IR 05000443/1990016

Insp Rept 50-443/90-16 on 900730-0904.No Violations Noted. Major Areas Inspected:Operations,Radiological Controls, Maint/Surveillance,Security,Engineering/Technical Support, Safety Assessment & Quality Verification

ML20059L887
Person / Time
Site:	Seabrook
Issue date:	09/21/1990
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20059L886	List: IR 05000443/1990016 ML20059L883
References
50-443-90-16, NUDOCS 9010020116
Download: ML20059L887 (27)

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J 0.S. NUCLEAR REGULATORY COMMISSION i'

REGION I

-Docket / Report No.: _.50_443/90-16 License No.: NPF-56 .

Licensee: Public Service Company of New Hampshire  ;

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Facility: Seabrook Station, Seabrook, New Hampshire

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. Dates: July 30 - September 4,1990 '!

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! Inspectors: 'N. Dudley, Senior Resident Inspector .

J R. Fuhrmeister, Resident Inspector.

[iL A. Cerne,' Senior Resident Inspector, Construction  ;

P. Sena, Reactor Engineer l

l Approved By: & D.~ b M .h Ebe C. McCabe, Chief, Reactor Projects SectioT TB

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f OVERVIEW Operations: The loss of offsite power test and the 250-hour warranty run were Ip safely completed.' -An emergency power reduction, a. turbine setback and an un-

- planned reactor trip recovery were appropriately controlled.

! Radiological Centrols: Contaminated areas were identified, controlled and de-

contaminated. Continuing problems with controlling access to locked high radiation o- Jareas were identifie Maintenance / Surveillance

Work was generally well-controlle Maintenance and:

surveillance activities resulted in_a reactor trip and a turbine setbac Securi ty.: Licensee investigation of the work of individuals identified as sub-t

"' stance. abusers identified no adverse affect on plant safet . Engineering / Technical Support: Engineering analysis of the P-9 setpoint and-cbarging. pump line vibrations were' acceptabl E iSafety Assessment / Quality Verification: NRC inspection of weld quality records

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i w identified no safety inadequacies, b

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9010020116 900921 PDR ADOCK0500g3

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TABLE OF CONTENTS f '

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PAGE L 1.0 ' Summary of Activities,.................................... 4......... I'

2.0. Operations'(71707,71710,92700,'92701,93702)....................... '2

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2.1 Plant Tours..................................................... 2 2.21-Plant Events........................................... ........ 3 2.3 Engineered Safety Feature System Wa1kdown..........'.............- '4 2.4 (LER 90-014) Blowdown Flash Tank Drain Radiation Monitor Inoperable..................................................... 4 2.5. (LER 90-018) Reactor Trip Due to Low EHC 011 Pressure Signal.... 5 g 2.6' Facility: Tour Issues............................................ 5 b . .

3.0'~ Radiological Controls (71707, 92701)................................. 5

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'3.1' Plant Tours..................................................... 5 L 3.2 ( LER 90-017) High Radiation Area Requi rement Not Met. . . . . . . . . . . . 6 . .0 Maintenance / Surveillance (61726,62703,92701)....................... 6

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4.1~ Maintenance.....-................................................ 6 Surve111ance.............................. ....................- 7 5.0 . Security (71707,92701)..,........................................... 7 16,0 ' Engi neeri ng/ Technical Support (92701) . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . '

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6.1 P-9 Setpoint....................................................- 8 .

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n 6.2 Charging' Pump Vibration.........................................- .9-6.3 (89-09-02) Atmospheri c Steam Dump Val ve Desi gn . . . ... . . . . . . . . . . . . . 10'

6.41 Design of Breathing Air Supp1y................................... 10

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L6.5 Potential MSIV Circuit Deficiency............................... .11

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7.0 Safety Assessment / Quality Verification (92701)....................... 11-

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, 7.1: Reviews of Welding Quality Records.............................-... 11'

7.2' (90-07-01).10 CFR 50.55(e) Evaluation of a YAEC Deficiency Report........................................................ :12

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1 8'.01 Meetings.............................................................,. 13

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DETAILS 1.0 Summary of Activities *

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1.1 NRC Activities '

- Three resident inspectors were assigned. The 150 inspection hours  !

