' ovide specific instructions on 4

venting of the RHR pump or its associated suction and discharge lines prior to  ;

starting a pump. During the attempt to start the "B" RHR pump on September 1, the operators did not conduct any specific venting evolutions for the "B" RHR j pump. Operations Department instruction (ODl)-1, Conduct of Operations, provides

general guidance fnr venting pumps and piping high points whenever possible  ;

} following draindown of a system or component. No draindowns of the system or i

the "B" RHR pump occurred between securing the pump and the attempted restart.

i The RO attempted to start the "B" RHR pump and noted that the ampere indicatinn 1

in the control room maintained greater than 400 for approximately five seconds.

l The RO expected to see the ampere indication go off-scale at greater than 400 l amperes for a couple of seconds and then come to a value of approximately 180 i

amperes. The RO secured the pump, consulted with the SM, and then attempted to restart the motor after approximately one minute. The restart attempt had the same i indications of off scale amperes. The RO again secured the pump. The SM sent a j Nuclear System Operator (NSO) to "bar over" the pump shaft and to verify if any 4

protective relays came in locally at the breaker in the "B" switchgear room. The

, NSO reported that the shaft did not move when " barred over." No protective relays

were noted to have tripped at the breaker. ODI 1, section 5.3.8 provides guidance on resetting protective devices and states that if protective devices trip, an attempt l should be made to understand the cause prior to resetting the device. For the "B"

, RHR pump no protective devices tripped during the start of the pump. The team concluded that the RO actions to restart the "B" RHR pump were not precluded by 1

) ODI 1. Based on a subsequent motor current spectrum analysis, no observed I i damage to the motor was identified by the licensee. The analysis verified that no l rotor bar damage occurred during the short time that the motor experienced locked j rotor current.

j On September 13,1996, the licensee approved revision 32 to NOP 2.91 which

! made the following major changes: upper flow limit established for RHR flow I

through a heat exchanger; attachments added to vent the RHR pump prior to starting or shifting pumps; and steps added to secure the one pump when both are no longer required.

I c. Conclusions The team concluded that NOP 2.91 was inadequate in providing guidance for shifting RHR pumps, upper RHR flow limitations, and isolation of heat exchangers. l

<, The licensee revised NOP 2.91 to address these procedural deficiencies, l 4

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03.6 Boration Iniection Flowcath Alianment i

j a. Inspection Scone 1

On August 28,1996, during performance of an alignment check to verify an l operable TS required boron injection flow path, a NSO incorrectly opened a valve 4

that resulted in injecting water and nitrogen into the RCS. The causes of the event were an inadequate pre job brief (discussed in 04.5) and the failure of the NSO to

comply with all procedural precautions. The team reviewed the applicable

surveillance procedure, SUR 5.1-1598, Boron injection Path Valve Lineup and 4 Metering Pump Test (Shutdown Modes 5 and 6), and discussed its performance with the operations personnel.

l b. Observations and Findinos i

Procedure SUR 5.1-159B requires a verification that each valve in the TS required j boron injection flow path that is not locked,

::aled, or otherwise secured is in its j correct position. On August 28,1996, the existing &ction flowpath was from the l boric acid makeup tank (BAMT) via the metering pump to the RCS. This flowpath l was to be rendered temporarily inoperable during the performance of preventative 2

maintanance procedure (PMP) 9.1-31 on Emergency Generator (EG)-2B. EG 2B is

the emergency source of electrical power for the metering pump. The NSO was.

. directed to verify alignment of another operable injection flowpath from the BAMT via a boric acid pump and a charging oump to the RCS. However, the flowchart

] methodology is not a pre-determined valve line up checklist and requires the operator to select the available flow path (s) from among several possibilities by l applying knowledge of existing system and component status in answering l questions regarding operability and availability, e.g., "BA Filter in Service? Yes/No."

Procedure step 6.1.2 directs the operator to " VERIFY" that each component for the

' selected flowpath is in the required position but does not direct repositioning of any

component. If a valve is not found in the position specified on the valve lineup

! checklist, precaution 5.1.1 directs the operator to immediately notify the shift

supervisor and not to proceed until the situation is resolved, i

According to the NSO's description of the occurrence, he believed that the direction of step 6.1.2, to verify each flowpath component in its correct position, authorized realignment of components as necessary and that precaution 5.1.1 did not specifically apply once thu flowpath had been determined. As a result, the NSu

- repositioned valve BA V-355 and injected water and nitrogen from the CVCS into the RCS. The licensee subsequently issued TPC No.96-358 that added a note to flowchart 1 precluding valves BA-V 354 and BA-V 355 being open at the same time with valves CH-MOV-257 or CH-MOV-257B shut. Refer to 04.5 of this report for

[ NRC findings related to operator knowledge and performance.

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c. Conclusions Operations procedural guidance for verification of an operable TS required boron injection flow path was adequate. Adequate direction is placed in the SUR procedure to preclude component repositioning during flow path verification. i

However, due to the many system and component line-ups possible during an

outage, the procedure relies on the operator to recognize and understand system and component status for proper performance. The causes of the event were an i inadequate pre job brief and the failure of the NSO to comply with all procedural F precautions, i

I 03.7 Loss of Residual Heat Removal i

j_ a. Inspection Scope

The inspection team reviewed the abnormal operating procedure for a loss of the

residual heat removal system.

f a

j b. Observation /Fmdenas 1

l The licensee has established an abnormal operating procedure, AOP 3.212, Loss of j Residual Heat Removal. This procedure has initial conditions of the reactor in Mode

! 5 or 6, cold shutdown and refueling, respectively, its initial strategy involves l diagnosis of plant conditions and actions to restore the RHR system, including l- restoration of power, component isolation and pump suction venting.

{ The licensee developed a plant condition action matrix within the procedure that i directs the operator to perform actions that establish alternate means of decay heat removal, depending on the plant conditions and what is known about the RHR system. The procedure plant condition action matrix directs the operator to a specific decay heat removal recovery action. Had the RHR system failed during the l

period of interest covered by this inspection, the operations staff would have attempted recovery by isolating and pressurizing the RCS and establishing a heat

{ sink with a steam generator with natural circulation or forced cooling with a RCP.

Had this primary method failed, the procedure directs the operators to alternate l methods. Feed and bleed cooling of the reactor core would be attempted with a charging pump. The RCS would be pressurized and coolant discharged through a l low temperature overpressure protection (LTOP) relief valve. Then, forced cooling would be attempted with a LPSI pump and the RHR heat exchangers.

In order to address various contingencies, AOP 3.2-12 addresses contingencies of

filling RCS loops with a charging pump or by gravity feed, unisolating an RCS loop, natural circulation or reflux cooling and starting a RCP.

i

!. The licensee has also developed abnormal operating procedures that address loss l the RHR support systems: AOP 3.2-10, Loss of Component Cooling Water; AOP i

3.2-15, Loss of a Vital Bus; AOP 3.219, Loss of Service Water; and, AOP 3.2-25,

. Low Voltage on Emergency Busses.

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! c. Conclusions The tasm concluded that the licensee had developed reasonable plans and q

j procedures that should be able to cope with a loss of residual heat removal event.

l 04 Operator Knowledge and Performance i 04.1 Reactor Coolant Svstem Draindown

a. Insoection Scone

The inspection scope was to review operator actions during a pre-planned draining l evolution of the RCS on August 29,1996.  !

! b. Observations /Findinas l 1 On August 29,1996, during two separate occasions the control room operators

! aligned the RHR purification system and letdown the RCS to the ADT through valve

{ WD-V-210.

l At approximately 1609 after a total of approximately two hours of draining the

!- RCS, the evolution was stopped due to an outage planning decision that postponed j the reactor disassembly until September 3,1996. The RCS draining evolution i j supported removal of the reactor vess il head.-

! Discussions with the US and the RO is dicated that the total inventory change was i estimated at 3,000 gallons. However, the second draindown evolution was

! stopped at 280 inches cavity level aft sr approximately 4,000 gallons were drained I to the ADT. (The licensee later calculated 4515 gallons as the total inventory

change based on run time of the ADT pump.) The operators attributed the i

! difference between the estimated draiixiown volume of 3,000 gallons and their ADT l j draindown of 4,000 gallons to in-leaki ge from the RCS loops. The operators j manually closed in on RCS isolation va lves for two loops and isolated the loop

! overpressure protection valves for all our loops. After this action, operators noted i that cavity level continued to increase with no make-up but the rate of increase j slowed after the valves were verified closed and the loop overpressure protection 3 valves were closed. The next shift added approximately 1,000 gallons to the RCS and returned cavity level to its approximate initial value.

i 1- Based upon team interviews with the shift operators and operations management,

{. the magnitude'of the RCS inventory mismatch was not identified in an ACR, nor j was it explained to operations management. Nor did the operators solicit E&TS to

- assist in resolving this discrepancy. The team asked the licensee for the history of loop stop valve leakage. The licensee has experienced leakage of the loop stop valves during past outages, however quantification of that leakage has not occurred. Based on the inventory imbalance, the team concluded that leakage of this magnitude could not have occurred through the closed loop stop isolation valves. ,

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40 Further, the team learned that isolation of the loop overpressure check valves by the operators resulted in no overpressure protection for the RCS loops. The team evaluated the temperature changes in the loops, containment temperature, and the RHR suction temperature between August 29 at 1427 until August 30,1996 at 0416, when the loop overpressure protection was isolated. The maximum recorded rise in loops temperatures was 3*F and the containment temperature decreased 3*F based upon the startup of the second containment air recirculation fan. The TS limit based upon heat up limitation curves of TS Figure 3.4-4 at the RCS temperature of approximately 80*F was approximately 480 psig. Licensee evaluation of the loop temperature variations indicated that this limit was not exceeded.

The team learned that the normal operating procedures to isolate an RCS loop did not requee that the overpressure protection check valves be left in service, and the locked VAe checklist for the isolation of the overpressure check valves required the valves to be opened in Operational Modes 1 through 4, but not cold shutdown.

c. Conclusions Operators failed to inform management or engineering / technical support of the significance of the inventory balance given the insupportable magnitude of leakage from the loop stop isolation valves. The operators isolation of the loop overpressure protection valves was inappropriate in that it removed the only means to relieve an overpressure condition within the 'solated loops. The normal operating procedures did not preserve the pressure protection of an isolated loop. The operators failed to display a questioning attitude in not investigating the cause of increasing RCS level indication with no makeup.

1 04.2 Return of the Reactor Coolant Looos to Service il

a. Inspection Scone On August 31,1996, a smallleak was identified on the "A" RHR heat exchanger inlet valve, RH-V-791 A. The heat exchanger was isolated and the "A" loop of RHR was declared inoperable. On September 1, the operators performed NOP 2.4-7, Return of a Loop to Service with the Plant Shutdown, and unisolated the RCS loops to increase the predicted time to boiling in the RCS should the remaining RHR pump be lost. The team reviewed NOP 2.4-7 and riiscussed its performance with the cognizant operations US, SM and the Chemistry Department.

b. Observations and Findinas Procedure NOP 2.4-7 provides the detailed instructions necessary to return a RCS loop to service after it has been isolated or idled and requires that prior to opening

, the loop stop isolation valves, the isolated or idled loop shall be determined to have a boron concentration greater than or equal to the boron concentration required to j- meet the shutdown margin requirements of TS 3.1.1.4 and to have a cold leg

, temperature within 20 'F of the operating RHR loop. Prior to sampling the isolated

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j -41 or idled loops for boron concentration and determining loop temperature, steps ,

, 6.2.2 and 6.4.3 of NOP 2.4-7 direct that the associated RCP be started in  ;

accordance with NOP 2.4-2, Reactor Coolant Pump Operation.

l The US for the 0600 - 1800 shift on September 1 indicated that the procedure

steps required for operation and start of a RCP in accordance with steps 6.2.2 and ,

, 6.4.3 of NOP 2.4-7 were not performed but marked as "N/A" since the loops were  !

j being returned to service for availability and decay heat removal concerns only, and j_ the RCPs and plant were to remain shutdown. The US further indicated that the j verification of adequate loop boron concentration was based on the sample results '

j provided by the Chemistry Department where loop #1 boron was 2874 parts per  :

million (ppm), loop #2 boron was 3111 ppm, loop # 3 boron was 1790 ppm, and 3

} loop #4 boron was 2235 ppm taken prior to opening the loop stop isolation valves

{ and the last known loop boron concentrations of approximately 1400 ppm taken  ;

i prior to loop isolation 'on August 12,1996. The SM for the 1800-0600 shift on i September 1 and 2 indicated that isolated loop temperatures were verified by

{ contact thermometer readings on the secondary side steam generator hand hole j- covers and were within 20 'F of RHR temperature. Management review of i j completed procedure NOP 2.4 7 was completed on September 2,1996, by the j Operations Manager.

i The team discussed the loop boron sample results with the Chemistry Department and reviewed the relevant records to determine whether the boron samples were oroperly obtained and if the results were: credible. The ::hemists indicated that the l

soop boron sample results were not crecible since the samples could not be obtained

in accordance with approved proceduren, i.e. the RCPs were not started and the l loops were at atmospheric pressure restiting in the samples not being taken at the i

sample sink but being drawn locally in t is containment off the loop drain header.

The chemists further indicated the opers. tors were advised that the results would i not be valid indicators of loop boron concentration. When the boron samples for i loops 1 and 2 indicated boron concentra tions greater than the RHR system, the i operators requested another sample anc obtained similar results. The operators

failed to follow up the boron sample dis:repancy with chemistry prior to unisolating j the loops. The team reviewed the samt el records and determined the approximate loop boron concentration prior to isolation was 1430 ppm boror. and that no evolutions with the potential for dilution of loop boron concentration had occurred j after loop isolation on August 12,1996. The team verified in surveillance i procedure (SUR) 5.3-19, Boration Requirements for Reactor Shutdown, and in the i fuel cycle report that the required shutdown margin boron concentration in this

condition was approximately 850 ppm.

The Procedure Compliance Policy of ACP 1.2 6.5 states that "... if the procedure i appears to be inadequate for the intended task ... then task will be stopped" and i

limits the use of "N/A" for cases where a step or steps of a procedure are clearly not applicable. The lack of representative boron samples indicated a f ailure to I

! comply with the intent of the procedure. Furthermore, the licensee identified on j September 13,1996, that the boron samples may not have been taken within 30 l minutes of unisolating a loop as required by TS 4.1.11.2.

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) a 42 I The team made the following observations regarding the licensee's performance of NOP 2.4-7: '

o The operators use of "N/A" for action steps dealing with operation of the RCPs was improper and resulted in unrepresentative boron samples and ,

isolated loop temperatures, and a failure to comply with TSs 4.4.1.11.2 and i 4.4.1.11.1. The as-performed procedure was inadequate for these two ,

tasks and should have been stopped until appropriate procedure changes  ;

were made or the RCP could be started.

o The safety consequence of the operators failure to obtain accurate loop boron and temperature results was minimal based on the loops' previous ,

boron concentrations and ambient temperature conditions.  :

e The management review of the completed procedure was inadequate since it also failed to identify the procedural non-compliance.

c. Conclusions The team concluded that NOP 2.4-7 was improperly performed. The licensee failed to ensure that isolated loop boron concentrations and loop temperatures were correctly determined prior to opening the loop stop isolation valves. Although not directly contributing to the occurrence, the management review of the completed procedure was inadequate since it also f ailed to identify the procedural non- ,

compliance. '

04.3 Operator implementation of Abnormal Oneratino Procedure for Reactor Coolant System Leakane Indication

a. Inspection Scone The inspection scope was to evaluate or erator response on September 1,1996, to a unexpected reduction in RCS inventory,

b. Observations /Findinas On September 1,1996, at approximately 0900, the NSO isolated nitrogen gas to the VCT while conducting an investigation into increased nitrogen use noted over the past four days. When nitrogen was isolated to the VCT, the RO noted a decrease in cavity level and pressurizer level. Based on interviews with the RO, the RCS leak rate was estimated at approximately 20 gpm. The SM directed the NSO to restore nitrogen slowly to the VCT and increase VCT nitrogen pressure at 5 psig increments from approximately 20 psig to 30 psig.

The US executed the action steps of AOP 3.2-31 A, Reactor Coolant I System / Refueling Cavity Leak (Mode 5 and 6), to identify and isolate the source (s) of the leak. The RO initiated makeup to the RCS using a suction path from the RWST through the "A" RHR pump. AOP 3.2 31 A also requires a review of

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1 43 Emergency Plan implementing Procedure (EPIP) 1.5-1.A, Emergency Assessment, for determination of potential reportability, incident classification and initiation of ,

emergency actions. Based on interviews, the SM stated that he consulted the EPlP l and concluded that he did not meet the threshold provisions for an emergency j j classification declaration or the immediate notification criteria of 10 CFR 50.72. On )

September 11,1996, the licensee reported the event under 10 CFR 50.72 1 (b)(2)(iii)(B). .

The operators' leak investigation for areas both outside and inside the containment was completed by approximately 1015 and 1057, respectively. No potential leak sources were identified. However, based on the indicated pressurizer level decrease

and the patential RCS inventory reduction, the SM directed the RO to control RCS i inventory and by maintaining indicated pressurizer level between 5% to 15%. The licensee calculated that approximately 5,000 gallons of borated water were added  ;

j to the RCS based on the change in RWST level, t

l l During the RCS leak investigation, the operators had suspected nitrogen gas

intrusion into the vessel as the cause for the cavity and pressurizer level changes.

In parallel, with implementation of AOP 3.2-31 A, the operators cycled and verified closed the two valves connecting the VCT gas space to the "B" charging pump

suction header. Valves BA-V 354 and BA-V-355 were verified closed, and the nitrogen supply to the VCT was isolated. The above valves were tagged out at approximately 1353. This action isolated the nitrogen intrusion through the charging system into the RCS. At approximately 1255, on September 1,1996, RCS inventory was stabilized and indicated pressurizer level returned to its  !

approximate initial value of 10%.

! The team reviewed the SM's logs and concluded that this record failed to  !

adequately describe the sequence of events recounted above in that all of the major i activities were summarized under a single log entry for 0900 on September 1,1996. The operations crew initiated ACR 96-0966 to document this event and the actions taken. Based upon interviews, operations requested no l additional support to evaluate the worst case inventory reduction within the reactor l

, vessel given en apparent significant nitrogen gas displacement of reactor coolant.  !

l Three days later, the licensee performed a preliminary volumetric balance on l September 4,1996, and concluded that worst case inventory reduction within the )

vessel was 39 inches below the reactor vessel flange.

c. Conclusions f c

1 The team noted acceptable diagnosis and AOP implementation by the operators to l

q identify potential RCS leakage source (s). After further review by the licensee staff,  !

the SM's initial reportability decision of September 1 was reconsidered and the event was later reported by the licensee on September 11,~1996. Operations requested no additional technical support to evaluate the worst case inventory reduction within the reactor vessel given an apparent significant nitrogen gas I displacement of reactor coolant.

