IR 05000354/1986026

Insp Rept 50-354/86-26 on 860501-0609.No Violation Noted. Major Areas Inspected:Operational Safety,Surveillance Testing,Maint,Esf Sys Walkdown,Ler Followup & Biscoseal Deficiency

ML20199F450
Person / Time
Site:	Hope Creek
Issue date:	06/17/1986
From:	Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199F409	List: IR 05000354/1986026 ML20199F403
References
TASK-2.K.3.16, TASK-TM 50-354-86-26, NUDOCS 8606240330
Download: ML20199F450 (29)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

050354-860413 Report N /86-26 050354-860415 050354-860416 Docket 50-354 050354-860417 050354-860420 License NPF-50 050354-860424 050354-860425 Licensee: Public Service Electric and Gas Company 050354-860426 050354-860502 Facility: Hope Creek Generating Station Conducted: May 1 - June 9, 1986 Inspectors: R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector (In Training)

E. L. Conner, Project Engineer R. L. Nimitz, Radiation Specialist R. J. Summers, Project Engineer K. H. Gibson, Reactor Engineer (In Training)

Approved: C-L. Norrholm, Chief, P/ojects Section 2B ' Uate Inspection Summary:

Inspection on May 1, 1986 - June 9, 1986 (Inspection Report Number 50-354/86-26)

Areas Inspected: Routine onsite resident inspection of the following areas: followup on outstanding inspection items, operational safety verification, surveillance testing, maintenance activities, engineered safety feature system walkdown, preoperational testing, licensee event report followup, Biscoseal deficiency, allegation followup, and safety issue surve Information on the management meeting between PSE&G and NRC Region I is also provided in this report and its enclosur This inspection involved 271 hours0.00314 days <br />0.0753 hours <br />4.48082e-4 weeks <br />1.031155e-4 months <br /> by the inspector Results: No violations were identified; however, our review of control rod drive mechanism maintenance activities indicates that an increased focus on procedural adherence and detail is warranted. Although the potential impact on health and safety was minimized because the plant was in a I pre-initial criticality phase we are concerned that a lack of attention to detail in the future could have an adverse effect on plant operations and worker safet Review of this maintenance activity is documented in paragraph 5 of this repor .

8606240330 860618 PDR ADOCK 05000354 Q- PDR

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d DETAILS 1 Persons Contacted

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Within this report period, interviews and discussions were conducted i with members of the licensee nanagement and staff and various contractor personnel as necessary to support inspection activity.

1 Followup on Outstanding Inspection Items

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! Inspector Follow Items l (Closed) Inspector Follow Item (85-45-01); OPS Surveillance /

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Calibration of Temperature Switches on ECCS Room Coolers, Because the ECCS pump room coolers are controlled by room temperdture i switches and not by the applicable pump start control circuit, the i

s inspector developed a number of concerns relating to the calibration of temperature switches and functional testing of room coolers. The inspector verified that the room cooler temperature switches have been calibrated and are specifically identified for a calibration j frequency of 18 months. In response to the inspector's concerns, Operations Procedure OP-FT.ZZ-001(Q) " Emergency Area Coeling Systems j

(EACS Room Coolers Functional Test - 18 Months)" was written to test the auto initiation-of the EACS room coolers and the associated

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Safety Auxiliary Cooling System valves. This procedure was reviewed and found to be acceptable. The inspector has no further questions and this item is close (Closed) Inspector Follow Item (86-02-02); Maintenance Surveillance '

Procedure Update. During a previous inspection, several discrepancies

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were noted between the draft Technical Specifications (T.S.) and mainte-nance department surveillance procedures. In response to the inspector's concerns and as part of the licensee's surveillance procedure review program, a comparison was made between all maintenance department T.S.

surveillance procedures and the the T.S. This review included

the veri-fication of the T.S. reference number, setpoints, periodicity and proper format; and, the revision of data sheet l

This review was completed on April 23, and all necessary' corrections have been incorporated. The inspector independently verified that

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the specific instances cited in Inspection Report 86-02 have been

! corrected. The licensee's procedures used to satisfy T.S. surveillance requirements will continue to be reviewed as part of the routine inspec-tion program. This item is close (Closed) Inspector Follow Item (86-06-04); Various Deficiency Reports.