[ included'40 backshift hours, of which 27 were' deep backshift hour '

A team of region-based and headquarters-based inspecto_rs provided l l 24-hour per day coverage of operations until August 8, 1990. Aug-mented site coverage was then provided until September 4.-199 {

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i On July 30 - August 2, 1950, the Special Test Programs Section. Chief I

["j from the Region I Division of Reactor Safety visited the site and i observed power ascension tectin !

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On August 8, NRC Commissioner Curtiss visited Seabrook Station. The- '

Commissioner, accompanied by Mr. W. Kane, Deputy Region I Administrator, toured the f acility and met with plant and corporate managemen ;

L Representatives from the Seacoast Anti-Pollution League:and the New l England, Coalition on Nuclear Pollution participated in the plant tour ;

and meetin Slides used during New Hampshire Yankee's (NHY's) pre- ,

, sentation are attached _to this repor '

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On August 14; the Director, Project Directorate'I-3, NRC Office of  !

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. Nuclear Reactor Regulation (NRR), toured the facility with the resident inspectors and met with the NHY Senior Vice President and Chief of ;

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-On August 17, the Director of NRR toured the facility with the resident !

inspector.and met with the President and Chief Executive Officer and !

,, his staff. The meeting included discussions' of reliability centered j maintenance, corrective.-actions on submerged electrical-cables, and

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.the status of the licensee's check valve testing progra t On August 20-24, there was a region-based inspection of radiological ,

n he results will be documented in NRC-Inspection Report On August 23, she Atomic Safety and Licensing Board reacct-d claims :

that the Vehicular Alert Notification- System is inadequate and. ejected an appeal based on the board violating its own procedure ,

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On August 27-28, two region-based inspectors conducted final rey h s 3 of the power ascension test program. The inspection results will be j,

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r documented in NRC Inspection Report 50-443/90-8 *

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n 2 On August 2B-29, the Deputy Executive Director for Nuclear Reactor

[t Regulation, Regional Operation and Research and the authors of NUREG-1425

" Welding and Nondestructive Examination Issues at Seabrook Nuclear .

Station," met at Seabrook Station with congressional aides to discuss

.NUREG-142 .2 Plant Activities At the beginning of the inspection, the plant was in Mode 3, Hot F

Standby, following completion of a planned reactor trip from 100%

> power and the Natural Circulation Test. The reactor was taken cri-tical on July 30 and power was raised to 30%.

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On August 1, the plant was tripped from 20% power as part of the planned Loss of Offsite Power Test. The reactor was taken critical on August

3, and power was raised to 100% to perform a 250-hour warranty run.

On August 13, reactor power was reduced to 15% to repair an electro-hydraulic control (EHC) oil leak in the turbine buildin Power was returned to 100% on August 1 On August 17, the warranty run was completed. The station was declared !

to be in commercial operation on August 19, 199 :

On August 22, a reactor trip resulted from troubleshooting performed ,

in the turbine EHC cabinet. The plant was restarted on August 2 !

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On August 25, a setback of the turbine reduced turbine power to 55%. !

The setback signal was produced by improper shif ting of iso phase bus !

cooling unit fans.- Power was restored to 100% the same da l

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2'. 0 ' Operations

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2.1 Plant Tours The inspector conducted daily control room tours which included reviews of operator log-books, technical specification action statement tracking a logs, tagout logs and night orders. Assessments were made of technical specification action statements in effect, control room staffing, management oversight, operator awareness of plant conditions and alarms, and operator' responses to abnormal events. No unacceptable condition were note ,

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The inspector conducted plant tours which included the primary auxiliary b'uilding, fuel handling building, waste processing building, turbine building, switchgear rooms, diesel generator building, service water -

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building, cooling towers and intake structures. No equipment or struc- l tural problems were identified. Minor discrepancies were turned over to the licensee and adequately resolved, e

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2.2 Plant Events

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The inspector witnessed the loss of offsite power tes The crew was well-trained and effective communications were maintained throughout the test. Plant equipment responded as designed with the exception e of Group A pressurizer backup heaters. New Hampshire Yankee determined that the heaters were incorrectly wired because information from the electrical prints were incorrectly transferred to wiring sheets. The