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04.4 Placino Residual Hect Removal Purification System in Service d

s. Insoection Scone I

The inspection scope evaluated operator perf

rmance during a system alignment to restore RHR purification on August 22,1996.

I b. Observations /Findinas On August 22,1996 a non-licensed NSO failed to adhere to NOP 2.7-4.

Specifically, the operator failed to perform step 1.4 of NOP 2.7-4 which restores the system alignment from a temporary shutdown of the purification sub-system of RHR. Failure to perform the alignment resulted in maintaining valve PU-V 261 A l open. Valve PU-V-261 A is the suction valve from the RWST to the purification

] pump. When the operator performed step 1.5 of NOP 2.7-4, it required opening l valve RH-V-874A which is the outlet valve from the RHR heat exchangers to the suction of the purification pump. With this alignment, RHR flow, instead of being directed to the purification pump, was diverted to the RWST.

! When directed to place the RHR purification back in service, the NSO assumed step 1.4 had been completed and restarted the temporary shutdown procedur=, at step 1.5 " COMMENCE purification of the RHR System"- and opened valve RH-V-874A 1

which diverted RHR flow to the RWST. On receipt of a cavity low level alarm in the control room, the RO, after consulting with the NSO, directed the NSO to shut valve j RH V-874A isolating the flowpath from the RHR system to the RWST. The consequersce of skipping step 1.4 was the inadvertent diversion of approximately l 500 gallons of RCS water from the RHR system to the RWST over a period of approximately five minutes.

Based upon interviews with the shift operators, the team noted that this was the j first time the NSO had performed the procedure for restoration of RHR purification.

. The on shift US indicated his expectation was that if the NSO had questions about the NOP implementation that they should be discussed prior to the evolution. The

licensee's IRT found that no pre-job brief occurred prior to the restoration of RHR J

purification.

} c. Concluti2D 1 .

. During the restoration of the RHR purification system alignment on August 22, the j NSO failed to adhere to NOP 2.7-4. The licensee failed to conduct an adequate pre-job brief for the cperator who was performing NOP 2.7-4 for the first time. The result of this lack of procedure adherence was the diversion of approximately 500 gallons of RCS water to the RWST.

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j 04.5 Imoroner Alianment of Boration Flowoath i 1

a. Inspection Scope l The inspection scope was to evaluate the events and details associated with 1

changing the boration flowpath on August 28,1996.

l

! b. Observations /Fmdinas I f '

J On August 28, a NSO misaligned valves in the boric acid system which connected

the VCT gas space to the suction of the charging system, through valves BA V 355

and BA-V 354 and subsequently into the RCS. At the time, the VCT was 1

pressurized with nitrogen gas to approximately 30 psig consistent with NOP 2.3-4 l which allowed for a makeup flowpath to the RCS using VCT pressure. The RCS  ;

i was approximately atmospheric pressure and vented to the reactor vessel vent '

1 header system. The boric acid system valve misalignment resulted in CVCS

! nitroger' and water being injected into the RC3. The pressurizer level increased j approximately 2% and VCT pressure decreased approximately 2 psig.

Verification of another boration injection flowpath was required in preparation for an i e

expected emergency diesel generator surveillance later in the day. The emergency diesel generator is considered inoperable during its surveillance test making the j existing flowpath via the metering pump to the RCS also inoperable. l 3

j Based on interviews with the US, RO, work control supervisor, and the NSO on i shift at the time of the event, the team concluded that the NSO received a less than

' adequate pre-job briefing. It was apparent that the NSO discussed various aspects of the task separately with all of aforementioned individuals but that an integrated pre job brief, with clear expectations for task performance, was not conducted.

Additionally, based upon interviews with the shift operators, the team noted that

this was the first time the NSO had performed a boron injection flowpath >

verification using procedure SUR 5.1-1598.

As a result of an inadequate pre job brief, the NSO chose a flowpath incompatible l with the existing VCT status and valve line up - approximately 30 psig nitrogen in 4

the VCT and valves CH MOV-257/2578 blue tagged shut - and incorrectly opened  ;

BA V-355. Opening this valve allowed the nitrogen pressure in the VCT to inject CVCS system water and nitrogen into the RCS. The RO's initial response to the boration flowpath error was to readjust the high alarm setpoint for cavity level and ,

l increase it by approximately 5 inches higher than actual level. On receipt of the '

} second high cavity alarm, the RO directed the NSO to close mis-positioned valve

.' BA V-355 and stop performance of SUR 5.1-1598. The team learned that no specific procedural expectation exists for controlling the high level alarm setpoint for

, cavity level. (Annunciator Response Procedure (ANN) 4.24-1, Cavity High Level,

indicates the setpoint is determined by the operator. NOP 2.3-4 and NOP 2.4-10, Reactor Coolant System Mid-Loop Operation, require the adjustment of the low level i

i 1

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j' 46

{

2 alarm setpoint to 255.5 inches and 208 inches, respectively, prior to conducting RCS draindown evolutions.) The licensee estimates that the amount of inventory added to the RCS was 500 gallons, in addition to nitrogen. The licensee prepared ACR No. 96-0946 to document the event and to identify potential corrective i actions.

4 l c. Conclusions During the evolution to shift boration flow paths, nitrogen gas entered into the RCS.

The evolution to shift boration flow paths was inadequately performed, in that, the

]

t NSO, who lacked experience and familiarity with the job task, was given an I

inadequate pre-job brief by shift management.  !

I 04.6 Failure to Control Maintenance Testino on Containment Sorav Valves

a. Insoection Scope

{

3 The inspection scope was to evaluate operations performance during an inadvertent i RCS draining evolution during maintenance testing of RHR motor-operated valves.

i

b. Observations /Findmos

}

l On September 4,1996, between 200 to 300 gallons of reactor coolant was j discharged to the containment sump from the RHR system. The leakage path was j through a pair of motor-operated valves (RH MOV-23 and RH MOV-34) and a

common isolation valve RH-V-23A to the containment spray automatic drain valve

! (RH-V-888). At the time, the licensee had a stop work order in effect on RHR and systems affecting RCS inventory.

l A preventative maintenance activity was being performed to disassemble the two l motor-operated valves (RH MOV-23 and RH MOV 34). In support of this i disassembly the valves were moved off their closed seats. The RHR flow was j diverted through the valves, and through a manual valve that was thought to have been closed (RH-V-23A) and then subsequently through valve (RH-V-888) to the j containment sump.

l l

The SM noted a small decrease in pressurizer and cavity levelindication. An operator was dispatched to containment to locate the potential leakage.

Approximately five minutes after the event started, operators reenergized the two

, motor-operated valves, closed the valves, and terminated the loss of RCS inventory.

l The leakage resulted in approximately 1.5% decrease in pressurizer level and i

approximately 5 inch decrease in cavity level. The operator in containment, closed i RH-V 23A another 1/8 of a '. urn.

l l- The maintenance should rot have occurred as a result of the stop work order issued on September 2,1996. This particular maintenance activity was not pulled by outage management, and the electrical maintenance department assumed that since

it was acceptable to work, they proceeded on September 4,1996. Additionally, 1

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47 the team learned that the SM was not informed about the maintenance activity.

The control room operators were aware of the activity, but felt that it was low risk a since isolation valve RH V 23A was closed. They assumed that the work was approved since the work control center and tagging office had cigared the work -

order.

l i c. Conclusions The operations crew did not thoroughly question the decision to proceed with the 2 work given a limited stop order work on systems involving RHR and other primary systems affecting RCS inventory. The team determined that the operator response  ;

to the reduction in RCS inventory was appropriate.

i 04.7 Operator Response To Excessive Liouid Nitronen Use ,

1

a. insoection Scone l The inspection focused on the operator sensitivity to liquid nitrogen use during an i outage.

! b. Observations /Findinos 1 Based upon the team's interviews with a variety of operators, it was concluded that l

their concern for high levels of liquid ni

rogen usage dering the outage was only a 4

function of the need to resupply the ta *s. In general, the operators were aware of l past large demands.for nitrogen to sup) ort outage activities such as a steam generator sparge or draindown. The N50's logs require recording nitrogen tank

level every eight hours. At 20 inches 11 the tank it recommends actions be taken to refill the tank.

1 During the period of August 29 until September 1,1996, the use of nitrogen by the j facility had dramatically increased frorr the trend prior to August 29,1996. Prior

demand for liquid nitrogen resulted in a n approximate three inch tank level decrease over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Between August 29 th ough September 1,1996, the liquid nitrogen tank level decreased on average 15 inc has per day. One use of nitrogen during this time frame was the draindown and refill of the No. 2 steam generator.

4 On September 1,1996, the shift NSO raised to the US the significant increasing trend in nitrogen usage since the tanks had just been filled on August 29,1996.

Based upon the NSO's concern, the US coordinated a systematic search for excessive nitrogen use by plant equipment. The troubleshooting process involved the US using the P&lD for the nitrogen system, and coordinating with the NSO to isolate and unisolate various loads. Plant operators failed to solicit help from E&TS to locate the source of the nitrogen par leakage. ,

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At approximately 9:00 a.m., the NSO isolated the nitrogen supply to the VCT. The i NSO noted that primary auxiliary building nitrogen pressure increased from 40 psig to 63 psig,' and VCT pressure decreased from 29 psig to 18.5 psig. The control

room also noted a corresponding decrease in cavity and pressurizer leval. The control room personnel felt that they had a RCS leak and entered AOP 3.2 31 A. On entry into AOP 3.2 31 A, the SM told the NSO to unisolate the nitrogen supply to the VCT. Based upon team interviews, the SM felt it was prudent to place the system back in its normal configuration, and perform actions of AOP 3.2 31 A.

On September 1,1996, operations management developed ODI 190 to, in part,

! maintain an inventory of liquid nitrogen on-site and log its usage for monitoring of abnormal trends. On September 11,1996, the team confirmed implementation of 4

the instruction.

i

c. Conclusions

in general, operators were knowledgeable of historically known large liquid nitrogen i loads during refueling outages, however they lacked a timely questioning attitude for the unexpected nitrogen usage between August 29 until September 1. The team concluded that the operations and E&TS staff were not timely in investigating i the source of the high nitrogen gas usage.

04.8 Event Notifications f a. ((1soection Scone j The team reviewed the event notifications, made by plant operators to the NRC, tc l j verify that the notifications were completed in accordance with NRC requirements. l i

! b. Observation /Findinos l Prior to the arrival on site of the NRC inspection team the licensee had made one report in accordance with the requirements of 10 CFR 50.72 concerning the through-wall leak in an RHR system valve. There were no events classified under j their emergency plan. l The licensee reported discovery of a through wallleak in the "A" RHR heat l exchanger inlet isolation valve in accordance with 10 CFR 50.72(b)(2)(i) at 13:52  !

on August 31 (Event No. 30945). However, the licensee had not reported the l' potential loss of RHR caused by the accumulation of nitrogen in the reactor vessel before this concern was identified by the team. The licensee reported this event in accordance with 10 CFR 50.72(b)(2)(iii)(B) at 15:04 on September 11 (Event No.

30992). .

l The inspection team reviewed the classification of events that were the subject of l this inspection in regard to the station emergency plan. EPIP 1.5-1, Emergency l Assessment Using EAL Tables, addresses this classification procedure. None of the events required emergency classification. j l

2 = .

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! 49 f c. Conclusions On September 1,1996, the licensee failed to report the event that Eggld have

! prevented the fulfillment of the safety systems that era needed to remove residual ,

a heat within four hours as required by 10 CFR 50.72(b)(2)(iii)(B). The team

] concluded that the accumulation of nitrogen gas in the reactor vessel gegld have prevented the RHR system from functioning. The proper notification was made by I the licensee to the NRC on September 11,1996.

05 Operator Training and Quellfications 05.1 Level Indication and Vent Header System Trainina Durina Mode 5 Shutdown i Conditions i

j a. Inspection Scone l The team reviewed the licensed and non-licensed operator training material j regarding operation of RCS levelinstrumentation and the reactor vessel vent header 1 systems during cold shutdown conditions, and interviewed several operators and a l training department representative with respect to their content.

I b. Observetions and Findinas l The team reviewed the Refueling Malfurictions lesson p'an (CY-OP-LO AOP-L11),

i the Mid Loop Operations lesson p an (CN OP-LO-NOP-L10), and the Fuel Handling

! lesson plan (CY-OP-LO-PRISYS-LOO 900) and discussed their content with a i licensee training department representat ve. The training representative indicated

,. these lessons are used in the initial and :ontinuing licensed operator training

, programs to cover RCS level instrumentttion operation and industry events. The

! lessons provide a de' s cription of the RCL level indications, controls, and alarrn available during shutdown and refueling conditions. The lessons mandate review

for two events - reactor seal cavity failu e event (NRC IN 84 93) and undetected j accumulation of gas in reactor coolant s/ stem (NRC IN 94-36). However, the

! lessons provide no detailed information >n RCS levelinstrumentation operation and its response for mode 5 operations or the effects that reactor vessel vent header operation or other system changes can have on levelindication. The term 3

confirmed this observation during interviews with several operators who indicated i that existing training material does not present detailed information regarding RCS

. levelinstrumentation or reactor vessel vent header operation for the irregular conditions existing with the plant in cold shutdown.

The team reviewed the licensee's continuing training schedule for the previous and current periods. The training representative indicated that presentation of the information discussed above was planned for August and September 1996 during i Licensed Operator Requalification Training (LORT) Cycle 96 5 t' Jt was postponed until the week of September 30 due to the early plant shutdown. He indicated that this information was last presented to licensed operators during LORT Cycle 94 6 in October and November 1994. In lieu of LORT Cycle 96-5 training, originally 1

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50 scheduled for one week, the licensee substituted one day of training prior to the

outage on selected refueling topics that focused on various refueling evolutions, loss of RHR and AC events, and containment closure TS requirements. This training j did not discuss RCS level instrumentation or reactor vessel vent header operation.

i j Finally, the training representative indicated the current training information provided i

on the reactor vessel vent header system does not discuss system operation and is 4

limited to a description of the system's connection points to the RCS (lesson CY-

OP-NLO-PRISYS-LOO 210). Updated training information for this system is planned

for development and presentation in LORY Cycle 6 beginning the week of i September 30.

l i

l c. Conclusions I

The team concluded that the quality of existing training information for operation of l RCS level indication and the vent header system during cold shutdown conditions j was weak and failed to provide any cubstantial information on their operation during j the irregular conditions existing with the plant in cold shutdown. The licensee failed

to provide the scheduled detailed training on outage and refueling operations due to

the early plant shutdown. The team determined that the failure to conduct this j training was a contributing cause for this event.

I i

l ll. Main'* nanst l M1 Conduct of Maintenance

! M 1.1 Timeliness of RHR Pumo Reolacement i

a. Insoection Scooe

! The team reviewed the maintenance activities to repair the "B" RHR pump.

1 1

! b. Observations and Findinas l On September 2,1996, the maintenance mechanics began a disassembly and i

inspection of the spare RHR pump. The inspection was performed to verify that the spare pump was in compliance with the pump manufacturers specifications. By September 4,19h the maintenance mechanics had identified that the spare pump had several out-of tolerance dimensional readings that would require corrective

! action by the pump vendor. The specific out of-tolerance readings were the j- concentricity and perpendicularity between the bearing and head casings. The head i

casing and the pump impeller also required additional machining and balancing. The '

pump casing was returned to the pumo vendor for machining on  !

September 8,1996, and the impeller was returned the following day. The spare pump, casing wear rings, and new vent valves were not available for installation

until September 13,1996.

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! The team noted that the lack of vendor information on the pump resulted in delays

, during the pump inspection activities. The licensee had to request the i specifications from the vendor. Certain pump drawings were not received from the l vendor until September 7,1996.

On September 15,1996, the post maintenance test of the "B" RHR pump failed due to inadequate pump venting and vent locations. The second post maintenance test also failed on September 16,1996, when operators observed low oil level in the  ;

bearing housing and the pump was secured. On September 16,1996, the maintenance staff identified two through-wall cracks on the motor upper end bell.

This caused further delay while an analysis of the motor was performed. The post maintenance testing of the pump resumed on September 21,1996, at which time oilleakage was noted near the pump thrust bearing. The oilleakage was caused by the slip of an oil seal batfle that had an improper interference fit. The first attempt to repair the oil seal baffle was unsuccessful and additional repairs were required.

The RHR pump was returned to operable status on September 25,1996,25 days after the pump was identified as having failed.

c. Conclusions The team determined that repair activities for the RHR pump were generally methodical and conservative. However, several avoidable delays resulted in the untimely restoration of the pump to service.

M1.2 Removal and Reolacement of the "B" RHR oumo

a. Inspection Scone The team observed the removal and replacement of PAB access floor blocks and the "B" RHR pump repair to assess the licensees maintenance practices.

b. Observations and Findinas On September 4,1996, the PAB floor blocks were removed. The team noted that the pre-job briefing for the removal of the floor blocks was thorough. The floor blocks were removed in accordance with WCM 2.2 7, PAB/ Pipe Trench Floor Block j

Lift Procedure. The team noted that step 1.3, which required the floor blocks to be ,

sealed with a silicone rubber adhesive upon reinstallation, was not completed. The I system engineer was not aware of this step and had not taken this into account during the development of the engineering evaluation for the removal of the floor blocks. The licensee posted a flood / fire watch until further engineering evaluations were performed.

i i

Corrective Maintenance Procedure (CMP) 8.5 99, Residual Heat Removal Pump l Maintenance, step 6.2.11, requires the installation of a foreign material exclusion (FME) cover over the "B" RHR pump discharge case opening, and step 6.2.12 requires a Quality Assessment Services inspector to verify the work is performed.

During a walk-through of the RHR cubicle, the team observed an appropriate FME

= .

t 52 cover on the pump case; however, the pump casing drain line and the seal water ,

heat exchanger lines were not taped off for FME control. The team discussed this issue with the maintenance manager and was informed that this discrepancy did not meet management expectations. The proper FME covers were installed.

c. Conclusions The team concluded that the maintenance activities were generally performed  ;

according to the appropriate procedures; however, the team noted a few examples I where procedure requirements were not adequately implemented.  :

M2 Material Condition of Facilities and Equipment M2.1 Material Condition of the RCS Vent Header & Isolation Valves f

a. Inspection Scope On September 3,1996, members of the team walked down the reactor coolant vent l header system with a reactor operator. The purpose of the system walkdown was to verify system alignment and configuration as described in the operating  !

procedure. The team also assessed the material condition of equipment associated '

with these events. ,

t

b. Observations /Findinas ,

a  :

The reactor coolant vent system is a temporary system that is installed during cold i

shutdown conditions. The purpose of the system is to remove gases from the RCS.  !

l The controls for installation and operation of the system are identified in NOP 2.9-6.