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The inspector reviewed the disposition of three deficiency reports (DR)- l which had been written to document the inspector's questions. In DR !

HQA-86-005 the valve manufacturer's engineer and the system sponsor engineer determined that sufficient thread engagement existed on HPCI

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valve 1-BJ-HV-F006. No corrective action was required. In response to DR HQA-86-004 a Hope Creek resident engineer verified that sway strut

! 1-P-BC-019-H24 was correctly installed. The missing snap ring was

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installed on hanger 1-P-FD-224-H07 in accordance with DR HQA-86-003. The inspector has no further questions at this time and this item is closed.

(Closed) Inspector Follow Item (86-07-01) Licensee to train and i qualify an adequate number of radiation protection personnel to

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support fuel load, power ascension, and fuel power operation. The i

licensee has trained and qualified an adequate number of radiation protection personnel to support fuel load, power ascension, and full power operatio (Closed) Inspector Follow Item 86-07-04) Licensee to repair the damaged whole body counter. The licensee obtained and placed in >

service a new stand-up whole body counter. Procedures have been established for use of the counter. Minimum detectable activities for the counter are adequate for principal radionuclides of interes (Closed) Inspector Follow Item (86-07-05) Licensee to establish and implement a contamination control program consistent with guidance in IE Circular 81-07, " Control of Radioactive Contaminated Material."

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The licensee revised procedure SA-AP.ZZ-029(Q), " Radioactive Waste and Material Control Program," and other appropriate procedures to

address NRC identified concerns. The procedures include: release criteria; surveys of inaccessible surfaces; and selection of appropriate instrumentation to meet IE Circular 81-07 criteria. The licensee is not releasing material in aggregate quantities. Consequently, no procedures or methodology for release of aggregate quantities of matter are in place.

l (0 pen) Inspector Follow Item (86-07-06) Licensee to establish methodology and procedures for determining personnel intake of airborne radioactive material. The inspector met with cognizant personnel and reviewed appli-cable procedures and guidance in the area. The following matters needing licensee attention were identified: the equation used for estimating intake of airborne radioactivity based on whole body counting is incorrect; action levels for initiating action to preclude recurrence of j intakes would not ensure compliance with 10 CFR 20; and the basis (10%

MPOB) for performing assessment of intakes of alpha and pure beta emitters was not provided. The licensee's actions on these matters will be reviewed during a future inspectio (Closed) Inspector Follow Item (86-07-08) Licensee to upgrade procedure

guidance for use of additional personnel monitoring devices and placement of personnel dosimetry to address concerns discussed in IE Information j Notice 83-59, " Dose Assignment for Workers in Non-Uniform Radiation Fields." The licensee placed additional guidance in procedure SA-AP.ZZ-024(Q), " Radiological Protection Program," to address IE Infor-mation Notice 83-59.

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(Closed) Inspector Follow Item (86-07-09) Licensee to upgrade

, procedure guidance and controls to ensure airborne radioactivity sampling equipment used in the field was set up and operated

!, consistent with calibration limitations (e.g. use of proper hose i length). The licensee upgraded procedures to address this matte :

l (Closed) Inspector Follow Item (86-07-10) Licensee to complete j preoperational and surveillance testing of safety related ventilation

! systems. The inspector reviewed the outstanding issues discussed in Inspection Report 50-354/86-15 (Section 4.2.b). The licensee ,

adequately addressed the described concerns except the following:

one bolt was found broken on the Technical Support Center Ventilation i System charcoal addition hatch. Although the system met in place i filter testing criteria, the broken bolt should be replaced prior to ,

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the next charcoal change ou ! (Closed) Inspector Follow Item (86-07-13) Licensee to train personnel i in locations of transient and unusually high radiation areas and

access requirements thereto. The licensee contacted other facilities to determine the location of such areas. The licensee established

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procedures to control access to these areas and is training

appropriate radiation protection personnel in the location of such

areas. High radiation area access control requirements are discussed i in general employee trainin t l (Closed) Inspector Follow Item (86-15-01); HPCI/RCIC Steam Isolation i

Timer. The inspector discussed the basis for the deletion of the time requirements for the High Pressure Coolant In,iection (HPCI) and

the Reactor Core Isolation Cooling (RCIC) steam supply isolation

] valves with the NRR staff. The issued Technical Specifications

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indicate that a maximum isolation time is not applicable for the HPCI and RCIC steam supply isolation valves. This item is close i