! wiring sheets were corrected and the heater group was rewired properl The wiring sheets for the other pressurizer heater group was verified by the licensee to be correct. The inspector verified that the other heaters functioned as designed during the tes On August 22,-troubleshooting was conducted on the circuit card for E the Early Valve Actuation System, which provides rapid closure of intercept valves on a sensed mechanical to electrical load imbalance

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on the main turbine. When jumpers were installed to repeat a previously performed test the turbine tripped due to a voltage drop on the 24V DC control circuit power supply. The plant was stabilized in Mode The feedwater isolation signal could not.be reset because an erroneous high steam generator water level signal was produced. The reactor trip breakers were closed to allow resetting of the isolation signa Also, relief valves on the high pressure feedwater heaters lifted and did not reseat, Feedwater to the heaters was isolated and the relief -

valves were repaire During the emergency shutdown of the turbine generator on August 13, the operators were unable to maintain control rods in the required ban The operators reduced power with rod controls in manual and over-borated the primary coolant. When power was reduced below 20%

prior.to manually tripping the turbine, the operators entered the action statement for Technical Specification 3.1.1.3a, which defines maximum withdrawal limits for control rods When the moderator tem-perature coefficient is positive. Sufficient demineralized water was added to the primary coolant to allow the rods to be driven below th withdrawal limits. A " lessons learned" review of the event determined that the control rods should have been left in automatic and a limited amount of boron injected to initiate the power reduction. The inspector

observed other crews effectively employing that technique to achieve l more controlled decreases in later power reduction Operator response to abnormal events was generally conservative and in accordance with required procedures. No regulatory concerns were noted. However, even though information on plant response is provided through operating experience sheets and the night order book, the inspector observed some examples where available information on plant response was not applied on shift. For example, following the reactor trip on August 22, the plant responded as it had during the planned trip from 100% power during the Power Ascension Test Program. Operators stabilized the plant, However, the observed -4 degrees Tave-Tref

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signal caused by isolation of the turbine and feedwater heaters on i the turbine trip was not expected by the Unit Shif t Supervisor. The  !

Shift Superintendent did not understand that the feedwater isolation 'l signal could not be reset because an erroneous high' steam generator 1

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water level signal was produced and was not indicated in the main  :

control room. The operators unnecessarily isolated the steam dumps '

when control was transferred to the pressure control mode, since they "

were unaware that the integral function of the pressure controller .i should be zeroed prior to the transfer, j As a result of the post-trip review, the unexpected plant responses were detailed in the night orders. Also, results of power ascension  :

tests and unexpected plant transients were incorporated in the operator  !

qualification program. The inspector had no further question .3 Engineered Safety Feature System Walkdown f

Operability of the Train "A" Diesel Generator Cooling Water System i was verified by performing a walkdown of the system using PID drawin DG-B20461. The walkdown was conducted to identify conditions that: '

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could degrade performance and to' verify that valves were properly positioned. The inspector noted that the jacket coolant standby cir-culating system was operating properly'as part of the engine's keep-warm system, maintaining jacket coolant temperature at 110F. The ,

inspector. verified that the auxiliary coolant pump and the mechanically '

-interlocked suction and discharge valves were properly aligned as a  ;

backup source of coolant circulation. No deficiencies were note .

2.4. LER 90-014: Noncompliance with Technical Specification - Blowdown Flash Tank Drain Radiation Monitor Inoperable l This event occurred on June 12, 1990, when it was discovered that the

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i steam generator (SG) Blowdown Flash Tank Drain Radiation Monitor-Sample  :

pt..np was not running. The radiation-monitor. control panel indicated- t flow through the monitor. This condition was discovered by NHY. while investigating low flow indications in 2 of the 4 individual SG blowdown ,

radiation monitors. It was determined that the flow sensing switch ,

was stuck due to an accumulation of solids, The accumulation resulted '

from high suspended solids in the SG blowdown lines due to the breaking j free of scale and corrosion products during power ascension (an expected . i condition). Any release would have been detected by the individual >

SG blowdown radiation monitors and the routine daily chemistry sample Corrective action. consisted of operators performing a daily flow veri-  ;

fication using the sight glasses. In addition, the flow switch is being' evaluated by NHY for suitability for use in flow streams con-taining high suspended solids. This item is close ,

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' r5 LER 90-018: Reactor Trip Oue to Low Electrohydraulic Control Oil Pressure Signal r

This event occurred on July 5,1990 when the unit tripped from 75%

l power due to an erroneous low Electrohydraulic Control (EHC) Oil Pres-sure Si_gnal. The false low pressure signal was due to pressure sensing 1 '

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switch contact closure resulting from high vibration of the switche '

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The switches were supplied as part of the turbine control system and ,

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were_ mounted on the turbine stop valves. Steam flow through the valves

caused vibration which was conducted to the switches causing contact !