I The system is not described in the UFSAR, and there are no license requirements for its operability. The vent system is connected to the RCS loops, pressurizer, pressure relief tank, and the reactor vessel head.

i The team identified various equipment and procedural deficiencies during the I system walkdown. Hardware deficiencies included kinked tygon tubes, lack of an t i adequate levelin the moisture drain tank and condensate in the reactor head vent  !

I hose.

I The team noted several valves in the CVCS and RHR that failed to perform there ~

! isolation function. The leaking CVCS valves allowed nitrogen gas to enter the RCS. {

j Leaking RHR valves allowed water from the RCS to be diverted into the containment 4

sump. l i

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c. Conclusions The temporary reactor coolant vent system was in poor material condition as observed by the team on September 3,1996. The team noted that a series of G leaking isolation valves were the root cause for two of the events reviewed. The team concluded that based on the large number of valves that leaked, that this may I

be a broader problem that requires additional attention.

s. .

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Ill. Enoineerina u 'Facili les p e i E2.1 Technical Response to the Accumulation of Nitroaen Gas in the Reactor Vessel l

a. Insoection Scone The team assessed the licensee's response to the identification of nitrogen gas in the reactor vessel.

b. Observation /Findinat The E&TS organization was not involved in identifying the source of the excessive nitrogen gas usage or evaluating the cause for the increase in pressurizer and cavity levelindications. The E&TS staff were also unaware of the system configuration changes that allowed inappropriate maintenance to be performed on the RHR spray valves and the inappropriate closure of the RCS loop stop relief valves.

Following the identification of nitrogen gas in the reactor vessel oa September 1,1996, the licensee was unable to verify that all the nitrogen gas had been removed or that the reactor vessel was full of water. The. E&TS staff initiated several actions to verify the actual reactor vessel level. Plant l&C technicians obtained local RVLIS and CET readings on September 3,1996. The local RVLIS and CET readings were taken using available test equipment and existing procedures.

The readings indicated that the reactor vessel was nearly full. A special test was necessary to positively verify that the vessel was full.

The E&TS organization developed two special test procedures to determine reactor vessel level and to verify that the reactor head vent was not blocked. The tests were completed on September 5,1996. The test determined that the reactor vessel was not completely full and refilled the vessel.

On September 4,1996, a temporary modification was installed to connect three CETs to provide indication in the control room. The RVLIS control room indication was restored, using a jumper, on September 11,1996.

4 e ,

e I 54 l The shift manager had requested that the Shift Manager's Staff Assistant (SMSA) 4 determine the worst-case (lowest) reactor vessel level that occurred during this ,

event. The SMSA acquired background information, but did not determine a i; minimum reactor vessellevel. On September 3,1996, the Operations Manager

initiated actions to perform this calculation. A preliminary analysis was completed j on September 4,1996. At the conclusion of the onsite inspection activities, the i

i licensee had yet to complete a final volumetric inventory balance calculation.

[  ;

j c. Conclusions The E&TS organization was not proactive in identifying and resolving several j important technical anomalies that contributed to these events. The team 4

concluded that a communications failure between operators and E&TS was a j contributing cause for several of the events reviewed.

The initial actions taken by shift personnel, to refill the reactor vessel and the

CVCS, were appropriate following discovery of nitrogen gas in the RCS. However, >

l this initial corrective action was treated as a success without fully evaluating the

overall safety implications.

l The team concluded that the E&TS response to this event was not commensurate 3 with its safety significance. The E&TS organization failed to take prompt action to I

obtain RVLIS measurements or verify the actual reactor vessel level. The team noted that confusion among plant staff on the ability to collect this information con.ributed to this delay. The reinstallation of the control room CET and RVLIS

indications was not completed in a timely manner. The special tests, to verify j reactor vessel level and vent header operation, were also delayed.

t l E2.2 Technical Resoonse to the RHR Pumo Failure I

a. Insoection Scope

5 The team assessed the licensee's response to the failed "B" RHR pump.

l 1

! b. Observation /Findinas l

l The RCS loop stop valves were opened to allow the steam generators to be a backup heat removal source, if the operating RHR pump was to fail. The operators opened the RCS loop stop valves at 2124, on September 1,1996, about seven j hours after the "B" RHR pump was identified as having failed. By the evening of l September 2, the electric auxiliary feedwater pump was verified lined up to feed the

! steam generators from the domineralized water storage tank (DWST); the RCS j integrity and modified containment integrity were verified; the AOP for loss of RHR j was reviewed, and the nitrogen banks were isolated.

i I

i e

, 6 I 55 A RCP could be used to provide forced circulation of the RCS in the event the RHR system was unavailable. The RCPs could also be used to fill the steam generator U-

, tubes if nitrogen gas accumulates in the steam generators. A RCP and associated l support systems were aligned for service on September 3,1996.

< 1

On September 7,1996, a personnel barrier was in place to control access around l the operating RHR pump. At the same time, operator rounds in the RHR pump area

were increased from every eight to every six hours.

! c. Conclusions Following the failure of the RHR pump, the operations staff appropriately focused j their efforts on aligning a steam generator for natural circulation. The team '

concluded that actions, were not timely for restoring a RCP to standby service.

The team concluded that the actions taken to monitor the operating RHR pump were not timely or comprehensive. The licensee appropriately provided a personnel barrier to limit access around the operating RHR pump. However, the team concluded that a non intrusive pump performance monitoring surveillance should have been performed to provide assurance of proper pump operation. The "B" RHR pump was found to have significant thermal damage in area of the pump seals, yet no attempt was made to measure the operating pump's sealinlet and outlet temperatures.

E2.3 Technical Response to the RHR Heat E chancer Valve Leak

a. insoection Scone The team assessed the licensee's respnse to the "A" RHR heat exchanger inlet valve leak.

b. Observation /Findinas On September 1,1996, the RHR syste m engineer was on-shift as the SMSA. The operations crew requested SMSA input on the design limitations of flow through the RHR heat exchanger. The request was based upon maintaining RCS temperature, during an abnormal alignment of the RHR system, with only one of two heat exchangers. The applicable procedure, NOP 2.91, did not provide limitations on RHR flows or guidance on the alignment of the RHR system with one heat exchanger. The SMSA provided guidance to the crew on flows to limit RHR pump "run-out", and maximum flow rate through a heat exchanger based on design values identified in vendor information. The crew implemented this guidance on flowrates during the evolution to remove the "A" RHR heat exchanger from service.

The SMSA guidance was subsegemly incorporated into revision 31 to NOP 2.91.

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i i The RCS system engineer supported the operations crew by evaluating the location 4

of the valve body leak on RH-V 791 A, and provided recommendations to isolate the leak. The system engineer recommended closure of the valve, since the defect was l

j located on the valve body, whereas the valve disc would isolate RHR system

pressure to the leak. Operators subsequently isolated the "A" RHR heat exchanger j based upon the system engineer's recommendation.

! The licensee completed several actions following discovery of the defect in the RHR  ;

e heat exchanger inlet isolation valve, RH-V-791 A, on August 31,1996. For  !

example, operators shut the leaking valve as a precaution to protect the RHR I l system integrity, because little was known about the wxtent of the flaw or the risk f that it presented to the RHR system pressure boundary. An early decision was made to radiograph the valve. Although a contract radiographer arrived on site with a radiograph source that evening, the radiography was delayed pending an

evaluation of the issues concerning rigging the 600 pound source container into the )

1 "A" heat exchanger cubical.

l c. Conclusions i The licensee acted decisively following discovery of a through wall leak in the "A" 3

RHR heat exchanger inlet valve. The E&TS staff promptly addressed the decay heat

removal and system integrity issues. The actions taken in response to the leak j identified on the "A" RHR heat exchanger inlet valve body were appropriate.

E2.4 Review of Generic Information

! a. Inspection Scope

} The team assessed the licensee's corrective actions taken in response to industry j information on nitrogen gas intrusions into the RCS.

! b. Observations /Findinas l i l NRC Information Notice (lN) 94-36, Undetected Accumulation of Gas in Reactor 4

Coolant System, describes an event at another nuclear power plant where nitrogen gas accumulated in the reactor vessel. The E&TS staff documented an evaluation of this IN in a letter (ODM 94-134, dated July 23,1994). This evaluation stated a nitrogen gas bubble in the reactor vessel could cause a level decrease to four inches

, above the RCS hot leg center-line. The evaluation assumed that at this level the i nitrogen gas could pass through the pressurizer surge line to the pressurizer and be l vented to the vent header.

i The recommendations documented in this evaluation were to maintain the RCS pressurized in cold shutdown when feasible and to isolate the VCT if the RCS is depressurized. If the VCT cannot be isolated, the evaluation specified that (1) the

? reactor vessel head should be vented to the vent header; (2) to maintain at least

one channel of RVLIS available; and (3) to monitor the cold calibrated pressurizer i

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level for indications of a gas void in the vessel. The analysis explains that a gas l void formation in the vessel would result in an increasing level in the pressurizer.  !

The evaluation also recommended revising procedures to included indication of gas accumulation in the reactor vessel.

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c. Conclusions '

The team concluded that the E&TS staff conducted a thorough evaluation of the NRC IN. However, the recommendation to maintain one channel of RVLIS available i was not implemented. The recommendation to revise procedures was narrowly focused and did not provide guidance in the procedures used during this event. The team determined that the failure to implement actions based on this evaluation was a contributing cause for this event.

E2.5 Outaae Plannina and Schedulina

a. Inspection Scope The team reviewed the process for incorporating risk considerations in planning and control of outage activities. )

b. Observations /Fmdenas The licensee has established Wor ( Contiol Manual (WCW procedure 1.2-9, Outage Planning, Scheduling, and Implementatic ;, to provide the administrative controls to manage shutdown risk. The concept of designating system out-of service windows was selected as the planning control mechanism. The licensee has also developed planning schedules for the current outage, Refueling Outage 19, that define the i

scheduled windows for systems. The s'iutdown risk profile shows a significant increase in relative risk during reactor di tassembly until the reactor cavity is flooded. Therefore, during the events re viewed by the team, the shutdown risk profile indicates a relatively higher risk.

The licensee failed to apply appropriate :ontrols to work that resulted in a diversion

. of water from the RCS. This event resu.ted while working on motor operated j containment spray valves on September 4,1996. The activity was performed i

without the refueling water cavity filled and without contingency planning. This was contrary to WCM 1.2 9, Outage Planning, Scheduling and Implementation, i Section 1.2.7.b, which states that activities which impact the RHR system should be scheduled during seriods of low decay heat, maximum coolant inventory or defueled conditions. The containment spray valve work was inappropriately authorized after the RhR pump had been found failed.

Plant management made a decision en August 29,1996, to postpone refueling  ;

l outage maintenance activities over the 3-day weekend. A contractual obligation '

l granted the personnel supporting the reactor disassembly a day off on Sunday, j i September 1. This outage period has a high relative risk. The licensee decision to '

delay the reactor disassembly process before opening seal connections on the i

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• 1 reactor head was conservative. When RCS integrity is not maintained risk increases i because the steam generators are unavailable for decay heat removal. However,  !

the licensee failed to fully evaluate the reactor status or take compensatory I measures during the time that refueling activities were delayed. All four steam }

generators were isolated from the reactor vessel and the RVLIS and CETs were l disconnected. There was no evaluation of the risk presented by this configuration l- and there was no contingency planning made for obtaining RVLIS or CET data.

. The licensee's letter, dated June 15,1992, provides CYs response to NRC Generic Letter (GL) 88-17, Loss of Decay Heat Removal. In that letter CY committed to

.having at least two independent temperature measurements, representative of the l core exit, whenever the reactor vessel head is in place. Failure to connect the CETs l deviated from this commitment.

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c. Conclusions l The licenses has developed an appropriate work control procedure to minimize shutdown risk. However, certain decisions, such as, postponing the maintenance activities over the weekend without reactor vesselindications, were not thoroughly l reviewed for risk significance. In this case, the reactor water level was lowered and

the vessel isolated from all four RCS loops for an extended period, while the vessel  ;

temperature and levelinstruments were disconnected. The team concluded that while detailed guidance existed for minimizing shutdown risk, the implementation of l

} the guidance was not adequate. It apperrs that the decision not to connect the CETs was not consistent with the licenses's GL 88-17 response.

j E2.6 Reactor Coolant Vent System

a. Inaggetion Scone 1

The team assessed the quality of E&TS provided to install and operate the l

l temporarily reactor coolant vent header system. The quality of the vent system l

normal oparating procedure was evaluate

I in report section 03.2, operator training j was evaluated in report section 05.1, anri the material condition of the vent header l was evaluated in report section M2.1.

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b. Observations /Findinas 1

The team learned through interviews with various operators, that on at least two occasions, operators requested engineering support to improve the temporary vent

system. Licensee management did not initiate engineering activities to improve the

system, primarily due to the priority of other engineering work requests, or a lack of '

] a viable engineering solution.

i The team noted that very limited engineering support existed for the temporary vent  !

system. The system was controlled under NOP 2.9-6; however, no accurate i I drawings depicted the system alignment, nor were there any engineered design l limitations on the capability of the system. Additionally, the team learned that no

5 59 material specifications existed for either the valves or hoses, or any guidance on proper installation of the temporary hoses. The licensee improved the control of the system and provided technical guidance on design limits for the system in a procedural revision to NOP 2.9 6 on September 14,1996,

c. Conclusions The team concluded less than adequate engineering support was evident in the design, installation, and monitoring of the temporary vent system.

E2.7 Root Cause Determination of the Failure of the "B" RHR Pumo

a. Insoection Scoce The team reviewed the root cause failure analysis for the "B" RHR pump.

b. gbservations and Findinas j

The licensee determined that the following four factors contributed to the failure of

! the "B" RHR pump: (1) the pump shaft deflection reduced the clearance between

! the shaft and the throttle bushing; (2) the 316 stainless steel throttle bushing did

} not have good wear properties relative to the 316 stainless steel shaft sleeve. This led to galling of the throttle bushing when the bushing was allowed to contact the

shaft sleeve; (3) the out of round casing wear ring, the looseness of the head to i casing fit, and the eccentricity of the head to casing fit resulted in a reduction in

i. wear ring clearances, which allowed the shaft to contact some sections of the i throttle bushing before the wear rings contacted; (4) the manufacturer specified

! clearances in the wear rings and the throttle bushing areas did not provide sufficient

! margin to prevent contact between the wear rings and between the throttle bushing l and shaft sleeve.

The root cause for the pump failure was a combination of a marginal design and manufacturing defects. The combination of these manufacturing deficiencies led to the occasional contact of the shaft with the stationary throttle bushing. The j

stationary throttle bushing was located at the impeller end of the shaft. The

! occasional contact apparently broke tack welds that held the bushing in place.

4 Once the tack welds were broken, the bushing was free to move along the pump e shaft. The bushing may have caused the increase in vibration levels observed i'

during the September 29,1996, inservice pump test. The bushing eventually cocked and locked between the stuffing box bore and the shaft sleeve. This t

apparently occurred during a 45 minute pump run on August 19,1996. The locked j

bushing cut the shaft sleeve, and heated the shaft leading to severe bowing and

! additional shaft rubs. The rotor seized when the pump was stopped. The "A" RHR pump, rebuilt in 1990, eliminated several manufacturing defects, similar to the ones l

found on the "B" pump in 1996. The licensee plans to inspect the "A" RHR pump, l prior to entering mode 4 operation.

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c. Conclusions The team concluded that the licensee's root cause analysis was acceptable. While this root cause evaluation did not positively identify a root cause for this failure, a reasonable apparent cause was identified.

E7 Quality Assurance in Engineering Activities t

E7.1 Indeoendent Event Revim

a. Insoection Scope The team reviewed the licensee's independent event review response.

b. Observations /Findinas ,

The licensee established root cause investigation teams for the R4R pump failure and the leaking RHR isolation valve on September 2,1996. These investigation teams were later disbanded when an IRT was formed. Licensee senior management directed that an IRT be established on September 3,1996. The IRT was functioning with 9 of 14 members on September 4,1994. The IRT was to provide an independent assessment of the events and to determine the root cause for each event.

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c. Conclusions <

The team concluded that plant management's failure to fully appreciate the i

significance of this event resulted in a poor event response and in delays in initiating an integrated event recovery plan.

The team also concluded that the licensee was slow in developing a functional IRT.

The IRT investigation was not completed at the conclusion of the AIT onsite

inspection activities. Therefore, the quelity of the IRT's assessment was not reviewed.

E7.2 Manaaement Oversicht Summarv Management oversight of the nitrogen bubble event and the degraded RHR system was fragmented and protracted, resulting in untimely corrective actions for significant conditions adverse to quality. The untimely response was reflected in (1) the failure to fully appreciate the significance of the event; (2) the delays to establish actual reactor level

(3) the delays to reestablish control room reactor vessel level and temperature indications; (4) the delays in aligning a reactor coolant pump for service; (5) the untimeliness of actions taken to monitor the operating RHR pump following the "B" RHR pump failure; and (6) the delays to establish and

implement an independent review team.

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I-61 i 'm'nt Meetinos X1 Exit Meetina Summary l The team presented the inspection results to members of licensee management at the conclusion of the inspection on October 2,1996. The exit meeting was open for public observation. The slides used at the exit meeting are provided as

! Attachment 2 to this report. The licensee acknowledged the findings presented.

No proprietary material was knowingly retained by the team or disclosed in this inspection report.