, 2.2 TMI Action Plan Items

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j (Closed) TAP II.K.3.16 Reduction of Challenges and Failures of Relief

Valves - Feasibility Study and System Modifications. Failure of the 3 power-operated relief valve to reclose during the TMI-2 accident i

resulted in damage to the reactor core. As a consequence, relief valves at Hope Creek were examined with a view toward their possible role in a small break loss-of-coolant accident. This TMI Action Plan item required the licensee to-conduct a feasibility study and make

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j necessary modifications to reduce the number of challenges and

failures of safety relief valves (SRV). The following three features have been incorporated into the Hope Creek desigr and have been I accepted by the NRC staff in the Safety Evaluation Report

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Providing a low-low set (LLS) relief logic system,

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Lowering the reactor pressure vessel water level isolation setpoint for main steam isolation valve (MSIV) closure from j level 2 to level 1, and

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Instituting a preventive maintenance progra I i

The NRC staff evaluation of the BWR Owner Group response to II.K.3.16 l

! concluded that the 3 items listed above are acceptable and effective  ;

in reducing SRV challenges and failures.

l There are 14 Safety Relief Valves installed with the following lift i settings:

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4 SRVs - 1108 psig i

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5 SRVs - 1120 psig

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5 SRVs - 1130 psig The low-low set (LLS) function changes the setpoint of valve F013H from 1108 psig to 1017 psig and valve F013P from 1120 psig to 1047 psig. Since these 2 SRVs now have a lower lift setpoint, their lifting may prevent a challenge to the remaining 12 SRVs. By

lowering the MSIV trip setting from level 2 to level 1, fewer SRV i challenges will occur as a result of MSIV closures. The SRV preventative maintenance program is based on operational feedback

experience derived from NRC, INP0 and vendor programs. This item is j close , 2.3 Construction Deficiency Reports '

! (Closed) Construction Deficiency Report (86-00-02);_Tobar Pressure j Transmitter Terminal Connections. One hundred and eighteen Tobar pressure transmitters were incorrectly installed such that the

transmitters were not environmentally qualified. The correct design

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called for the use of butt splices at the connection of the external

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cable conductors to the pigtails of the transmitters. Contrary to the design, the transmitters were installed utilizing the internal terminal blocks. Use of these terminal-blocks does not provide-adequate moisture sealing of the electrical connectors. The inspector reviewed Discrepancy Report SE-86-002, Startup Deviation

! Reports EA-0740, EA-0756, EG-0776, GJ-0302, GS-0468, GS-0476,

KP-0019, GU-0590, GV-0592 and the corresponding completed work orders which controlled the rework. The rework consisted of removing the j connections at the terminal blocks and making up the required parallel butt splices as per Bechtel drawing No. - E-1408-0. All work

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I was completed prior to commencing fuel load activities. This item is ,

closed.

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(Closed) Construction Deficiency Report (86-00-06); Environmental

! Qualification of Tobar Pressure Transmitter Housings. During environmental qualification testing of Tobar pressure transmitter '

polyester housings and base assemblies, the electrical amplifier assembly of the transmitters failed during irradiation testing. This deficiency was applicable to 53 transmitters included in the harsh Environmental Qualification Program. These transmitters are located outside of the primary containment but inside of the reactor building in following systems:

Residual Heat Removal (RHR)

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Filtration Recirculation and Ventilation System (FRVS)

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Fuel Pool Cooling (FPC)

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Main Steam Isolation Valve Sealing System (MSIVSS)

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Containment Atmosphere Control Rather than make modifications to the Tobar transmitters, the i

licensee decided to replace them with Rosemount 11538 transmitters.

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" The 6 transmitters located in SACS and 1 in RHR were replaced prior to fuel load and the remaining 46 were replaced as of May 15, 198 .

The inspector reviewed design change request (DCR) 4-BC-J-86-294,

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Instrument and Controls procedure IC-DC.ZZ-088 "Rosemount i Differential Pressure Transmitter Calibration" and special i instruction 86-03-31-066-1 "Tobar to Rosemount Replacement" which

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controlled the associated work activities. The inspector also '

witnessed portions of the transmitter replacement throughout the

! inspection period. Calibration data for the new transmitters were

verified to be within tolerance The inspector has no further questions.