- bounce. No precursor of this vibration effect was note Corrective <

action consisted of removing the switches from the valve bodies and b rigidly mounting them to turbine building structural steel member This issue is close .6 Facility Tour Issues  !

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Observations made during a tour of the facility on January 3,1990 were documented in Inspection Report 50-443/89-2 l The status of !

several of those items is given below: .i

< Item 3 - Secondary plant areas need cleaning up: completed.

T u L Item 6 - RM6510 and 6511 have broken glass on flow elements (WR90- !

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0000018): The inspector verified that the sight glasses have been replaced. In addition, radiation monitors have been placed in service i'

.and flow indication is functioning properl f Item 11 - Evaluate removing the paint on the latching mechanism for ;

.the turbine-driven EFW (emergency feedwater) pump: The inspector >

verifed that paint had been removed from the mating surfaces.of the ,

. latch for the mechanical overspeed tri '3.0' Radiological Controls '

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' Plant Tours r The inspector toured the radiological controlled area, reviewed posted survey maps, and observed radiological practices. .No discrepancies -

were note ;

Contamination was identified in the "A" charging pump room and a con-tamination area was established. The contamination was verified to-be from fission products. During a two week investigation, health physics technicians and management prevented the spread of contamina-tion, attempted unsuccessfully to identify the source of contamination, ;

decontaminated the area, and restored the area to general acces The inspector had no further questions on the response to the unexpected contaminatio >

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3.2 .(Closed) Licensee Event Report (LER) 90-017: Noncompliance with Technical Specifications - High Radiation Area Requirement Not Met This' event was discussed in NRC Inspection Report 50-443/90-15. The inspector subsequently verified that signs were placed at key card readers and doors for locked high radiation areas, that the number of ,

persons permanently allowed access to locked high radiation areas was -

reduced from 112 to 23 and that a training deficiency report was issue This item is close '

u After implementation of corrective actions, a locked high radiation I area door was found unlatched by New Hampshire Yanke The event and i follow-up actions are discussed in NRC Inspection Report 50-443/90-1 '

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4.0 Maintenance / Surveillance -

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4.1 Maintenance On August 22, a turbine trip / reactor trip occur ed during troubleshooting i,

of the Early Valve Actuation (EVA) system. NHY was unable to reproduce the fault that caused the tri ,

The inspector monitored troubleshooting activities on the EVA circuitry following the trip. Troubleshooting was conducted in accordance with ,

WR 90 WOO 3810. Operations personnel were briefed on the maintenance and were aware of possible system responses. Technicians monitored 24V DC trip circuit bus voltage while simulating a loss of the permanent  ;

magnet generator power supply. Bus voltage remained at 24V DC as the

? in-house power supply properly assumed.the load. The technicians had the work request in hand while performing the maintenance and had

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adequate supervision and technical assistance. The technical manual was available at the work site and was referenced during the wor Troubleshooting activities.and results were documer.ted en the work i request. However, the inspector noted that the level of detail describing the troubleshooting was insufficient to reconstruct all the steps ,

taken by the I&C technicians. New Hampshire Yankee agreed that docu-  ;

i mentation of troubleshooting needed improvemen During the trip, the thermal relief valve on High Pressure Feedwater-Heater 26B lifted and failed to reset. Work on this valve, 1-MVD-V132, ,

was. conducted in accordance with WR 90W00443 Prior to work commence-~ a ment, the inspector verified that safety tags were in place, assuring

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that the work on the relief valve could be performed safely. Th .

inspector witnessed the removal of the valve from service, initial ,

bench testing of the valve and its subsequent disassembly ai.o inspectio i Repair was conducted in accordance with approved procedures. The '

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inspector reviewed the results of the valve's retest and verified

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that the lift pressure setpoint and seat leak tightness met acceptance i criteria.