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62 PARTIAL LIST OF PERSONS CONTACTED -

Mike Baca, Operations Unit Supervisor Lee Blaede, Operations Work Control Supervisor l

Rick Borg, Supervisor of Information Technology Gary Bouchard, Director of Work Services '

Jonathan Bower, Operations Shift Manager Will Chestnut, Operations Unit Supervisor John Calderone, System Engineering Sean Doody, Operations Nuclear System Operator Gerry Dreschler, IRT Member j Joseph Eldridge, Nuclear System Operator John Ferguson, Manager of internal Review Jeff Folden, Operations Unit Supervisor Jim Hawxhurst, IRT Member Mike Hess, IRT Member Steve Hutton, System Engineering Don Kidder, IRT Member Dave Lazerony, IRT Member Pierre L'Heureux, IRT Team Leader John Majewski, System Engineering Mike Marino, System Engineering Clark Maxson, IRT Member Roger McBeth, Nuclear Training Doug McCrscken, Acting Operations Manager Pat Motes, Nuclear System Operator Buster Orf, Nuclear System Operator

' John Pointkowski, Operations Shift Manager Phil Rainha, Operations Shift Manager Chuck Reid, Operations Reactor Operator Bill Rein, System Engineering i

Jack Stanford, Start Up issues Manager Larry Wellbrock, Operations Reactor Operator Marty Williams, IRT Member Dick Willis, Operations Shift Manager Neil Young, Nuclear Training LIST OF ACRONYMS USED AOP Abnormal Operating Procedure ACP Administrative Control Procedure ACR Adverse Condition Report ADT Aerated Drains Tank ANN Annunciator Procedure BAMT Boric Acid Makeup Tank CFR Code of Federal Regulations CET Core Exit Thermocouples 1

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e 63 CMP Corrective Maintenance Procedure CVCS Chemical and Volume Control System CY Connecticut Yankee EG Emergency Generator l EDG Emergency Diesel Generator EPIP Emergency Plant implementing Procedure E&TS Engineering and Technical Support

'F Degrees Fahrenheit FME Foreign Material Exclusion gpm gallons per minute HPSI High Pressure Safety injection IN information Notice IRT independent Review Team l&C Instrument and Control I kV Kilovolts l LORT Ucensed Operator Requalification Training LPSI Low Pressure Safety injection N/A Not Applicable NOP Normal Operating Procedure NRC Nuclear Regulatory Commission NSO Nuclear System Operator ODI Operations Department Instruction Psig Pound per square inch gage l P&lD Piping and instrument Drawing  !

PAB Primary Auxiliary Building PORC Plant Operations Review Commi. tee l

PM Preventive Maintenance '

PMP Preventative Maintenance Procedure ppm parts per million RO Reactor Operator RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RVLIS Reactor Vessel LevelIndication System

! RWST Refueling Water Storage Tank SI Safety injection

! SM Shift Manager SMSA Shift Manager's Staff Assistant SUR Surveillance Procedure TPC Temporary Procedure Change TS Technical Specification UFSAR- Updated Final Safety Analysis Report US Unit Supervisor WCM Work Control Manual VCT Volume Control Tank

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. i i Figure 1 1

s l The Undetected Nitrogen Gas

! Introduction into the

, Reactor Vessel from the Charging Line I

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. HE UNDETECTED NITROGE.e GAS INTRODUCTION INTO '

THE REACTOR VESSEL FROM THE CHARGING LINE Steam Vent Header generator # '

V Pressurizer Pressurizer

__ / "*"

v, r - 891* EL

" Head c '

Pressurizer Y'"I Air Level

- Head Full 355* EL-12%

- 351" EL Top of Head 4 - 301" EL '

p

. . , w Cavity - J - <

Reactor i r-233* EL e,h*U,'y 1hll. d .Y. I "

_ 3,; no" n Hot Leg * ^ ^ ~

1 Loop

-c ggoi .- o+ 1 I l o+ .WC, 3 Charging Stop l

_L- k 3 - . M.; , ' ~

-J _, Line vaeve Loop g l=

4 g , g.

-l Cold Leg f ,

~RHR '

Stop Valve N Cooling RHR inlet Fuel _=

piow

! Outlet .6 Cold leg 10* Pressurizer to*RHR Reacter ,

. , . , ,  !!!,? g ,' - *f* ,

Surge Line Suction Line Coolent ,

J .

L (Loop 4) (Loop 1)

""E N -

{!.* g- \ / '

EL 198* - f

/ EL 185* - 271r2" I  ;

EL 172* - h i

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y, ,[ Hot Leg Pipe l

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FIGURE 1 L_._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ________ _

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Figure 2 Chemical & Volume Control System j

CHEMICAL & VOLUMt CONTROL SYSTEM .

NITROGEN GAS LEAK PATH FROM VOLUME CONTROL TANK TO THE RCS To RCS L pf2 Volume Cor. trol Q

' To RCS Loopf4 Gas 30s  :

-^^^'

- ^- - - -

Hot Leg e A-v4

Open u

Regenerative Hxs M M **+8" y l '

A A A Velves #

7, Leaking By

= g r,

~

, V V V y

,, Bypass Open

[

)(  : 1 1- Boric t

Ble er  !

=

)(  :

73- "

l Valves Closed Charging l

Leaking By Pumps From H pumps FIGURE 2

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1 Figure 3 Residual Heat Removal System (RHR?

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- . . . . . . . . . . - . . . - _ . _ = . . . .. .-._ . . _.

RESIDUAL HEAT REMOVAL SYSTEM (RHR)

R ST From RCS Loop #1.

Hot Leg

, , , r Failed Pump'

,r X X B RH-V-7948 , RH-V-7918 l

To Loo #2 Cold Leg RHR Pumps w2 w2 RH-V-794A RH -791 A fA RHR Heat Exchangers Pinhole '

Leak in Valve '

Recirc to RWST FIGURE 3

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Attachment 1 i 1 i . AIT Charter i

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pmaseg

., w e, UNITED STATES 2 a NUCLEAR REGULATORY COMMISSION l

$ REGloN I 4 475 ALLENDALE ROAD g .... KING oF PRUSSIA, PENNSYLVANIA 19406-1415 September 4,1996 MEMORANDUM TO: James T. Wiggins, Director Division of Reactor Safety FROM: Hubert J. Miller o Regional Administrator Region 1

SUBJECT:

AUGMENTED INSPECTION TEAM CHARTER FOR THE REVIEW OF THE HADDAM NECK NON-CONDENSABLE GASES IN THE VESSEL HEAD AND LOSS OF DECAY HEAT REMOVAL CAPABILITY.

On September 1,1996, Haddam Neck operators responded to a lowering pressurizer level event with the plant in cold shutdown. Pressurizer level was maintained between 10%

and 5% over a four hour period by adding water from the refueling water storage tank l (RWST) to the reactor coolant system (RCS). Approximately 5000 gallons of water was added. Operators had been in the process of correcting excessive nitrogen usage to the volume control tank (VCT) and isolated a path to the charging line and into the RCS. The

] licensee believed that the RCS had a continueus vent established and that the non-l condensable gases in the upper head were bong vented. An earlier evolution to support RCS vessel head removal, which had lowered 3CS level to below vessel head flange, was -

terminated, and operators had reestablished 1.ivel at 10% in the pressurizer prior to the

start of the event. The licensee believes tha the gases introduced while lowering the RCS

, level were not vented upon reestablishing pressurizer level. Throughout the event

] adequate core cooling was apparently mainte ned.

?

Another problem involving through wall leaka ge from the body of the A residual heat

' removal (RHR) heat exchanger inlet valve had placed the decay heat removal system into a ,

one heat exchanger configuration. The A RH 3 pump was running with the B RHR pump i aligned in standby to that RHR heat exchange r. The actual valve body defect still exists

and assessment is being done of this valve to determine if it is considered functional; the licensee may examine the other 3 similar valves. Lastly, in order to align the B train RHR

! with the B train heat exchanger to support radiography of the A RHR heat exchanger inlet valve, the operators attempted to shift to the B RHR pump and identified indications of

pump seizure. '

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! Because the event is considered to be complicated and the probable causes are unknown or difficult to understand, per M.C. 0325, Paragraph 05.02, item c, I have determined that an Augmented inspection Team (AIT) should be initiated to review the causes and safety implications.

5 The Division of Reactor Safety (DRS) is assigned the responsibility for the overall conduct of this augmented inspection. James Trapp is appointed as the AIT leader. Other AIT members are identified in Enclosure 2. The Division of Reactor Projects is assigned the r

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a Haddam Neck AIT Memorandum 2 responsibility for resident and clerical support as necessary; and the coordination with other NRC offices, as appropriate. Further, the Division of Reactor Safety, in coordination with DRPis responsible for the timely issuance of the inspection report, the identification and processing of potentially generic issues, and the identification and completion of any enforcement action warranted as a result of the team's review.

Enclosure 1 represents the charter for the AIT and details the scope of the inspection. The inspection shall be conducted in accordance with NRC Management Directive 8.3, NRC Inspection Manual 0325, inspection Procedure 93800, Regional Office Instruction 1010.1, and this memorandum.

Enclosures:

1. Augmented inspection Team Charter

2. Team Composition 1

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4 I

ENCLOSURE 1 AUG,." TED INSPECTION TEAM CHARTER The objectives of this AIT are tc:

1.) Develop a sequence of events for the introduction and removal of nitrogen gas from the RCS. The sequence of events should also include the component failures and restoration of the decay heat removal system capability. The chronology should start early enough to include the last successful use of the B RHR pump and include any evolutions that may have impacted the pump performance. Also include key findings from Objective 7 below.

2.) Assess the shutdown risk and recovery actions available to the operators given the condition of the decay heat removal system during this event.

3.) Assess the quality of the operator actions, procedures, and training which resulted in the introduction of nitrogen gas to the RCS, as well as the second attempt by operators to start the 8 RHR pump af ter the first attempt had failed.

i 4.) Assess the operator actions taken to identify and respond to the nitrogen bubble in the reactor head.

5.) Assess the quality and timeliness of the response by the operations staff to implement compensatory actions for the decay heat removal system component )

failures.

6.) Conduct a root cause failure analysis, ir, parallel with the licensee's effort, to 3

establish the causes for the failure of th e RHR pump, reactor head vent and the RHR heat exchanger inlet valve leak. Include reviews of other involved equipment such as the VCT and the nitrogen source.

7.)

Assess CY management and technical s:aff (Engineering, Maintenance, Operations, Health Physics, Security, etc) response *o this event. Include an evaluation of the timeliness and extent of integration of ti e CY organization initial event response and recovery, as well as performance relativo to reportability decisions, prompt and final operability reviews and use of the correc tive action process.

8.) Review the August 28,1996 manipulation of the CVCS 355 valve, to determine this event's relationship, if any, to the introduction of nitrogen gas to the RCS.

Examine the potential for tha nitrogen gas introduction into other portions of the RCS piping and loops.

9.)

Review all aspects of this event, including the RCS instrumentation available to the operators for monitoring or assessing the extent of core coverage and reactor vessel level while shutdown, for generic implications, particulerly relative to industry operating experience with use of N2 as the motive force for the water movement during shutdown conditions.

l.

2 Schedule:

The AIT shall be dispatched to Haddam Neck so as to arrive and commence the inspection on Saptember 3,1996. During the site portion of the inspection resident and clerical support is available. The inspection report is due out 30 days after the formal exit. The team leader will also make recommendations to the DRS Division Director regardirig the need for conducting a public exit, the appropriate enforcement action, including draft NOVs and enforcement panel materials, followup needed for any issues left open by the team, and, potentially generic issues. Also, the team leader is expected to schedule a post-AIT lessons-learned meeting between DRS and DRP to examine enforcement opportunities relative to the conduct and administration of the AIT.

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l ENCLOSURE 2 i

TEAM COMPOSITION The assigned team members are as follows:

Team Manager: James Wiggins, DRS Onsite Team Leader: James Trapp, DRS Onsite Team Members: Tom Shediosky, DRP Eben Conner, DRP l

Robert Depriest, DRP Peter Habighorst, DRP John Munro, NRR .

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Attachment 2 l

j AIT Exit Meeting Slides j

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HADDAM XECK AUGMENTED INSPECTION TEAM l

1 DECAY HEAT REMOVAL SYSTEM CHALLENGES NRC INSPECTION 96-80 SEPTEMBER 3-16,1996

MAJOR' EVENTS REVIEWED

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1. UNDETECTED ACCUMULATION OF NITROGEN l

GAS IN THE REACTOR VESSEL HEAD i i

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2. TWO INADVERTENT DIVERSION OF REACTOR COOLANT SYSTEM INVENTORY l

3. DECAY HEAT REMOVAL EQUIPMENT FAILURES J

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EXIT MEETING FORMAT l f

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1. SEQUENCE OF EVENTS i

2. SAFETY SIGNIFICANCE

3. CONTRIBUTING CAUSES FOR EVENTS

4. EVENT RESPONSE 4

1 SEQUEXCE OF EVENTS UNDETECTED INTRODUCTION OF NITROGEN GAS INTO THE REACTOR VESSEL On August 28,1996, an operator incorrectly opened a valve, which resulted in injecting water and then Nitrogen into the RCS.

The valve was closed, but did not seat properly. This allowed nitrogen gas to continue to leak into the RCS from the volume control tank.

• Nitrogen gas accumulated in the reactor vessel head.

Displaced water from the reactor vessel entered the pressurizer.

• On August 29,1996, ~5C00 gallons of water was drained from the RCS in preparation for refueling activities. Later that day, ~1000 gallons was added to the RCS, when the decisio) was made to suspend refueling activities.

August 31,1996, the nitrogen gas bubble growth reached an equilibrium.

September 1,1996, nitrogen gas leakage was stopped by isolating the nitrogen to the VCT. Operators added

~4000 gallons to fill the RCS.

INADVERTENT DIVERSIONS OF REACTOR COOLANT SYSTEM INVENTORY

• On August 22,1996, operators failed to properly align

valves while placing the RHR purification system in service. This failure resulted in inadvertently diverting

! ~500 gallons of RCS inventory to the refueling water i

storage tank.

• On September 4,1996, plant operators inappropriately

. conducted work activities on the RHR containment spray

valves which diverted ~300 gallons of RCS inventory to

the containment sump.  :

{

DECAY HEAT REMOVAL EQUIPMENT FAILURES

• On August 31,1996, a plant operator identified a leak in the "A" RHR heat exchanger inlet valve. The valve was closed, removing the heat exchanger from service, to 4

isolate the leak.

• On September 1,1996, the "B" RHR pump failed to start and was found seized. The pump had previously operated on August 19, 1996.

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SAFETY SIGNIFICANCE l

THIS WAS A SIGNIFICANT PRECURSOR EVENT  !

Event Parameters

• Operators were unaware of actual reactor vessel level during this event.

• The time to an RCS temperature of 200 F was approximately 52 minutes if RHR was lost.

• Reactor vessel level decreased to approximately 21/2 feet above the point of RHR pump cavitation.

1

• If the nitrogen gas bubble had continued to grow, it would have caused cavitatio.a in the operating RHR pump.

Decay Heat Removal Challenges

• The two inadvertent RCS drain downs, had operators not taken prompt actions to terminate the diversion, could have resulted in cavitating the operating RHR pump.

The required boric acid flow path was compromised by the nitrogen gas in the CVCS.

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! l SAFETY SIGNIFICANCE  !

(Continued)

Decay Heat Backup Systems

• The "B" RHR pump and "A" RHR heat exchanger were not available for service.

• Venting the "B" RHR pump following maintenance was difficult even with the installation of additional vents.

• The loss of RHR procedure requires the charging pumps be operable to refill the RCS. These pumps were not available during this event.

e If the RHR system was lost, timely restoration of forced l circulation using a reactor coolant pump would have been i difficult.

e Nitrogen gas could have adversely impacted the ability of the steam generators to remove decay heat by natural circulation.

e The low pressure injection system would have been available to remove decay heat.

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l I CONTRIBETING CAE SES FOR EVENTS

1. Procecure Qua'ii:y was Poor i

• The use of ACP'1.2-5.3, Evaluation of Activities \ Evolutions Not Controlled by Procedure, was not consistent with technical specifications requirements.

The RCS drain down guidance was inadequate The vent header procedure was inaccurate and did not provide adequate detail.

The RHR purification system procedure lacked key sign-offs.

• The RHR system operating procedure guidance lacked i important guidance.

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2. Failure tO Implement Procedures 1

• Failure of a non-licensed operator to properly perform the RHR purification restoration procedure resulted in the August 22,1996, inadvertent drain down of the RCS.

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• On August 28,1996, a non licensed operator failed to align the boric acid flow path valves in accordance with l; procedure steps.

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• Operators failed to properly determine temperature and boron concentration prior to opening loop stop valves.

i Operators isolated nitrogen to VCT without a procedure.

Operators inappropriately closed the loop stop relief valves without a procedure.

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3. Lack of a Quest:ioning Attialce

• The lack of a questioning attitude resulted in the operators failing to adequately evaluate reactor coolant system level anomalies.

4

• Operations failed to respond in a timely manner to the excessive use of nitrogen gas between August 29 -

September 1,1996.

L. Inappropriate Decision Making

• Control room operators failed to properly control plant work activities which contributed to the September 4, 1995, RCS drain down event.

• Operators isolated the loop stop pressure relief valves without properly assessing the technical implications.

5. Pre-Job Briefings were Inadequate

• No pre job briefing before lineup of RHR purification pump on August 22,1996.

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• No pre job briefing before boric acid flow path alignment on August 28,1996.

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6. Lack of Ins 1:rumental: ion

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• The RVLIS and CET had been disconnected and were  ;

l not available. 1 4

• The cavity level system did not provide reliable or 4

adequate . reactor vessel level indication.

I 7. Training Not Adequa:e 1 1

• Failed to conduct scheduled re-qualification training on l outage and refueling operations. l

• Training conducted on RCS level indication and vent header operation was weak.

8. Poor Equipment Condition

• Several isolation valves in the CVCS system leaked by causing a leak path for nitrogen.

• A leaking manual RHR containment spray isolation valve allowed RCS to be diverted the sump.

• Poor design control and material condition of the i temporary reactor head vent system.

9. Poor Implementation of Generic Information l

• In 1994 CY staff failed to implement several recommendations made by technical support in response to NRC Information Notice 94-36.

10. Weak Use of Technical Suonort 1

• Operators failed to solicit technical support for l identifying the source of nitrogen gas usage.

• Operators did not solicit technical support to evaluate perceived loop stop valve leakage or to develop corrective action plans.

i 1 e Operators did not solicit tes:hnical support in developing a more detailed drain down procedures.

• Technical Support did not provide operators with local

^

i RVLIS or CET readings.

l e Work occurred on RHR containment spray valves

! without System Engineers knowledge.

I i

T

. l l

11. Inaoorooriate Planning and

^ ^ ~

i Scheduiine l i

• The schedule change to stop refueling activities over the weekend was not thoroughly reviewed for safety.

• Implementation of the September 4,1996 work activities that resulted in RCS inventory diversions was not consistent with the work control procedure to minimize shutdown risk.

i 4

4 l

l i

1 i

1 4

4 I

? ..

EVENT RESPONSE 4 1. Slow Initiation of Event Review Team

• The incident response team was not in place until 4 days following this event.

i

2. Unt:imeLv Technical Resoonse

!

• Local RVLIS and CET readings were not taken for 2 l days. .

4

.'

• Additional surveillance of the operating RHR pump were  !

not implemented for 5 days.

• The RCP and seal injection were not aligned for service for several days.

• Special tests to establish vessel level and vent header

operation were not performed for 4 days.

l

3. Avoicable Delays in RHR Purup Repair l 1

• Several avoidable delays were encountered during the replacement of the "B" RHR pump.

f

• Time to restore "B" RHR pump to operable status was 25 days.