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The inspector had noted end documented, in IR 50-443/89-20, a head gasket leak on the'#9 cylinder of Diesel Generator.1A. Inspector ;

monitoring during ensuing months confirmed that the leak had stopped, t During the loss of offsite power test with the engines loaded, the inspector again confirmed that the head gasket leak had stopped without corrective maintenance being necessary. This concern is resolve .2 Surveillance +

i On August 25, a turbine setback was initiated when improper shifting i of iso-phase cooling fans occurred. The potential for a turbine setback was identified by the technician based on previous events on similar systems and therefore the Unit Shift Supervisor observed the surveil- l

lance test. With one fan running, the second fan control switch was !

taken to " start." Unknown to the operators, the control switch operates a damper which takes 30 seconds to open before automatically starting the fan. The first fan's selector switch was taken from automatic to

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manua That, unknown to the operators, produced a momentary " fan not. running" signal. With two " fan not running" signals, a turbine setback was initiated. New Hampshire Yankee revised the surveillance procedure by adding a caution to ensure that the second fan had started before transferring the first fan to manual control. Also, all non-technical specification surveillances which could cause a trip or setback were suspended until a review of the adequacy of the procedures was complete *

The inspector observed two performances of Surveillance Tests 1431.02, !

03 and 04, " Valve Testing for TG Stop, Control and Intercept Valves."

The sur esiliance tests were completed successfully. The tests were coordinated with station engineers and GE contract personnel to evaluate i vibrations on the EHC oil lines. As a result, additional pipe hangers-were permanently installed on the EHC oil lines,

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5.0 Security The inspector observed routine security activities including personnel and vehicle searches, responses to intrusion alarms, and compensatory measures for work under and along the protected area fence. No discrepancies were '

note The inspector reviewed New Hampshire Yankee's investigations of the work performed by three individuals who were suspended for substance abus The inspector concluded that adequate investigations were conducted in leach case and that plant safety had not been degraded. Details follo One. investigation included a review of work conducted by a supervisor who had access to vital areas. That review identified four quality-related routine task sheets for which the individual had been responsible. The i equipment worked on was walked down and no deficiencies were noted. The

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I employee allegation referral program conducted a review of its files and a i a canine search was performed of the Individual's work area. No safety problem l t was identifie >

In another case, the licensee determined that the individual involved did I- not have security access to vital areas. This individual had been in trainhg i r for the previous two months and was under constant supervision. The indi- ' '

l vidual's work performed for the two months prior to his training was reviewe A sample of Ste of his tasks was evaluated and determined to have been properly  ;

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performed. No problems affecting plant safety-related equipment were iden-  !

i tified, although one common documentation error was identifie .

In the third case, the individual did not have access to vital areas and  !

k did not perform work on safety-related equipmen . 0 Engineering / Technical Support 6.1 P-9 Setpoint f On June 20, 1990, while in Mode 1 at 30*4 power, a turbine. trip / reactor

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trip occurred due to actuation of a main generator station ground ,

L fault relay. -The power level.at which the reactor directly trips as  !

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. a result of a turbine trip is determined by the."P-9" interlock setpoin ~

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Westinghouse Standard Technical Specifications (WCAP-12159, Merits Program - Technical Specifications and Bases, March 1989) provides the basis for the setpoint and indicates'that the P-9 setpoint is intended to be selected on the basis of load reject capabilit *

A 50?4 setpoint for P-9 was approved for Seabrook in the March 1983 NRC Safety Evaluation Report (SER), Sections 7.1.5.2 and 7.2.1. This is consistent with the basis in the Standard Technical'Spectiication '

Subsequent to issuance of the'SER, the licensee issued FSAR Amendment ,

56, changing the P-9 setpoint; from 50*; to 20% power. The setpoint  :'

change is reflected.in Technical Specification 2.2.1, which allows a P-9 setpoint of from 0?4 to 22.1% of rated thermal powe *

Technical Specification 5.7.1 (component fatigue limits) states that reactor coolant system components were designed and shall be limited, l over their life, to 80 loss-of-load events without immediate reactor -

trip from power levels greater than 15*4, and to a total of 400 trips, l The P-9 setpoint determines which of these two limits is more likely to be approached later in lif The licensee provided background information relating to the setpoint .i change including two 1983 memoranda prepared by the licensee's Transient -

Analysis Group. Using.RETRAN computer program analyses for various power levels, the effect of the P-9 setpoint was examined in detai As a result, the licensee (a) lowered the P-9 setpoint from 50% to 20*4 and (b) raised the pressurizer power-operated relief valve setpoint i

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to 2449 psia, which is above the high pressure reactor trip setpoint.

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These changes were expected to reduce challenges to relief and safety valves for turbine generator trips with loss of steam dump capabilit .' Plant procedures specify that the Rod Control' System (RCS) normally l be placed in autonatic at 15'o power. However, the RETRAN analyses '

upon which the P-9 change decision was based assumed, more conserva-tively, that the RCS is in manual and no operator action mitigates the transien The lower P-9 setpoint may increase the number of automatic reactor trip That will be routinely monitored by the licensee and routine NRC inspectio Also, primary and secondary side relief valve challenges ;

may be reduced for some turbine generator trips that occur coincident ~

with loss of main condenser steam dump capability. Overall, the effect ,

on safety of the P-9 change was assessed as minor and the inspector concluded that the basis for reducing the P-9 setpoint was acceptabl .

6.2, Charging Pump Pipe Vibration To resolve concerns about vibrations noted on the positive displacement ,

charging pump discharge line during a plant tour, the inspector reviewed '

Request _ for Engineering Services (RES)90-145, " Evaluation of Pipe '

Vibration Associated with Positive Displacement Charging Pump,1-CS-P- ,

128." The RES addressed both the discharge line vibrations and the

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i root cause of the pump drain line failure which was-identified on 5 March 6, 199 '

Vibration measurements taken on the 3-inch charging line indicated that the highest reading was 66 mils in the east-west direction at a flow rate of 75 gpm. Calculation SBC-379, " Steady State-Vibration in ,

Line CS-355-1-2501-3, Charging Pump C5-P-128 Discharge Line," calculated

, an acceptable limit of 156 mils. peak to peak. The inspector concluded that the measured vibrations were well within the calculated-accept-ability limit The root cause analysis of the 3/4-inch drain line weld failures de-termined that welds F0118 and F0120 failed due to fatigue. Therefore, even though RES90-145 concluded that the measured vibration was within 90'4 of the calculated acceptable limits, a pipe support was added per minor modification MM00 90-56 The failed welds were examined by Pullman-Higgins in November 1984 using a visual test and liquid penetrant examination. Atomic Nuclear Insurers later performed a surveillance of the liquid penetrant examinatio No nonconformances were written against the line or weld. A metallurgical analysis of the welds found minor weld imper-fections, caused by poor workmanship, which provided the initiating sites for the fatigue failure n

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The Nuclear Quality Group (NQG) performed an independent assessment of the weld failures. The inspector reviewed the report and noted that an incorrect minor modification number was referenced. The licensee corrected the-error. A single welder-performed all of the welding on the failed welds. A check of that welder's qualification records established that the welder was adequately qualified. The orientation of the socket welds which interfered with the welding was unique to the system. Nonetheless, configurations of other unsupported drain lines with isolation valves attached were modified to add additional support NQG concluded that the weld failures c'id not have generic implications due to the unique orientation of the pipe and the expecta-tion that similar failures would be identified by pin hole leaks 'versus-catastrophic failure ,

The inspector concluded: that an adequate evaluation was conducted'of the vibration on the discharge line and of the generic implications of the drain line weld failure; that the calculations and metallurgical -'

analysis were comprehensive and adequately documented; and that the i NQG evaluation of the drain line socket weld failures was complete and reached acceptable conclusions.

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6.3 { Closed) Unresolved Item 89-09-02: Atmospheric Steam Dump Valve (ASDV) Design

The adequacy of actions taken to verify operability of atmospheric steam dumps valves (ASDVs) was reviewed in NRC Inspection Report 50-443/90-08 and determined to be incomplete based on lack of monitoring of the characteristics of the ASDVs under dynamic conditions. The '

licensee implemented quarterly dynamic testing using procedure OX 1430.02, Revision 4 ,which strokes the ASDVs 20% under load. The procedure became effective on August 8, 1990. The inspector reviewed- ;

the procedure and concluded that an adequate test program was in place ;

o to monitor ASDV performanc This item-is close l4 Design _of Breathing Air Supply