1

i -

1

4. Failure to Report Event

• The required 4-hour 50.72 event report was not made for 10 days.

t l

s 4

I 4

5 4

i j

Y r OfficicI Trcnscript cf Precggdings

)

NUCLEAR REGULATORY COMMISSION

Title:

CONNECTICUT PUBLIC TELEVISION

BROADCAST RE NUCLEAR SAFETY

. ISSUES i

i Docket Number: (not applicable)  !

i Location: TELEVISION BROADCAST -

l l

Date: AUGUST 29,1996  ;

l l

i Work Order No.: Pages 1-59 NRC-951 NEAL R. GROSS AND CO., INC.

. Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.

Washington, D.C. 20005 (202) 234-4433 47oirioTrcr-o Gof f. Enclosure 3 4

O < 1 .

1 UNITED STATES OF AMERICA ]

2, NUCLEAR REGULATORY COMMISSION 1 ' +++++

4, SAFETY ISSUES  ;

5 +++++

6 CONNECTICUT PUBLIC TELEVISION BROADCAST 7 +++++

8 AIR DATE: AUGUST 29, 1996 9 +++++

10 IN THE MATTER OF 11 EXHIBIT A TO 10 CFR 2.206 12 PETITION DATED 11/25/96 ON BEHALF OF 13 CITIZENS AWARENESS NETWORK (CAN) 14 AND 15 NUCLEAR INFORMATION 'ND RESOURCE SERVICE (NIRS)  ;

16 17 18 19 ,

I IO The interview was held at Connecticut Public 21 Television, Susan Perry Luxton, moderator, and Gary 22 Verdun, guest speaker.

23 24 j l

25 NEAL R. GROSS COURT REPORTERS AND TRANSCRSERS 1323 RHODE ISLAND AVE., N.W.

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o .-

P-R-O-2-E-E-D-I-N-G-S 2 (Introductory music.)

MS. LUXTON: Hi, I'm Susan Perry Luxton, and 3 l' i

4 welcome to Nuclear Safety Issues brought to you by the 5 Citizen's Regulatory Commission. This evening I have with 6 me a guest who was a former employee of Millstone, Gary 7 Verdun. Gary, glad to have you here.

8 MR. VERDUN: Pleased to be here, Susan.

9 MS. LUXTON: Gary is a very powerful speaker 10 who we started to hear at the last few meetings, and we 11 have never heard before that. And so, he has quite a 12 story to tell relating to his experience working at 13 Millstone, and also the termination of January 11, 1996.

14 And so I thought it was something that you 15 should all hear because I ras duly impressed by -- not 16 impressed, upset's the worii more, I think, by -- by what I

, 17 heard from Gary. So Gary, let's talk a little bit about 18 your background first.

19 How long have you been at Millstone or were 20 you at Millstone before you were terminated?

21 MR. VERDUN: Well, I started at Millstone in 22 1980, and I worked there for 12 years as a Senior 23 Engineering Tech in a group called the Liability I

24 Engineering. And after 12 -- 8 years -- 12 years with the 25 company, I should say, I was promoted to Site Maintenance

, 1 NEAL R. GROSS

< court REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W (202) 234-4433 WASHINGTON. D C 20005-3701 (202) 234-4433 I

w e 3-1 Supervisor.

2 And at the point in time, I took on a major 3 responsibility of assuring that the site was in good 4 [::nd:. tion with respect to material condition of the site, 5 .the :leanliness of the site, the --

6 MS. LUXTON: The " site," you mean all three 7 reactors?

8 MR. VERDUN: All of the support buildings, 9 grounds, the maintenance of all support buildings on the 10 units and some support for the units themselves.

1~ MS. LUXTON: " Material condition," you mean 12 hardware and housekeeping?

13 MR. VERDUN: All -- all the housekeeping on 14 the site and all the maintenance of the support buildings 15 --

16 MS. LUXTON: Yes.

17 MR. VERDUN: -- HVAC work, repair of all the 18 doors on the site, which there's a couple of thousand 19 doors, and all of the janitorial services --

20 MS. LUXTON: Yes.

21 MR. VERDUN: -- for the site.

J 22 MS. LUXTON: Yes.

23 MR. VERDUN: So, the job was a pretty big job.

24 MS. LUXTON: It sounds like a big job.

25 MR. VERDUN: And it --

it was different from I

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a 4 lm

. the previous job because the previous job was all 2, technically oriented, and it involved a lot of work on 3Uschingrotordynamicsproblems--

4 MS. LUXTON: Yes.

5 MR. VERDUN: -- problems with rotating 6 equipment, vibration problems on -- on large rotating 7 equipment. So --

9 MS. LUXTON: In nuclear -- in the nuclear i 9 field?

10 MR. VERDUN: All in -- well, we did work on 11 the nuclear plants as well as fossils and hydro plants.

12 MS. LUXTON: And this was when you were with 13 Combustion Engineering --

14 MR. VERDUN: No, this was --

15 MS. LUXTON: --

or when you went --

16 MR. VERDUN: -- this was with North --

17 Northeast Utilities --

18 MS. LUXTON: All right.

19 MR. VERDUN: -- from 1980 until about 1992, I 20 did that kind of work. And then from '92 until 1996, I 21 worked at the site as Site Maintenance Supervisor.

22 MS, LUXTON: Okay, so that's 19 -- so, 19 --

23 is that 16 years?

24 MR. VERDUN: Sixteen years, yes.  ;

25 MS. LUXTON: (Laughing).

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o , 5 1 MR. VERDUN: a lot of the years at Northecst 2 Utilities were wonderful years. If I was really -- since  ;

3 M'?B7, when the plants really began to have problems and we 4 saw -- began to see a degradation at the plant in terms of tne etndition of the plant, in terms of procedural 6 violations, in terms of, you know, morale problems 7 beginning to surface at the plants, and this all kind of 8 coincided with cut-backs that were imposed on the staff as 1 9 a result of the company trying to position themselves in a 10 competitive situation -- ,

11 MS. LUXTON: Yes, yes, 12 MR. VERDUN: -- for things that were coming 13 down the pike with retail wheeling --

14 MS. LUXTON: Deregulation?

15 MR. VERDUN: -- and deregulation and -- and 16 that kind of thing.

17 MS. LUXTON: Yes.

18 MR. VERDUN: So, a lot of problems began to 19 emerge in 1987 as a result of cut-backs at the site:

20 money cut-backs, manpower cut-backs, and that sort of 21 thing. And by the time I became a supervisor in 1992, the 22 cut-backs were really becoming very severe.

23 And -- and then they were beginning to talk l I

24 about staff reduct;< .s to the Northeast Utilities' staff. j 1

25 And of course, that --

that heightened and increased the NE%L R. GROSS COURT REPORTERS AND TRANSCRIPERS 132f RHODE ISLAND AVE., N W l

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e. e  ;-

1 concern of the employees for their jobs and i: had a 2l detrimental effect on morale --

MS. LUXTON: Yes.

~

) 4 MR. VERDUN: -- along with severe budget cuts 5 from 1992 until 1996. That bunch of cuts were so severe 6 that the degradation at the plant began to increase at a 7 faster rate.

8 MS. LUXTON: And you were involved with -- -

9 talking about degradation of the plant -- of identifying 10 those kinds of problems --

11 MR. VERDUN: Yes.

12 MS. LUXTON: -- by -- by your MCP --

13 MR. VERDUN: Material Conditions Program.

14 MS. LUXTON: -- program? Right. That was --

15 I so, let's talk a little' bit -- but first, I think, talk 16 about the Material Condition Program, becat.tse that was a 17 big part of what -- what you -- you did at Millstone in 18 those last final years.

19 MR. VERDUN: Yes, it was.

20 , MS. LUXTON: But was -- that had to do with 21 the housekeeping, right? And -- and when you talk about 22 " housekeeping," we were talking about that earlier today, 23 and that was an interesting thing to learn: how important 24 that is in a nuclear power plant.

25 MR. VERDUN: Housekeeping --

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o. o 7 .

F 1 MS. LUXTON: I mean, it's not just like 2 sweeping the floor.

i 3 ll MR. VERDUN: No, there's a lot more to i'; than  ;

-i 'enat. Housekeeping is a very important issue --

i 5 MS. LUXTON: Why?

6 MR. VERDUN: -- at nuclear power plants' 7 because it involves aspects of fire safety. And fire I 8 safety at nuclear power plants is a very important issue.

9 So, it's very important to keep the plants very clean and 10 in a condition that looks good to everybody, and not only 11 looks good, but is safe from the standpoint of 12 combustionable materials and --

13 MS. LUXTON: Right.

14 MR. VERDUN: -- and that sort of thing.

15 MS. LUXTON: ' hey're not tripping over things.

16 There's not rags --

17 MR. VERDUN: Right.

18 MS. LUXTON: --

and whatever --

19 MR. VERDUN: Right. I 20 MS. LUXTON: --

hanging out.

21 MR. VERDUN: Right. Part of it is keeping 22 things tied down, keeping everything in order, and 23 everything has a place and everything should be in its 24 place.

25 MS. LUXTON: And that's the fire safety part, NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RMODE ISLAND AVE., N W.

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e- , B-1

but there's also the radiation part --

)

2 MR. VERDUN: Yes, that --

it 3' MS. LUXTON: -- for instance, in rad : ste.

-i , MR. VERDUN: Yes, that -- that --

those kinds 5 of issues are normally taken care of by a whole crew of 6 people that are assigned to rad waste.

7 MS. LUXTON: Yes.

8 MR. VERDUN: But the - .one of the jobs that I 9 got --

10 MS. LUXTON: Then what were the people doing 11 if -- if they weren't --

if they had a whole crew of 12 people in charge of rad waste, Gary, what were they doing 13 all these years that they let rad waste go like that? l 14 MR. VERDUN: You're talking about Unit I rad 1 15 waste?

16 MS. LUXTON: Right, Unit I rad waste, when i

17 they --

18 MR. VERDUN: Basically, I think they became 19 complacent with a very bad situation, an antiquated .,

20 facility --

21 MS. LUXTON: Yes.

22 MR. VERDUN: --

that certain piping systems 23 that process radioactive waste were antiquated. They'd 24 been cut out of the piping loop. Some pumps had been shut 25 down, were no longer used.

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9 1l A lot of the systems in there were still I

2j radioactive pipes. They hadn't disposed of those pipes or I

3 nll:ut them out of the system because it was going to be just 4 000 expensive, the Utility thought, to dispose of those 5 systems at that time --

6 MS. LUXTON: Right.

7 MR. VERDUN: --

because of the radiation 8 aspects of it.

9 MS. LUXTON: So, they allowed it to just kind 10 of deteriorate and deteriorate and they didn't get the 11 money --

12 MR. VERDUN: So --

13 MS. LUXTON: -- the funding --

14 MR. VERDUN: Yes.

.5 MS. LUXTCN: -- the department people didn't 16 get the funding to clean it up, so they just kind of lived i 17 with it?

18 MR. VERDUN: "he -- the technicians and the 19 people that had to work ir. those areas lived with that 20 situation. They kind of became used to it. And it really l 21 wasn't until I was assigned a major project by Don Miller 22 about a year and a half ago.

1 23 It was near Thanksgiving in 1994, that I -- I j 4

24 was asked to do a special assignment where I reported 25 directly to him and -- and created a program called the NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHoDE ISLAND AVE.. N W.

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l

1 Meterial Condition Program for the site, which was 2 basically a comprehensive inspection program, a visual 3 ! inspection program, of the site to identify any kinds of 4  :+ n ::.encies that we could see, hear or sense by sensory 5 .! perception. And at that time, 6 MS. LUXTON: What do you mean by "senschy 7 perception?"

8 MR. VERDUN: Well --

9 MS. LUXTON: I know there's something wrong --

10 MR. VERDUN: (Laughing).

11 MS. LUXTON: --

(Laughing).

12 MR. VERDUN: No, no. We -- we didn't have any 13 --

AS MS. LUXTON: No fortune tellers there, right?

15 MR. VERDUN: - ESP kind of stuff going on 16 there, or at least I don't think we did. (Laughing).  !

l l

17 MS. LUXTON: So, who was in charge, you? '

18 MR. VERDUN: I was in charge of the program,

19 and the program basically -- I was -- I was told what the

^

1 20 company wanted for a program and some of the fundamental i 1

21 aspects of the program. And then I was turned loose to )

22 design the program and come up with something that would 23 work for the site for a comprehensive inspection program.

24 So, . what I came up with after a fashion was --

25 and -- and working with Don Miller on this also was we

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• * .;j

-l would have 28 managers from the site, the Millstone site,

- perf:rm inspections once a month.

3y It would take about four to six hours of their 4  ; me per month. Each of the managers would be assigned a 5 zone to inspect, which was a -- like a small geographical 6 area of the site.

7 MS. LUXTON: Yes.

8 MR. VERDUN: Their job would be to go into 9 that area, or those areas, that they were assigned to, and 10 they would look for deficiencies --

l 11 MS. LUXTON: Yes. <

12 MR. VERDUN: -- any kind of deficiencies:

13 housekeeping deficiencies, leaks in pipes, noisy 14 machinery, machinery that was vibrating excessively that 15 might be developing a problem.

16 MS. LUXTON: Okay, why?

17 MR. VERDUN: To -- to basically put the plant 18 in a better material condition. We were trying to l 19 identify problems that we could later, you know, get them 20 identified in a database, and then organize an effort to 21 get these things repaired.

22 MS. LUXTON: So you could become an exemplary 23 nuclear power plant?

24 MR. VERDUN: That's right, because the 25 condition of the site was beginning to look very haggard.

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,e e _.

1 MS, LUXTON: Yes.

2 MR. VERDUN: And Don was, and many other

! 3 pe:p'e were, very concerned about that.

1 4., MS. LUXTON: Yes.

ll 5 MR VERDUN: And it slipped, like I said, 6 since 1987. So, this assignment came along in November of 7 1994. It had slipped into a pretty seriously -- a pretty 8 bad state of affairs --

9 MS. LUXTON: Yes.

10 MR. VERDUN: -- from appearance and also from 11 the condition of the machinery and the performance of the 12 machinery. So --

13 MS. LUXTON: So, this was a good thing?

5 14 MR. VERDUN: Oh, it was an excellent program.

15 MS. LUXTON: Yes, i

16 MR. VERDUN: I spent a lot of time designing 17 the program, setting up tracking systems for the program.

18 MS. LUXTON: Did --

19 MR. VERDUN: Eventually, we created a huge 20 database.

21 MS. LUXTON: Yes. How many -- how many things i 1

22 did you find wrong that needed to be fixed?

23 MR. VERDUN: In a year and a half, we l l

24 identified 2,613 items that were entered into the database 25 that needed to be addressed. Like I said before, a lot of l NEAL R. GROSS court REPORTERS AND TRANSCRISERS 1323 RHoDE ISLAND AVE., N W (202) 2344433 WASHINGTON. D C. 20005 3701 (202) 234 4433 1

i

, _2 l

1 them were minor things: housekeeping issues, appearance )

i 2 , things.

Yes.

MS. LUXTON:

4 MR. VERDUN: But there were some things in the 5 database that had far-reaching implications in terms of 6 effort, money, engineering, time, and, you know, really 7 seriously impacted the --

8 MS. LUXTON: Performance of the plant?

9 MR. VERDUN: Yeah, and not only that, but the 10 -- you know, the radiation exposure of personnel and the 11 safety of people working aro"nd that. kind of material.

12 MS. LUXTON: Yes, okay. So, what happened?

13 MR. VERDUN: Well, probably the best example 14 really was the, like you mentioned earlier, the Unic I rad 15 waste facility. When I went into that facility for the 16 first time in my life, I was -- I was approached by people 17 that were working in those areas.

18 And basically, they were concerned about 1

1 19 conditions that they were asked to work under --

20 MS. LUXTON: Yes.

21 MR. VERDUN: -- because the area was so -- it 22 was strewn with temporary pipes, temporary hoses, 23 temporary processing equipment. There were lots of pipes 24 in there that were still radioactive pipes that were not )

25 being used to process fluid, but hadn't been removed that, I

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• * *4 L l

.1 you know,-people had to work around these. things and walk l 2 by them all the time. ]

3  ! So, the people felt that, you know, they.were 4 ggetting exposure, that -- radiation exposure that they i !I l 5 I!,didn't have the -- l 4 MS. LUXTON: That was not necessary.

7 MR. VERDUN: --

necessary, and there was a lot l

l 8 of contamination, of course, in that area because of the 1

9 fact that they're processing these fluids.

i 10 What was particularly bad about it is that all

11 of these hoses and tubes running all over the floors and 12 everything were not plumbed into the system as hard piping i

"~

13 systems in a permanent manner.

I 14 Everything was kind of temporarily set up.

15 You know, it looked to me like it --

l 16 MS. LUXTON: And it had been temporarily set 17 up for years you mean?

18 MR. VERDUN: For years and years. And -- and 1 .

i 19 -- and everything -- every time they did something down l

20 l there, it was always done temporary, you know, temporary.

i.

i 21 ' So, they'd take out one temporary thing, and put in 22 another temporary thing. As technology changed and better

23 things came along, they'd take out one temporary thing and 24 put in another, I guess.

i

'2 5_ So, but they never got to a point where they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE 15.AND AVE., N W.

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i* ,

1 ever plumbed anything back in hard, you know --

2l MS. LUXTON: Oh, I see.

4 3 MR. VERDUN: -- with piping systems. So, it i

4 .,was a mess.

5 MS. LUXTON: So, what finally happened, Gary?

6 Did they fix it? I mean --

7 MR. W E UN: Well, they didn't --

8 MS. LUXTON: -- they didn't fix it then.

9 MR. VERDUN: After -- after a lot of 10 discussion about it and after working with one of the 11 managers that was responsible for inspections in that 12 'a re a , we got an engineering effort mounted to correct 13 those problems.

14 One of the major things though that -- two 15 things -- two major thinss in that area that concerned me 16 a great deal: one of them was a tank that had a -- that 17 had rotted out. And apptrently, there was radioactive 18 sludge that was leaking <>ut of the base of the tank.

I 19 The technici.tns in that area referred to this 20 -- this condition that had developed there as the ant 21 hill. So, the sludge would leak out slowly and a mound of 22 radioactive sludge had built up underneath the tank that 23 was extremely hot stuff.

24 So, somebody somewheres along the line decided 25 that it was going to be too expensive to repair this tank, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISt.AND AVE., N W.

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l

}

. ,t take

_t cut or put a new rank in because of the high 2 radioactive levels. So, they essentially put up a 3 i::n rete wall around the whole thing and kind of roped it i

4 Off and, you know, put a lead shield in and stuff like  !

5 that, and left it there.