To determine the potential for inadvertent use of pure nitrogen in breathing air supply systems, the inspector reviewed the design and operation of the system. Air breathing carts can be connected to the service air system. The carts monitor air quality, including oxygen, and are required to be continually monitored while in servic Service air and instrument air (IA) systems are supplied from the same air receivers. The service air system is automatically isolated from the instrument air system and the air receivers when IA system press'ure is low. Nitrogen supply bottles are used as backup to in-strument air for safety-related vaives ani '

. separated from the IA header by check valve ,

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< ji Nitrogen could inadvertently be used for breathing air if IA~ dropped below the pressure of the individual valve nitrogen bottle backup A '

supplies, check valves in the IA lines to the individual valves leaked-t by, the automatic service air isolation valves failed to isolated, there was sufficient volume in the nitrogen bottles to fill the IA s and service air headers, and the individual monitoring the air breathing i

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cart failed to respond to the reduced service air system pressure or low oxygen content. The inspector concluded that this presented no

significant potential for nitrogen to enter the breathing air system

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at Seabrook, In_considering whether partial pressure equalization of-oxygen between the air systems and nitrogen bottbs could-result in-an inadequate oxygen supply, the inspector assested the air quality-

assessment feature as an adequate safeguard and had no further question .5. Potential MSIV Circuit Deficiency Due to identified deficiencies at another facility, the inspector reviewed the logic of the steam line isolation circuitry and verified

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that the circuit contained an actuation' signal seal-in feature. The actuatioa circuitry was provided by the NSS$ vendor, Westinghouse, and no plant unique modifications were made to the circuit. The in-

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spector concluded that the steam line isolation circuit met the speci-fications of IEEE Standard 279-1971, 7.0 Safety AssessmentL/0uality verification

' Reviews of Welding Quality Records As documented in previous NRC Region I Inspection Reports, additional inspection of weld quality for piping systems was initiated as a result

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of Congressional interest. During this period, NRC resident inspecto review of construction' records' continued. . Region I also provided, support and assistance to'an NRC Independen.t Review Team (IRT). The results of IRT review of welding and Non-Destructive Examination (NDE)-

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at Seabrook are documented in NUREG-1425, published in July-199 In addition to providing support to the Independent Review Team, the resident inspectors separately assessed the acceptability of licensee record controls, retrieva'ility, o and quality verification for the welding /NDE areas of interest. Construction procedures were reviewed,

-quality assurance and engineering personnel were interviewed, and quality documents were examined to evaluate the sampled construction processes for documented verifiability of quality from a historical standpoint. During this report period,-record reviews and quality verification inspections related to the Congressional staff questions submitted prior to September 4, 1990 were completed. Resident inspector review during this inspection produced no evidence of defective work, improper process controls, incomplete records, or unresolved safety question )

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e 7.2 (Closed) Unresolved Item 90-07-01: 10 CFR 50.55(e), Evaluation of

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a YA T Deficiency Report j

As documented in NRC Region I Inspection Report 50-443/90-07, the licensee was requested to demonstrate that the contents of Deficiency Report (DR) 527 had been properly evaluated in accordance with the criteria of 10 CFR 50.55(e). This DR dealt with radiographic film package discrepancies identified by Yankee Atomic Electric-Company (YAEC) QA film reviewers during a-1983 review of pipe weld radiographs -

i transmitted for final turnover by Pullman-Higgins, the piping contractor, r Subsequent to the identification of this unresolved item regarding-the reportability of DR 527, a YAEC RT INTERPRETATIDN list enumerating ,

the specific welds categorized on the DR was found and evaluate '

This listing showed that none of the documented discrepancies were .

): weld quality' rejects requiring weld repai If a weld quality defect -

had been identified, the issuance of a nonconformance report (NCR).

was required in accordance with Pullman-Higgins Procedure XV-2. All  ;

such NCRs received-their own evaluation for reportability under the 3 criteria of 10 CFR 50.55(e). On March 19, 1990, the licensee issued i a memorandum explaining that an administrative error had resulted in--

the lack'of proper documentation of the reportability evaluation of DR 527, but that re-review concluded that the DR did not meet 10 CFR "

50.55(e) tests of reportabilit This memorandum further explained _ J

, the programnatic requirement that any identified hardware deficiencies  ;

i' would have resulted in an NCR which would have been reviewed-for re- ,i portability,

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> .NRC inspection confirmed that no- NCRs were initiated by Pullman-Higgins -

as a direct result of the discrepancies categorized on DR 527.. This-supported the-position that no hardware problems were evident in the' ,

DR deficiencies. An NRC inspector reviewed the applicable YAEC and: .

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Pullman-Higgins procedures and evaluated the licensee's explanation  !

and-reportability determination documented in the memorandum of March *

12, 1990. A records review found a YAEC Controlled Speed Letter'(CSL

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1. 095), dated five days after DR 527 was issued, forwarding the subject l DR to the YAEC Project Engineering Manager for 10 CFR LO.55(e) review .

and subsequently documenting closure of DR 527 as of January 10, 198 '

n NRC review of this issue concluded that, although cohfirmation of the results of a 10 CFR 50.55(e) evaluation of DR 527 do not exist in J licensee quality records, evidence of the submittal of the DR for t

, engineering review are available. Furthermore, current evaluation  ;

of the deficiencies documented on the DR validates the position that ,

D reportability was not required. Additionally, an NRC Independent i Review Team (IRT), established to review and respond to Congressional .

concerns, evaluated the licensee handling of DR 527 and the overall

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10 CFR 50.55(e) program. NUREG-1425 was published in July 1990 to document the findings of that IRT inspectio r (

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, review program and findings are documen ed Lin NUREG-1425, Section 1 IRT inspection findings related to DR 527 in this regard are also

' documented .in NUREG-1425 (p.14-3). ' The. IRT, in considering the NHY memorandum of Marc * 12,1990: and other relevant-information, concluded.:

that a 10 CFR 50.55(e) review of the DR 527 deficiencies would have resulted in a determination that the identified conditions were not'

reportable.-

ff, This unreso'1ved item is closed.

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. 8 .' O Meetinos The scope and findings of the inspection were discussed periodically-

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throughout the. irispection period. . An oral summary of the preliminary in- '.

spection. findings-was provided to the Executive Director of. Nuclear Production

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and his' staff at the conclusion of the inspectio Region-based inspectors conducted the following exit meetines, g Date' Subjec Report N Inspectori 8-24' Radiological Controls 90-18 R..Nimitz

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'8-31- Power; Ascension Test- 90-83 J. Trapp Program

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- Owners & Chief Executive Officer: l-l LE. MAGLATHUN, JR. - i

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' F.W. EDWARDS, CHAIRMAN ,

D 3. PARDUS,VICE CHAIRMAN *

.I Poky, &xiget & Fharckt Ltensing & Lkense tekted naners NHY President

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Power Ascension Test Program i f Startup Test Department '

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Staff System Support Departmert I I I I  :

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Director Director Director Director 'i SHIFT #1 SHIFT #2 SHIFT #3 SHIFT #4

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Test . Test Test Test Directors Diremors Directors Directors ,

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MAJOR TESTS. PERFORMED

ST22 Natural circulation test ST-35 Large load' reduction (2)

ST-38 Unit trip from 100% power (1)

ST 39 Loss of offsite power (1)

ST-33 Shutdown from outside the control room ST-34 - Load swing test (4)

ST-48 Turbine-generator startup test (6)

ST-4 Turbine torsionaltest (1)

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Professionalism Safety Excellence Integrity and Ethic 3 q

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LAREAS OF EVALUATION

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Team Manager 16 -

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Quality Assurance Senior Engineer 10

. Startup Test Director 18

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YANKEE ATOMIC Radiation Protection Manager 29 .

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Project Manager _ 27 1 Ex-NRC Site Section Chief 37-Maintenance Manager 27 ,

Maintenance Manager 19- t Maintenance Supervisor 20

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. MANAGEMENT OVERSIGHT COMMITTEE  :

COMPOSITION 1 i

Edward A. Brown 1

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President & Chief Executive Officer i Ted C. Feigenbaum -

Senior Vice President & Chief Operating Officer ;;

Bruce L Drawbridge -

Executive Director - Nuclear Production l - R. Jeb DeLoach -

Executive Director - Engineering & Ucensing .h

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. Director of Quality Programs

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Station Manager

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