6 So that, I thought, was, you know, an extreme 7 situation that absolutely should have been corrected. It 8 shouldn't have been a question of economics. It just 9 should have been fixed.

10 MS. LUXTON: Yes, yes.

11 MR. VERDUN: So, I made an issue out of that.

12 And I tried to get the films that people had taken on an i

13 inspection a year or two before. I never could get the i

\

14 films. I kind of was jerked around, so I never saw the 15 actual films.

16 I understand that they were taken during an 17 NRC inspection -- inspection. So, NRC people --

18 MS. LUXTON: Was aware of it?

19 MR. VERDUN: -- knew about it. They had seen 20 it, and they had not --

2".I MS. LUXTON: Years before?

22 MR. VERDUN: Years before, and they had not 23 imposed a situation on those plants -- on that -- on the 24 plant to correct that situation.

25 There's a pipe chase down there also that I NEAL R. GROSS COURT REPO8tTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.

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• i 17 1 undarstand is an extremely bad situation that is corroded 4

2 out, a rotted out system, that's an extremely hot area 3 i tnat people have been into, and is another area that 4 techn:. ians are very, very concerned about down there.

i 5 And I began to raise, you know, issues about 6 getting that area corrected also.

7 MS. LUXTON: Okay.

8 MR. VERDUN: So, there were -- there were big 4

9 things that came out of this program. It was not all 10 just, you know, painting --

11 MS. LUXTON: Right, right. ,

12 MR. VERDUN: -- floors and machinery and --

13 MS. LUXTON: And fixing valves and whatever?

' 14 MR. VERDUN: -- and fixing the valves. There i .5 were other things that wera involved that were big, big 16 items.

17 MS. LUXTON: ' es, like how about the thing 4

18 with the chips, the paint chips in Reactor Building II? i l

1 19 MR. VERDUN: Oh, where the containment --

l 20 MS. LUXTON: I love that story, in the 21 containment.

22 MR. VERDUN: Well --

23 MS, LUXTON: Was that part of your deficiency?

24 MR. VERDUN: Well, there -- there was a i l

25 condition in contain -- that you're talking about --

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4 . is

..I

" I

- MS. LUXTON: Yes.

2la MR. VERDUN: -- that's in Unit II --

h 3', MS. LUXTON: Yes. I 4 MR. VERDUN: -- containment where the --

5 there's a steel liner that circles containment --

6 MS. LUXTON: Yes.

7 MR. VERDUN: -- that is basically bolted to 8 the --

9 MS. LUXTON: Cement?

e 10 MR. VERDUN: -- the concrete --

11 MS. LUXTON: Yes. >

12 MR. VERDUN: --

the cement --

13 MS. LUXTON: Yes.

14 MR. VERDUN: -- which is eight or ten feet 15 thick.

16 MS. LUXTON: Yes.

17 MR. VERDUN: *lhis liner is a quarter of an 18 inch or a half an inch thick, something like that. And it 19 -- it's a painted -- the 31ner is painted.

20 MS. LUXTON: Yes.

21 MR. VERDUN: And my understanding of what 22 occurs in that area is that during thermal temperature 23 change that's inside a container, that the -- this shell 24 grows faster than the concrete does because it heats up at 25 a quicker rate from the changes in the air temperature and NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.

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fi l

, i:

1 the ambient temperature inside the-containment.

2 So that what's created there is a differential 3 h -hermal expansion --

4. MS. LUXTON: Yes.

~: MR. VERDUN: -- between the liner and the 6 concrete. And as that -- as that liner tries to grow --

l 7 tried to grow out against the concrete, which is not  !

i 8 growing at the same rate, so it -- since it tries to grow l

9 out, it can't go anywhere outwardly. So, it ends up 10 buckling.

5 11 So, that liner is buckled.in several areas 12 inside containment because of the differential thermal y 13 expansion problem. But what comes into play here is that 14 it put such tremendous stresses on the liner, that the r 15 paint that's applied to the liner peels off. And it peels l 16 off in chunks that are about the size of giant potato r

17 chips, probably so big. They just curl right off the

18. liner.

19 The -- the problem with the situation there is 23 that there are so many of these paint chips that have 21 peeled off the liner that when -- and this occurs on --

22 usually on start up --

23 MS. LUXTON: Yes.

24 MR. VERDUN: -- or during the process of f 25 running -- you know, while they're running the reactors, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N W (202) 234 4433 WASHINGTON. O C. 20005 3701 (202) 234-4433

l

'. that a let of these paint chips begin to appear. And what 2 ,has :cacerned the plant, and they've known about this for

.3 ' years, is that if there's ever an accident inside of Unit 4 ': and they have to spray down inside the containment, 5 there's a good possibility that all that water. rushing out 6 of the sparges and all the piping systems that are used to 7 quench the steam inside the containment, all that water 8 hitting those potato chips, or paint chips, could bust the 9 paint chips off, break them free.

10 They would fall to the floor --

11 MS. LUXTON: And clog up the drain.

12 MR. VERDUN: --

where there would be a torrent 13 of water running across the floors. They would go into 14 the drain, clog the drains. And in those sumps where the 15 drains are, those -- that's the suction for the pumps that i

16 then recirculate the water, bring it back up to the l 17 spraying headers and then re-spray it to continue 18 quenching steam and keeping, you know, whatever situation 19 has developed under control.

20 MS. LUXTON: Right. So, if it was -- so 1

21 ; technically then, theoretically, if that was clogged, then 22 they wouldn't be able to quench the steam.

23 MR. VERDUN: Right.

24 MS. LUXTON: And the steam would build up and 25 that would be a pressure --

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.' MR. VERDUN: That would create --

2 MS. LUXTON: --

problem.

3 i MR. VERDUN: -- some major problems --

4; MS. LUXTON: Right.

i 5

MR. VERDUN: -- inside there because they 6 would not -- no longer have a source -- well, it woUld 7 eliminate one major source of water that they would need .

8 to control a situation inside containment.

9 MS. LUXTON: Okay, now --

10 MR. VERDUN: So, that's a scary situation.

11 MS. LUXTON. It'a a scary. situation.

12 MR. VERDUN: And it's something that --

13 MS. LUXTON: And you identified several --

14 MR. VERDUN: --

needs to be addressed.

! 15 MS. LUXTON: --

of these.

- 16 MR. VERDUN: So,.that --

17 MS. LUXTON: Similar.

18 MR. VERDUN: -- the painting problems inside i

13 the containment were identified by --

20 - MS. LUXTON: By this program?

21 ' MR. VERDUN: --

this program, yes.

22 MS. LUXTON: Right. Okay now, the bottom line 23 is what happened to the program? Did they do the 24 inspections month -- every three months? And did -- did -

25 - and -- and so where's the program now?

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4 e ,

22

. You are no longer there. Where's the program?

I 2 That's the question?

t 3

MR. VERDUN: Well to my knowledge, when --

4 iwnen : was terminated or laid-off, whatever you'd like to

' I 5 call it, they -- they use the euphanism, you know, " laid-

! 6 off." But really, essentially what it was was an illegal 7 termination.

8 MS. LUXTON: Yes.

9 . MR. VERDUN: Nobody wanted to go and it was  ;

i 10 not necessary for anybody to leave. At that point in l 11 time, this Material Condition Program essentially was  ;

12 abandoned to the best of my knowledge.

13 I know of nobody that picked up on the program ,

14 to administer the program, to --

l '

15 MS. LUXTON: 'What about Mr. Miller? It was 16 his program.

17 MR. VERDUN: I don't know. Basically, I tried 18 to contact Mr. Miller after the termination to see if they i

19 would want to bring me back in through some, you know, job 20 shop --

! 21 MS. LUXTON: Contract, oh, yes.

22 MR. VERDUN: -- to work for Millstone to l

23 continue to administer the program. But, I was referred 24 to -- to the new senior person on the site, and basically I

25 didn't get a call back from him. Later called again to NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.

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t . 23-o

another vice president who had been involved n this, and 2 -- and he referred it to a director,-who was the director 3 . worked for prior to being terminated.

4' He did call and he told me in a phone 5 conversation that they -- they would decline the offer at 6 that time. They weren't interested in identifying any new 7 problems at the site. 'They were focusing all their 8 attention on problems that they alr.eady knew about, which 9 --

10 MS. LUXTON: Well, these aren't new problems -

11 -

12 MR. VERDUN: -- pretty much astounded me 13 because, you know, they should continue to look for 14 problems all the time at those sites. It's an ongoing, 15 living process.

16 MS. LUXTON: Right, right. I 17 MR. VERDUN: It's not something that you just, l 18 you know, abandon at som point in time. But the bad part 19 --

20 5 MS. LUXTON: Well, but --

)

21 l MR. VERDUN: -- about it --

1 1

22 MS. LUXTON: --

they don't really want to know 23 though, Gary, I don't think.

24 MR. VERDUN: I doubt it because --

25 MS. LUXTON: Because -- l NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.

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7 1 .e ,

4

~

MR. VERDUN: -- this program was essentially 2, acandoned as far as I can'see, and the database has not

-  :.anged from the time that I, you know -- I was -- I was  ;

4 nterminated at the site.  !

5 MS. LUXTON: Seven months ago or whatever.

6 Now,-do you think that this program that you did was just 7 getting a little bit too close for comfort for them, and a that's part of the reason why you got terminated?

9 MR. VERDUN: Well after L lot of thought on 10 the issue, .you know, I could think of no sensible reason 11 why I was terminated. . Essentially, I was doing two jobs 12 for the company. I was performing the duties of a Site 13 Maintenance Supervisor. I was doing everything I could to 14 do as much as possible in that area.

15 And I was performing the duties of this 16 Material Condition Program. It's a very large job. It 17 involved a lot of effort to maintain the databases and 18 organize the --

19 MS. LUXTON: Yes.

20 MR. VERDUN: -- the effort.

21 MS. LUXTON: And to take it home work, you did 22 at home.

23 MR. VERDUN: Oh yes, a lot of -- a lot of 24 times I took stuff home, and then get reports out to all 25 the senior level management, You 1:now, reports went out NEAL R. GROSS CoORT REPORTERS AND TRANSCRISERS 1323 RHoOE ISLAND AVE N W.

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I e ,

a. 5

to acout 50 people that were tailored to their level --

2 MS. LUXTON: Level.

3 MR. VERDUN: -- in the company, so that

-i everybody knew exactly what all the problems were at the 5 site, so that the appropriate people could put pressure on 6 the right people to get all these deficiencies cleaned up.

- 7 MS. LUXTON: But see, this is what I see, a Gary. This is exactly what they were talking about at the 9 NU meetings with the NRC. They had all these studies.

10 They put all this money into this enormous program to get 11 the place in better shape, and then they just let it go.

4 12 MR. VERDUN: And then they abandoned it, 13 right. ,

l 14 MS. LUXTON: And they abandoned it. And even l l

15 before they terminated yot and it was abandoned, people  !

l 16 weren't even doing it.

17 MR. VERDUN: fell, we had problems -- Miller 18 had problems right along <ith getting some of the managers 19 to do their inspections. And he was frustrated with that.

20 MS. LUXTON: fes.

21 l MR. VERDUN: And he tried 00 a number of times 22 I was at meetings where he actually raised some hell about 23 it with the managers. But, it didn't seem to make any 24 difference.

25 At first, it seemed to be a very effective NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISt.AND AVE., N W (202) 2344433 WASHINGTON O C. 20005 3701 (202) 234 4433

~

1 ,

26  ;

Frogram and the managers were -- managers were 2 participating. But as time went by, the effort tapered

~

ff until we got down to, you know, I think it was around

, 5. ;r 60 percent participation --

5 MS. LUXTON: Yes.

6 MR. VERDUN: -- per month by manager. And you 7 know, by then, you know, things were beginning to-fall off 8 rapidly. It felt that upper level management was not 9 supporting this program.

10 MS. LUXTON: Yes.

11 MR. VERDUN: And I was beginning to get some 12 hegative reactions from my director and my supervisor.

13 Some things happened in the office area that could 14 certainly be considered harassment by both my supervisor 15 and my director.

16 MS. LUXTON: '.'owards you?

17 MR. VERDUN: '.'owards me.

19 MS. LUXTON: * 'e s .

19 MR. VERDUN: and I began to feel that this 20 ; Material Condition Program was getting too hot for the 21 company to handle.

22 MS. LUXTON: Yes, yes.

23 MR. VERDUN: So, I went to Nuclear Safety 24 Concerns, who was supposed to be a safe island for people 25 to go to --

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t . 2-

- MS. LUXTON: Right, rignt.

I i 2j! MR. VERDUN: --

when they think they're having n

'l 3 'pr c' ems that involve nuclear safety concerns issues.

4 MS. LUXTON: Yes.

5 MR. VERDUN: I talked to them about it, and 6 basically I was told that, you know, they would discuss l

7 the situation with my supervision: my director and my l l

8 immediate supervisor and with Don Miller and, you know, 9 they would try to resolve the problem. ,

i 10 So after several weeks or so, there were --  ;

1 l

11 there were meetings about it. And we.-- I thought that we 12 , pretty much had resolved the issues, and that, you know,  !

i 13 this harassment would stop and -- ,

14 MS. LUXTON: Yes.

15 MR. VERDUN: -- and that things would work out l 16 --

17 MS. LUXTON: And what happened?

18 MR. VERDUN: -- okay. And about two weeks, 19 two or three weeks after the meeting, I was told that I l 20 was tarminated. So, you know, I basically -- I went -- I 1

2-

. followed the company procedures. I worked within the e

4 22 guidelines and framework of the company.

23 I believed in what they told me about working i 24 within the system --

25 MS. LUXTON: Yes.

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I e ~5' 1 MR. VERDUN: -- that the company set up --

2 MS. LUXTON: Yes.

3, MR. VERDUN: -- for a nuclear safety concerns 4  :.s s ue . I gave them due respect to work within that 5 system.

6 MS. LUXTON: Yes.

7 MR. VERDUN: And it -- they did not protect me.

8 They -- they -- they betrayed my trust, and basically 9 I was set up for the situation that --

i 10 MS. LUXTON: That you're in now?

11 MR. VERDUN: -- finally transp3 red.

12 MS. LUXTON: I'm sorry, Gary. I hate to say W

13 it, but you're one in a long list, an ever-growing list.

14 And hopefully, it won't be growing any longer now.

4 15 I mean, they keep saying that -- that they've 16 changed and that they're so much better, the Nuclear 17 Concerns Program, but I don't believe it. Let's talk 18 about this matrix business, about the -- about the

19 termination.

l 20 l MR. VERDUN: This is going to be the joke of I

21 all jokes.

22 MS. LUXTON: We've heard this before from --

1 23 MR. VERDUN: Yes.

24 MS. LUXTON: -- Jim Plum, from Harry Scully.

25 I mean, this is the joke of all jokes.

l NEAL R. GROSS  !

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. MR. VERDUN: Yes.

3 MS. LUXTON

How many years did you have 3 +:g e rience , 16 years experience? Guess what? In the 4 , rating folks, the rating -- they had -- they were rated in 5 different weights. In experience, the weight of 6 experience was two.

7 MR. VERDUN: Yes.

8 MS. LUXTON: I can't believe that experience 9 is only rated two.

10 MR. VERDUN: Well --

11 MS. LUXTON: Education was rated the next 12 lowest, five.

13 MR. VERDUN: There -- there's nothing sensible 14 about this process at all. I mean, it's -- it's a very 15 subjective process.

16 MS. LUXTON: One man did this?

17 MR. VERDUN: One man basically did this to me, 18 that's right. One man, who took probably ten minutes to 19 fill this matrix out, changed the course of my life 1

j 20 forever.

21 And nothing that was on this matrix was based 22 on past performance review; nothing was based on the i

23 effort that I put forth for the company; nothing was based 24 on, you know, the fact that essentially I was working two

, 25 jobs for the company.

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' e a.

I 1 MS. LUXTON: Yes, yes.

2 MR. VERDUN: It wasn't involved with -- it 3 i dn't have any bearing on -- with me or with anybody 4 e.se, by the way, that was involved with this matrix --

5 MS. LUXTON: So, you feel that --

6 MR. VERDUN: -- nobody was treated fairly.

7 MS. LUXTON: Right. So, you feel you were 8 wrongly terminated.

9 MR. VERDUN: Oh, absolutely.

10 MS. LUXTON: So, what -- so, this is a part of 11 the show that we want to talk about too. I would like, if 12 'there's anyone watching who also was on that January t

13 termination and would like to -- to call in and maybe I 14 share what happened with them and how they felt about 15 their termination -- maybe you felt it was fine.

!. 16 And I don't know if anybody feels that their i

17 termination is fine. But you're actually saying you think s

18 it was illegal.

13 MR. VERDUN: Well, I think it was illegal 2: because what -- what first alerted me to the fact that 2_ ,this might be illegal is I went to an NRC meeting --

22 MS. LUXTON: Yes.

- 23 MR. VERDUN: -- in a simulator building at --

24 at Millstone.

25 MS. LUXTON: Yes.

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. I

8 , 31 1 MR. VERDUN: And I think it was two or three 2 weeks ago And during that meeting, I was surprised to t

3 tifind out that the NRC made a statement that -- when they 4 were asked about these terminations and whether -- how the 5 investigation was going --

6 MS. LUXTON: Yes.

7 MR. VERDUN: -- from the standpoint of the NRC 8 looking into this stuff, that basically the investigation 9 of the firing of 104 people had not been completed, that 10 it had kind of been bounced around within NRC structure, 11 Office of Investigation, that they had pretty much come to 12 the conclusion that they might refer the whole matter to 13 the Justice Department.

14 MS. LUXTON: Yes.

15 MR. VERDUN: And if it went to the Justice I

16 Department, that there could be criminal prosecution --

17 MS. LUXTON: Yes, yes.

18 MR. VERDUN: -- as a result of the whole 19 affair. Because a lot c f these people that were i 20 terminated by use of this crazy thing that Northeast 21 Utilities developed to target people that they wanted to -

22 - to terminate, for whatever paranoid reasons that were 23 going through their mind, they -- a lot of them were 24 involved in nuclear safety issues.

25 MS. LUXTON: Yes, yes.

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' O 3-MR. VERDUN:

. 50 if that can be linked to --

2 y:n know, these terminations can be linked to illegal --

l 3I MS. LUXTON: Then criminal prosecution can be 5 MR. VERDUN: Yes.

6 MS. LUXTON: --

can be on the horizon. Now 7 listen --

8 MR. VERDUN: That's right.

9 MS. LUXTON: -- the key word here is " criminal 10 prosecution."

11 MR. VERDUN: That's right.

12 MS. LUXTON: Now, we were going to show the 13 folks at home the Code of Federal Regulations, remember?

l 14 MR. VERDUN: That's true, yes. j l

15 MS. LUXTON: Now, Gary brought in the C.F.R, l 16 the Code of Federal Regulations. Now, you've heard many 17 times on our show, and also in the newspaper -- I mean, 18 you can't have followed this issue without seeing these 19 letters, "C.F.R. 20.20" or "C.F.R. 50.54."

20 So, this is what C.F.R. is, the Code of 21 Federal Regulations. These are the laws that guide the i 1

22 nuclear power plant in its operations, right?

23 MR. VERDUN: That's right.

24 MS. LUXTON: Okay. So in this C.F.R. 50 -- 50 25 section, we looked up the employee protections part of it, NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W i (202) 234 4433 WASHINGTON. D C. 20005 3701 (202) 234 4433

e , :3-1 C.F.R. 50.7. And hare it says, " Discrimination by a i

2 licensed -- commissioned licensee against an employee for 3 engaging in certain protected activities is prohibited.

4 Discrimination includes discharge and other actions that S relate to compensation,. terms, conditions or privileges of 6 employment."

7 And then it goes on. And then it says, "Any 4

8 employee who believes that he or she has been discharged '

9 or otherwise discriminated against by any person for 10 engaging in protected activities specified in Paragraph 11 (a) -- (a) (1) of this section, may seek a remedy for the 12 discharge or discrimination through an administrative l

13 proceeding in the Department of Labor," which you've done.

I 14 Okay, and then it says, " Violation of 4

15 Paragraphs (a), (e) or (f) of this section by a 16 commissioned licensee," that would be like NU, "an 17 applicant for a commissiored licensee, or a contractor or l

18 a subcontrettor of a commissioned licensee or applicant 19 may be grounds for (1) denial, revocation, or suspension 20 of a license; (2) imposition of civil penalty on the 21 license or applicant; and (3) other enforcement action."

, 22 Now, this is a big point because " harassment 23 and intimidation of employee," not -- whether you're 24 wrongfully terminated or not. Whether you were wrongfully 25 terminated or not is one issue.

i NEAL R. GROSS COURT REPORTERS AND TRANSCRf8ERS 13E1 RHoDE ISLAND AVE.. N W.

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'i , 34 i But the harassment and intimidation, we know 2 that has gone on more than once, right?

3l1

, MR. VERDUN: Absolutely.

4 MS. LUXTON: More than twice. I mean --

5 MR. VERDUN: Absolutely.

6 MS. LUXTON: -- the people we've had on this 7 show: Blanche, Delcort --

8 MR. VERDUN: Plum.

9 MS. LUXTON: -- Plum, Tim O'Sullivan, Gladys, 10 and Benttencourt, all those people, you -- I mean, we're 11 talking up to six, seven, eight, nine.-- and Pete 12 Reynolds. I mean, if this -- with the (a) violation, 13 we've got five or six. Why isn't the NRC --

14 MR. VERDUN: Wait, we have nine. ,

l 15 MS. LUXTON: U 1at -- nine. l l

16 MR. VERDUN: Nine up to this point. Now --

17 MS. LUXTON: T.1at's with the --

18 MR. VERDUN: I believe that, you know, that I 19 -- that I am the tenth person who's come forward to say 20 that they think that they were treated wrongfully by they 21 company with respect to the protected activities.

22 MS. LUXTON: Okay, so it says right there --

23 MR. VERDUN: So I mean, this --'this is a l

24 repeating occurrence at this utility. It's something that '

25 needs to be corrected.

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35' 1 MS. LUXTON: By the NRC.

2 MR. VERDUN: And the NRC needs to take some 3 real tough action, and they need to take action against 4 the people that are doing this things, not against the 5 corporation.

6 There are people involved in this stuff that 7 need to be -- need civil penalties --

8 MS. LUXTON: Right, so -- so the civil --

9 MR. VERDUN: -- because they're violating 10 federal rules when they do these things.

11 MS. LUXTON: Right, or "the denial or 12 revocation or suspension of a license or other enforcement 13 action." Well what they've been doing is the "other 14 enforcement action."

15 MR. VERDUN: Right, but they have never --

16 they have never taken --

17 MS. LUXTON: Complied.

18 MR. VERDUN: -- their licenses away from them.

19 They've never revoked tneir license. They've never really 4

20 played hard-ball with this utility. They give them a 21 $100,000 fine or something. A S100,000 fine to this 22 utility is a joke.

23 MS. LUXTON: Right.

24 MR. VERDUN:  : mean, they laugh about it.

25 It's like you and I going out and getting a speeding NEAL R. GROSS COURT REPORTERS AND TRANSCRIS5RS 1323 RHODE ISLAND AVE N W 1202) 234 4 33 WASHINGTON D C 20(XM 3701 +202 2W33

• 26 1l ticke: --

2 MS. LUXTON: Yes.

h 3 :- MR. VERDUN: -- and then they got -- the 4 ":fficer writes it out and says that this is a four cent 5 fine, 6 MS. LUXTON: (Laughing).

7 MR. VERDUN: You know?

8 MS. LUXTON: That's a real deterrent, right?

9 MR. VERDUN: Yes, so, you know, it's that kind 10 of thing.

11 MS. LUXTON: Well, we've got a call.

12 MR. VERDUN: The penalties are not severe 13 enough. l 14 MS. LUXTON: Let's see who it is. Hi, you're l

15 on Nuclear Safety Issues. I 16 MALE PHONE CALLER: Yes, hi, Susan. How are 17 you?

18 MS. LUXTON: Good.

19 MALE PHONE CALLER: Great. And your guest, 20 good to see him, but not in these conditions that, you I

21 know --

22 MS. LUXTON: Yes.

23 MALE PHONE CALLER: -- that are prevailing.

24 But you know, he was talking, and you also were talking of 25 rad waste, et cetera. And at one time, I did work for rad NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.

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o 37' 9

1 waste.

2 MS. LUXTON: Oh, did you?

3 MALE PHONE CALLER: Yes. And I'm going to 4' 'tell you what --

just exactly what transpires out there, 5 and it's one of these things to hell with it -- to hell 6 with it. That's the big word, and the gentleman is 7 agreeing with me, shaking his head yes.

8 MS. LUXTON: What do you mean --

9 MALE PHONE CALLER: But it --

10 MS. LUXTON: -

"to hell with it?"

11 MALE PHONE CALLEP: I'll tell you exactly 12 what's happened. We've had episodes out there where I ,

13 have'actually been in Unit II. I've done work in there:

14 piping, et cetera. I've seen puddles on the floor.

15 And I've been told, " Clean that up." I've 16 said, "Let's get PHP in here and let's get a test going on 17 it," --

18 MS. LUXTON: Yes.

, 19 MALE PHONE CALLER: -- and so on. "Just wipe 20 it up and then we'll -- we'll check the rags up after you 21 wipe it up," blah, blah, blah. "We have messoline rags,"

22 et cetera.

23 Then you say, "No, I -- you know, get somebody 24 down here first." Well, they would raise cain with you.

25 And then the next thing would be, "To hell with it. Just NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHoDE ISLAND AVE.. N W.

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e- , 3a' i leave it. Go over and do that over there."

2 Now, many a time I have seen puddles that have 3 ,been -- been, excuse me, been tested, and the counters i,

4 dwculd just go so erratic, it would -- it would make your 5 head spin.

6 MS. LUXTON: The radiation counters?

7 MALE PHONE CALLER: Certainly, quite 8 certainly. You know, you had leaks here, there and 9 everywhere.

10 MS. LUXTON: So --

11 MALE PHONE CALLER: And every time you 12 reported a leak, you were a pain in the neck, you know?

13 "What were you doing over there? What were you doing over 14 here?" I, one time, worked out there for one solid week is in one particular area, and I don't know how many reports 16 I made for one valve dripping, one valve.

17 I kept reporting that valve and reporting that 18 valve. I was finally told, "Why don't you mind your 19 business?"

20 MS. LUXTON: Really?

21 MALE PHONE CALLER: That's right.

22 MS. LUXTON: I mean, I -- why do I say 23 "Really?"

24 MALE PHONE CALLER: That's right. Well, three 25 weeks later I finally went back in that area, three weeks NEAL R. GROSS CoORT REPORTERS AND TRANSCRIBERS 1323 RHoOE ISLAND AVE., N W.

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~

• o 39 1 later, and that valve was still dripping.

2, MS. LUXTON: Oh.

3 MALE PHONE CALLER: Okay? Now, I went back'to 4 y boss and told my boss. I said, "What in the hell is 5 going on around here?" He said, " Don't get involved in 6 junk like that." Can you imagine?

7 Well needless to say, I don't work for 8 Millstone. I got the hell out of there --

9 MS. LUXTON: I'll tell --

10 MALE PHONE CALLER: -- because I -- I just 11 couldn't take it anymore up there, you know?

12 MS. LUXTON: Well, you know, I think you 13 probably did a good thing for your health-wise too, 14 because those are not the kind of conditions we want is people working in.

16 MALE PHONE CJLLER: For certain. Well --

17 MS. LUXTON: You're talking about radiation 18 here.

19 MALE PHONE CJLLER: Well, you know, this Don 20 ( Miller and this whole gang out there, you know, each --

2. every one of them, again, I keep repeating the same thing.

l 22 Criminal charges should be brought up against each and l i

23 every one of them.

24 Mr. Fox, I don't know what the devil NRC is 25 thinking about with this Mr. Fox. But that's the guy that NEAL R. GROSS COURT REPORTERS AND TRANSCRSERS 1323 RHODE ISLAND AVE., N.W.

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a o 40' everybody's got to go after. He's known for many, many ,

t 2; years exactly what's happening in that plant.

3 'i MS. LUXTON: But you know what? You see, the i

4 j'NR --

l 5' MALE PHONE CALLER: But you know what's 4 happening? His -- his understudies are the ones that are -

7 --

that they're throwing hatchets at. You know what I'm 1

8 saying?

9 MS. LUXTON: Yes, right, right.

10 MALE PHONE CALLER: You can't hold --

11 MR. VERDUN: Let me tell you a story about Mr.

l 12 Fox, sir.

1 13 MALE i) HONE CALLER: Pardon me?

5.

14 MR. VERDUN: Let me tell you a story about Mr.

15 Fox that probably a lot of listeners don't know, but a lot 16 of the people at the plant do know about. One time as a 17 young engineer, he was an engineer on Unit I.

^

18 MALE PHONE CALLER: Yes.

4 19 MR. VERDUN: So, he certainly knew or should 20 have known about the refueling violations that were taking 21 place with respect to the full-core off-load. It has been 22 going on since day one.

23 And he was an engineer in Unit I Engineering.

24 He was around that site. He was intimately familiar with l 25 that site for years before he began to progress up the NEAL R. GROSS COURT REPORTERS AND TRANSCRSERS 1323 RHODE ISLAND AVE.. N W (202) 234-4433 WASHINGTON. D C. 20005 3701 (202) 23W33

e e 4. l

. ladder. j 2 MALE PHONE CALLER: Right.

it 3 'l MR. VERDUN: And incidentally, Bob Bush just 4 !;:t hred because of his attitude towards employees  ;

5 primarily I believe, and his arrogant attitude towards l 1

6 employees, and the fact that he was nothing but a -- a )

7 bully that pushed his weight around and --

8 MALE PHONE CALLER: I knew Bob Bush, sure. ,

I 9 MR. VERDUN: -- and basically exerting -- l 10 exerted his -- his influence to put people'in situations l l

11 where they took risks with safety, in.my opinion.

12 MALE PHONE CALLER: Yes.

13 MR. VERDUN: Mr. Fox condoned that behavior 14 from the beginning. He was Mr. Bush's boss.

15 MALE PHONE CAI.LER: Yes.

l 16 MR. VERDUN: That guy needs to go.

17 MALE PHONE CAI LER: Oh yes.

18 MR. VERDUN: They need to get him out of there 19 fast.

20 MALE PHONE CALLER: He needed a muscle-man out 21 there to look good with the stockholders. That's exactly 22 what it amounted to. And again, that situation you're 23 talking about, the tank that you saw the sludge, high  !

24 radiation sludge, and so on, that goes on in I, II, and 25 III all the time, you know?

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y - , _ __ ._ ..a . . -

• o 4-1 And again, it's the dollar over the safety.

2 You know what I'm saying?

3 MS. LUXTON: Yes. Thank you, sir. I'm going 4' t: --

I'm going to go to the next caller, okay?

5 MALE PHONE CALLER: My pleasure.

l 6 MS.-LUXTON: Thanks so much for calling.

7 MALE PHONE CALLER: Right.

8 MS. LUXTON: It was great to hear from you.

9 Wow, how about that, a confirmation of -- we've heard this 10 before.

11 MR. VERDUN: Well, it's kind of a confirmation 1

l 12 of what the employees in the -- in the rad waste facility j 13 told'me when I was down there, that they didn't feel i 14 comfortable about working under those conditions.

15 MS. LUXTON: *es.

16 MR. VERDUN: 'Ihey had grown kind of complacent 17 about it and realizing the.re was nothing they could really 18 do to change the situatier --

19 MS. LUXtON: Yes.

20 MR. VERDUN: -- by themselves. So, they were 21 kind of reaching out to me --

22 MS. LUXTON: What a shame.

23 MR. VERDUN: -- to help them.

24 MS. LUXTON: What.a shame. Unfortunately --

25 MR. VERDUN: And so I tried to do --

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e o 43 1, MS. LUXTCN: Unfortunately, you were cut off 2 before you could. Okay, let's go on. Now, we -- we 3 vanted to ask other employees to call if they could 4 because you'd like to do a support group, right --

5 MR. VERDUN: Absolutely.

6 MS. LUXTON: --

for the terminated -- for the  !

7 people who were terminated?

• i 8 MR. VERDUN: Right. One of the primary '

9 reasons why I'm here tonight is I want to try to get the l 10 word out to the 104 people that were terminated -- i

) 11 (Telephone ringing).

12 MR. VERDUN

-- on January lith to get in  ;

13 touch with us sogehow through either getting in touch with 1

14 you or calling me directly so we can get a support --

15 MS. LUXTON: Do you want to give your number? i i i

. 16 MR. VERDUN: -- group going. My home phone 17 number is (860) 442-6339. You can call me, tell me your

18 story. And we need to get a big database going so we can 19 get probably a class action lawsuit going against these 20 people.

21 MS. LUXTON: Good, i 22 (Dial tone).

23 MS. LUXTON: Uh-oh. Hi, did I lose it? You 24 know, I've been doing very well with the phone for the 25 last few weeks.

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15 7i 234 4433 WASHINGTON, D.C. 20005-3701 (202) 234 4433

44 1 MR. VERDUN: iLaughing).

2 MS. LUXTON: I really have.

MR. VERDUN: Yes.

3 !]

4' MS. LUXTON: I think I lost you. Please call 5 back. I'm sorry. I thought I was -- oh no, I think I 6 pushed -- wait a minute. Hello?

7 MALE PHONE CALLER: Yes, hello.

8 MS, LUXTON: Oh, I'm sorry. I thought I'd 9 lost you. Gee, I saved myself here. How you doing?

10 MALE PHONE CALLER: Okay, thank you. Good 11 evening.

12 MS. LUXTON: Good evening. What's on your 13 mind?

14 MALE PHONE CALLER: Just an easy question.

15 MS. LUXTON: Yes.

16 MALE PHONE CALLER: Any -- any chance the 17 Millstone could be shut down forever, and to make the --

18 safe?

I 19 MS. LUXTON: Who wants to take that? Do you 20 want to take --

21 MR. VERDUN: I'll take that one. I don't 22 think -- I don't want to see Millstone -- Millstone shut ,

And Susan doesn't want to see Millstone l

23 down forever.

24 shut down forever either, 25 MS. LUXTON: I don't? Wait a minute, now.

l NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W (202) 234-4433 WASHINGTON. D C. 200053701 (202) 234 4433 I

0 . 45 1 MR. VERDUN: Well -- 5 i

2 MS. LUXTON: Speak for yourself, Gary.

L 3 :i MR. VERDUN: Ah, well --

4 ,j MS. LUXTON: You go first --

V '

~

MR. VERDUN: Okay.

6 MS. LUXTON: -- and then I'll go. l 7 MR. VERDUN: I'll -- I'll speak for myself. I B don't want to see Millstone shut down forever. What I 9 want to see is I want to see Millstone adhering to the t

I'd want to -- I don't want to see 10 letter of the law.

l 11 these people breaking federal laws.

, 12 I don't want to see them discriminating 13 against employees. I don't want to see them breaking i

l 14 Department of Labor laws. I don't want to see them 15 challenging every single law and every single issue that -

! 16 - that comes up and trying to push it to the maximum --

17 MS. LUXTON: Yes. 1 i

18 MR. VERDUN: -- and stretch everything to the i )'

l 19 limit.

20 MS. LUXTON: Yes, i

21 MR. VERDUN: That's the kind of situation that

' 22 got them in this situation. That's the kind of thing that l

23 led to the current situation. They have no credibility 4

24 with the community. They have no credibility with the 25 NRC.

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1

,e-They have to no credibility with any of tha

)

people that were terminated. They have no credibility l 2

3 Swith their own employees because their own employees see

t. I 4 wnat's happening to other people,, and they know that it's 5 probably not too long before they're going to get the same 6 kind of treatment.

7 MS. LUXTON: Yes.

8 MR. VERDUN: They have no credibility with 9 their stockholders, and they shouldn't have any 10 credibility at this point with their Board of Directors.

11 But for some reason, upper level management 12 still seems to have retained some kind of credibility with 13 the Board of Directors. That -- that's something that's 14 beyond my ability to comprehend.

15  ! The Board of Directors needs to flush these 16 guys out, find out who the people are that have been 17 forcing the company into this situation, got them into 18 this embarrassing situation with all these violations of i

19 federal laws, state laws, employment laws.

20 They've got -- Board of Directors, you've got 21 to get these people out of here. Because if you don't, 22 all the stockholders in your company are going to lose 23 their shirts. That stock won't be worth -- ,

1 l

24 MS. LUXTON: Even more than they have, right? i i

25 MR. VERDUN: -- two cents --

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V . 47

MALE PHONE CALLER: -Excuse me, but --

2 MR. VERDUN: when these people are done with 3 y::u .

-i ' (End of Tape 1, Side A.)

5 (Start of Tape 1, Side B.)

6 MALE PHONE CALLER: You're probably right, Mr.

7 Gary. I shut up.

i 8 MS. LUXTON: You know what?

9 MR. VERDUN: No, you're still on.  ;

10 MS. LUXTON: It's --

11 MALE PHONE CALLER: Yes, please. The ,

12 gentleman there, Mr. Gary, is absolutely right. But this 13 -- another question. I've got the family here. They 14 can't sleep, you know, at night just think about what's 15 going on over there. A St of people they're moving out 16 from Waterford.

17 MS. LUXTON: Hmmm.

i 18 MALE PHONE C11LER: They say how -- you know, 19 you cannot trust anymore I'm surprised the -- the l I

, 20 government, Mr. Roland get involved more, no?

j. 21 MS. LUXTON: I'm surprised also.

22 MALE PHONE CALLER: I know. I nean --

l 23 MR. VERDUN: I think everybody is very 24 surprised with Mr.'Roland's stand on this issue. He came i

25 to Millstone one day and spent a couple of hours here, did 1 NEAL R. GROSS  !

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" <*-s - --.----- -_._--__-

• . . , 46- 1 I

1 a cursory inspection of the site, and then made some 2 pronouncement -- pronouncement that he thought the site 3 was' safe, and left, i

4 '

I mean, I think that that basically was --

5 MS. LUXTON: Irresponsible.

6 MR. VERDUN: -- totally irresponsible on the 7 part of the government.

8 MALE PHONE CALLER: Yes.

9 MR. VERDUN: He's not looking or listening or 10 reading or hearing what's actually happened down here --

11 MS. LUXTON: Right.

12 MR. VERDUN: -- at the site. If you're 13 listening, Mr. Governor, if you'd care to give me a call, 14 I'd be more than happy to take a couple of minutes of my I

. 15 time and talk to you about this. I do have a pretty busy 16 schedule, but I think I could fit you in.

17 MS. LUXTON: (Laughter.) Thanks so much for 10 calling, sir.

19 MALE PHONE CALLER: Thank you, Susan. Good 2, night.

21 MS. LUXTON: Good night. Well, you know, 22 getting back to what I think about it, it's not that I 23 want to see them shut down forever. But if they can't run 24 safely, I definitely don't think they should start up 25 until we can be convinced they do.

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o e 49-

. I think I'd feel a lot better if they waren't 2 here at all. But I think we, as a state, have to start 3 '!~Ocking to alternatives to nuclear energy. We're so

,; d e .t e n d e n t on them.

E And I think it's wonderrul that Mr. Blumenthal o

6 came out with this whole deregulation: all the 7 information on deregulation, making it known to the 8 public, step by step of the way, because that's looming on 1

9 the horizon. We're going to have a show on that too.

10 But, let's go on.

11 MR. VERDUN: Susan, I'd like to interject one 12 ' thing --

13 MS. LUXTON: Go ahead.

14 MR. VERDUN: -- before you go on to the next 15 issue.

I 16 MS. LUXTON: 'fou can interject anything, Gary.

17 MR. VERDUN: 'ie ' ve -- we have -- we ' ve l l

18 forgotten about 13 other tery important people --

19 MS. LUXTON: fes.

20 MR. VERDUN: -- in this process, and those are 21 13 people that were working for me in 1994 --

22 MS. LUXTON: Yes.

23 MR. VERDUN: -- early in 1994, that were also 24 terminated at Millstone. There are 13 housekeepers that j 1

1 25 were terminated in early 1994, and they were also l NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHoDE ISLAND AVE., N W.

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subjected to the sama kind of a process that th s 104 2 were.

There's really not 104 people's livelihoods

4. nat are at stake here. It's a 117.

I 5 MS. LUXTON: Seventeen.

6 MR. VERDUN: So if any of the 13 people 7 working for me that were terminated by the company, and 8 without my knowledge incidentally or my approval, without 9 any discussions with me whatsoever, I'd like to have those 10 13 people also get in touch with me because I want to try 11 to get them included in this support group.

12 The sooner we can take this thing with a --

13 some kind of a class action or law suit, but we need to 14 work together on this as a united effort to have the 15 greatest impact.

16 MS. LUXTON: Oh, that's great, Gary. Good for 17 you. Good for you. All ;ight, let's talk about what was 18 in the news today. Now, ve had two articles in The New 19 London Day today. One at the bottom of The New London Day 20 was "NRC advising quick evacuation in the event of a 21 nuclear accident." Well, isn't it about time they got 22 around to doing that? Two miles out, they want them 23 evacuated immediately.

24 But they said they've found that people 25 evacuate up to 50 miles out, Gary, up to 50 miles out NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.

(202) 234-4433 WASHINGTON. D.C. 20005 3701 (202) 234 4 433

4

• c.

,i

'l C  ; cecause like Mr. Sheridan said to us once with the 2 emergency evacuation plan, "If they find out the wind is go:.ng a certain way, they'll evacuate that neighborhood."

3 4 MR VERDUN: Because that's --

5 (Telephone ringing).

6 MR. VERDUN: -- where the plume -- that's 7 where the radioaction goes.

8 MS. LUXTON: Right where the plume is going.

9 But so I'm -- I'm right on the border of that 10 neighborhocd. .And I see everybody over there going, 11 right, and I'm --

12 MR. VERDUN: Yes.

13 MS. LUXTON: -- supposed to sit here and 14 think, "Well," --

15 MR. VERDUN: (Laughing).

16 MS. LUXTON: -- "okay, no problem. I think 17 I'll have a cup of iced tea."

18 MR. VERDUN: (Laughing).

19 MS. LUXTON: Forget it, they all move out.

20 MR. VERDUN: Yes.

21 MS. LUXTON: It's ridiculous, the evacuation l 22 plan. Let's see if I can get these people.

23 (Dial tone). {

l 24 MS. LUXTON: Oh my God. Hi, you're on Nuclear 25 Safety Issues.

l NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N W.

9 02) 234 4433 WASHINGTON. D C. 20005-3701 (202) 234 4433 4

. . -- - - . _ . . . . _ . . _ _ . -.- . - - . - . ~

!* e.- 52 1 FEMALE PHONE CALLER: Hello?

2 MS. LUXTON: H i '.

1 3

FEMALE PHONE CALLER: Hi, I have a question

?  ; i::: Jary.  ;

!. 5 MS. LUXTON: Sure.

l 6 FEMALE PHONE CALLER: Why did he wait 16 years f

7 waiting to get laid-off? Why didn't he leave before that ,

i t

8 if it was so unsafe?

9 MR. VERDUN: I did not say that the plant was 10 unsafe for 16. years. I said that the plant began to

! 2.1 degrads in 1987, and that -- and that.by -- by 19 --

i 12 because of budget cuts and because of management attitude, J 13 the arrogance of the management towards their employees, 14 that because of nuclear safety concerns that came along, i

15 the site began to degrade.

16 It degraded in steps. In 1987, it was a 17 world-class nuclear power plant. And there were a lot of 18 proud people working there. And certainly -- and 19 incidentally, there are a lot of proud people working 20 there today.

21 Most -- almost all of my friends in my life 22 worked there, the majority of my friends. They're highly 23 specialized, highly technical, highly educated people, and 24 they're real professionals.

25 But the problem is when -- when this NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS 1323 RHODE ISLAND AVE., N.W.

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~ , ., 53 t:

. management regima came in, the new management regime came 2 in, they began to drag this site down with them. And they

, 3 :reated -- they took a world-class performing nuclear l 1 4 power plant -- it became a level two plant.  !

l 5 And then recently, it became a' level three, 1

. 1 6 troubled plant. So, this -- this wasn't a situation that I l

1 7 existed for my whole career. And I have been very proud i

8 to say that I worked for Northeast Utilities for many of l

9 those years.

1 10 MS, LUXTON: Ma'am, I think you're under the j <

11 mis -- the misimpression that Gary didn't move. Gary got 12 terminated. He didn't lay himself off. They --

I 13 FEMALE PHONE CALLER: I understand that. I l l 14 MS. LUXTON: Okay, okay.

1 j 15 MR. VERDUN: Does that answer your question, 16 ma'am?

j 17 . FEMALE PHONE CALLER: I guess so, thanks.

1

! 18 MR. VERDUN: You're welcome.

19 MS. LUXTON: Okay, we have five minutes left, i

20 Okay, there was another article in the paper today saying l

21 that we've got a full commission now, Gary. Did you know l l

1 22 that?

2 23 MR. VERDUN: Yes.

24 MS. LUXTON: The NRC now has two new 25 commissioners.

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^@ .

y f-1 MR. VERDUN: I read that in the paper this 4

2 morning.

3 MS. LUXTON: Both -- they -- they have to be )

..f.rmed by the Senate, but this is good I think.

4 l

.i 5 MR. VERDUN: Yes. l 6 MS. LUXTON: The more commissioners, the 4

- 7 better.

8 MR. VERDUN: Yes.

9 MS. LUXTON: I mean, now they're working with r

i 10 a full board.

11 MR. VERDUN: That's right.

12 MS. LUXTON: So, that's good.

13 MR. VERDUN: They now have a full deck again.

14 MS. LUXTON: Full deck. Hey, we didn't talk 15 about --

hey, let's talk about that funny thing that we 16 were going to talk about, Gary, that I had forgotten about 17 the --

18 (Phone ringing).

19 MS. LUXTON: -- the Northville Mountain Team -

20 - MR. VERDUN Oh, yes.

21 MS. LUXTON: -- the team building fiasco.

22 MR. VERDUN: Yes.

23 MS. LUXTON: That's funny. Okay, hi, you're 24 on Nuclear Safety Issues.

25 FEMALE PHONE CALLER: Yes, I have a real NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N W.

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I

- __ __ ________4

~ ..,

v

.l; serious concern here that I would love to know how you l::uld be such experts on this if you can't even pronounce

.I It's not "nucular," and that really 3 q".u ~. ear" properly.

4 gets to me for you people to come across as if you're 5 experts. Do consider practicing saying " nuclear."

i 6 MS. LUXTON: Ma'am?

7 FEMALE PHONE CALLER: Yes. '

,i 8 MS. LUXTON: I never raid I was an expert.

4 9 FEMALE PHONE CALLER: Well, you're sure 10 sounding like you're talking about you have all the best i

1 11 interests of all of us in our hearts.

12 MS. LUXTON: Well, I -- I have my own best ,

e 13 interest at heart and my family's best interests at heart l

14 and I hope --

1 15 FEMALE PHONE' CALLER: We11 -- )

i i 16 MS. LUXTON: -- you do too.

i 17 FEMALE PHONE CALLER: -- it's very interesting i

18 when all these discussio1s with people who have created a ,

1 19 situation where the industry has to lay people off is now 20 shocked that they're laying them off. f 21 So I mean, let's -- you're creating something. j i

22 Let's deal with it, and not get too uptight about it. But 23 do please practice saying " nuclear," would you?

24 MR. VERDUN: Thank you for your call, ma'am.

25 FEMALE PHONE CALLER: Do you think you can NEAL R. GROSS COURT REPORTERS AND TRANSCRISERS  !

1323 RHODE ISLAND AVE., N.W.

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• v bo I

1 handle that? " Nuclear" is the way ic's pronounced, not 2 "nucular."

3 MS. LUXTON: Thank you, ma'am.

4 FEMALE PHONE CALLER: Thank you.

5 MR. VERDUN: Thank you for your call, ma'am.

6 MS. LUXTON: Hmmm, that was interesting. Are 7 we the disgruntled employees that made the situation?

8 MR. VERDUN: No, I -- I.was never a 9 disgruntled employee. I was a loyal employee for the 10 company. I did my job. I played by the rules. I was 11 terminated by the company.

12 Now, I'm an -- I'm a former employee. I'm 13 here to tell you things that people that work at Northeast I

14 Utilities are real'ty not free to say without fear of 15 reprisal.

16 (Phone ringing).

17 MS. LUXTON: Hi, you're on Nuclear Safety 18 Issues.

19 FEMALE PHONE CALLER: Hi, I've been told to 20 hurry, so I will. But ever since the day President Carter 21 pronounced it "nucular," I've been thinking that " nuclear" 22 was the proper pronunciation. But I looked it up in the 23 dictionary, and both pronunciations are accepted.

24 So, even though you don't know your " nuclear" 25 stuff --

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m  :.

r i MR. VERDUN:  ; Laughing).

2 FEMALE PHONE CALLER: -- you are pronouncing i

3 i- correctly, i

4 MS. LUXTON: Well, thank you, ma'am.

5 MR. VERDUN: Thank you very much, ma'am.

6 FEMALE PHONE CALLER: Bye-bye.

7 MS. LUXTON: Bye.

8 MR. VERDUN: Was that the same lady?

9 MS. LUXTON: No, that was a different lady.

10 MR. VERDUN: Oh, okay.

11 MS. LUXTON: But --

12 (Dial tone).

13 MS. LUXTON: -- I don't care how I pronounce 14 it, to tell you the truth. The point is " nuclear,"

\

15 "nucular," I could have a little twang in it. I might i

16 come from Texas and say "auclea" or someplace like that.

17 MR. VERDUN: (Laughing).

1 18 MS. LUXTON: The point is you understand the  !

19 word I'm saying I hope, o<ay? Northville Mountain, we're 20 not going to get a enance to talk about that.

21 MR. VERDUN: Well, we --

22 MS. LUXTON: We have about one minute left.

23 MR. VERDUN: Oh, one minute?

24 MS. LUXTON: Right, Mary, one minute left?

25 Yes, we won't be able to talk about it. We're going to NEAL R. GROSS COURT REPORTERS AND TRANSCRl8ERS 1323 RHCDE ISLAND AVE., NW.

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$ w , 55 1- have to --

2 MR. VERDUN: Well, we can talk about that on 3 another show.

4 MS. LUXTON: We can -- right, right. Okay, I 5 want to just give a couple -- an update of a couple of 6 things that are coming up.

7 (Phone ringing). (

8 MS. LUXTON: Why are you giving me another 9 call if we only have one minute left?

10 (Dial tone).

11 MS. LUXTON: Hi, this has to be quick if you 12 want to have a question.

j 13 MALE PHONE CALLER: Yes, I have a beautiful 14 request to you and to the fellow next to you.

• 5

. MS. LUXTON: 'i e s . I 16 MALE PHONE CALLER: Why do you have so much i l

17 talk and the -- no do the job once and for all? You do it )

18 the -- in this situation, : hey cannot operate anymore at 19 this particular nuclear lo:ation plant. So, I hope they 20 do something now, not too much talk. I 21 I read that paper, I read the -- the subject.

I 22 I read the - . They still hire these people. Why they j 23 not be prosecuted?

i 24 MS. LUXTON: Thank you, sir. I agree 4 25 completely. Bye.

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.' % 4 p'-

5)'

(l

". MALE PHCNE CA*.*.ER: Bye.

I I

2 MS. LUX'"CN : Ray, folks, meeting com:.ng up on l

3l September lith, Dr. Jay 30uld, Jordan Firehouse at 7:30.

l We're going to be talking It's a CRC lecture series.

4l ,

5 about cancer in our community and around nuclear power 6 plants. Thank you, Gary, for being here.

7 MR. VERDUN: Thank you very much, Susan.

8 MS. LUXTON: We have -- thanks so much.

9 MR. VERDUN: It's a pleasure to be here.

10 fiS . LUXTON: Come back again, okay?

11 MR. VERDUN: I think you'do a very good 12 ' service for this community.

13 MS. LUXTON: Thanks'.

14 MR. VERDUN: It's a pleasure.

15 MS. LUXTON: You too, thanks.

I 16 MR. VERDUN: Thank you.  ;

i 17 MS. LUXTON: See you next week, folks.

19 (Closing music.)

13 (End of Tape 1, Side B.)

20 i

l i

I 23 24 1

25 j 1

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. $m b '

\

' I CERTI FICATE l l

This is :o . r. fy that the attaches proceedings before the United States Nuclear

. Regulatory Commission in the matter of:

Name of Proceeding: Connecticut Public ~:. vision '

Broadcast re Nuclear Safety Issues on August 2*, *3s6 .

Docket Number: (not applicable)

Place of Proceeding: Television Broadcast were held as herein appears, and that this is the criginal transcript : ne r .3 : ::r ne file of the Un:.ted Stnes Nuclear Regulatory Commission taken by me and, thereafter reduced to typewriting by me or under the direction of the court reporting company, and that the transcript is a true and accurate record of the foregoing proceedings.

o I. J6

'j *auri Ward i

e l

i l

l l

1 f '

NEAL R. GROSS COURT REPORTERS AND TRANSCRI6(R$

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k "~i"A mweg hY 4E' ShkV tu U ',s e u i EDO Principal' Correspondence Control f

f FROMs DUE: 12 //J1-/96 EDO CONTROL: G960919 DOC DT: 11/25/96 1 FINAL REPLY: }

D borch Booth Katz o1/'21/f7 l Citizens Awareness Network '

TO:  !

l Emile Julian, DSB FOR SIGNATURE OF : ** GRN ** CRC NO: 96-1189 [

I I

DESC ROUTING: )

i 2.206 PETITION ON CONNECTICUT YANKEE, MILLSTONE Taylor UNITS 1, 2, AND 3 l Milhoan i Thompson i Blaha  ;

Miraglia, NRR j HMiller, RI i DATE: 12/06/96 Lieberman, OE '

ASSIGNED TO: CONTACT: l t

OGC Cyr  !

l SPECIAL INSTRUCTIONS OR REMARKS: ACT!ON I

I FPP RECEIVED: DECEMBER 16, 1996  !

NPR ACTION: SPO: TRAVERS 0.' " T f' ~tr:s . . s - , u h-

'- iu:n a v4c,,r y; -

i MR 00ljTING: MIRAGl.IA / .

l THADANI ZIMPERMAN l

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L/ }) g] -/

'~-

yg- '

SHERON  !

TRAVERS MARTIN h i B0HRER OGC 005082 O[8 i7h N l C .___ ._ . -

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OFFICE OF THE SECRETARY "3 CORRESPONDENCE CONTROL TICKET PAPER NUMBER: CRC-96-1189 LOGGING DATE: Nov 29 96 ACTION OFFICE: EDO AUTHOR: DEBORAH KATZ AFFILIATION: MAINE ADDRESSEE: EMILE JULIAN, SECY LETTER DATE: Nov 25 96 FILE CODE: ID&R 5 MILLSTONE

SUBJECT:

10 CFR 2.206 PETITION ON CONNECTICUT YANKEE, MILLSTONE UNITS 1,2 &3 ACTION: Appropriate DISTRIBUTION: CHAIRMAN, COMRS, OGC, D&SB SPECIAL HANDLING: NONE

- CONSTITUENT:

NOTES: 2.206 PETITION-- 2 EI! CLOSED VHS TAPES TO EDO j DATE DUE:

SIGNATURE: . DATE SIGNED: )

AFFILIATION:

l I

l EDO -- G960919 k- ___ _ _ _ _----------- a