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Operational Safety Verification l 3.1 Documents Reviewed l 4 -

Selected Operator's Logs

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Senior Shift Supervisor's Log

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Jumper Log

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Radioactive Waste Release Permits (liquid & gaseous)

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Selected Radiation Work Permits (RWP)

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Selected Chemistry Logs

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Health Physics Watch Log

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3.2 The, inspectors periodical?y: tourea' the piant during regular and

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backshift periods. These tours included '.ne contral room', Reactor, Luxiliary, Turbine and SeEvice J' ater buildings, tar.1 the drywelt (when caccess is possible).) During the inspection a._tivities, ciscussions

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"were held with operators,l technicians (HP & I&C), mechanics, super-

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visors, and plant rinagem9nt. The purpose of the inspection was to

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4 . affirm the licensee's commitments and compliance with 10 CFR,

echnical Specifications, and Station Procedure ,

(1) On a daily basis, pa/ticular attention was directed to the following areas:

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Instrume7tation and recorder traces for abnormalities;

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Adherence to LCO's directly observable from the

/ control rgom;

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Proper control room shift . manning and access control;

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Verificaricq.of the status of control room annunciators that are in alarm;

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hvaperuseofprocedures;

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Review c/' Logs- to obtain plant conditions; an'd,

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Verifica'.ior of surveillance testing for t'imely completir ' f

(2) On c. weekly nasis, the ' inspectors confirmed 'the operability 1 of selscted ESF trains by: '

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Verifying that accessible. valves in the flow path were

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in the correct positions;' ,

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Verifying that poder supplies and breake'rs were in the correc'. positions;

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Visually inspecting major components for leakage, lubrication, vibration, cooling water supply, and genera! operating conditions; and,

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Visually inspecting instrumsntation, where possible, for -proper operabilit '

(3) On a bi~weeklf bails, the inspe'ctors:

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Verified the correct application.of a tagout to a safety-related system;

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Observed a shift turnover; .

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Reviewed the sampling program including the liquid and ,

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Verified that radiation protection and controls were

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Verified that the physical security- plan was being  ;

j implemented;

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Reviewed licer.see-identified problem areas; and,

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Verified selected portions of containment isolation lineu .3 Inspector Comments / Findings:

The unit entered this report period in Mode 5 with the reactor '

shutdown, coolant temperature less than.140 degrees F and

preparations being made for the reactor coolant pressure boundary hydrostatic test.

On May 2, 1986 at 4:00 p.m., an alert was declared in accordance with Emergency Classification Guide section 9 when offsite power -

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was lost to the four IE 4160 V vital buses at 3:35 p.m. and only two of four emergency diesel generators (EDG) were available for loadin :

Although only 2 of 4 EDGs are required to satisfy Technical Specifi- '

cations in Modes 4 and 5, the emergency classification guida does not t

leave any room for interpretation and requires the declaration of an '

alert. The licensee made the required notifications, restored power '

to the vital busas and terminated the alert at 4:05 '

The loss of offsite power occurred when a relay technician i conducting relay maintenance on the 3-4 500 KV breaker failure

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circuitry re energized the wrong set of relays. This personnel error caused section 2 of the 500 KV ring bus to be de-energized and because the 13.8 KV ring bus was in an off-normal alignment the 4 vital buses were also de energized. This was not a total

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loss of offsite power in that portions of the 500 KV and 13.8 KV ring buses were still receiving power via the New Freedom 500 KV ,

transmission lines. Upon receipt of the loss of power signal '

the "A" and "C" diesels started and loaded as expected. The "B"

, and "D" diesels were inoperable at the time due to maintenance '

] and preoperational testing respectively. After restoration of i power, all systems were returned to norma The inspector i J reviewed the licensee's fact finding report and the licensee 1 i

event report and has no further questions at this time.

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i l On May 5, 1986, reactor vessel pressure was being increased to j hydrostatic pressure when the licensee observed that one of the  !

reactor vessel metal temperature indications was not being  ;

maintained in accordance with the requirements of Technical

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Specification figure 3.4.6.1- Pressure was'immediately reduced to approximately 600 psig so that the proper pressure-temperature relationship was established. The inspector reviewed safety evaluation number PSE-SE-M-039 " Evaluation of Reactor Coolant System Out-of-Limit Temperature / Pressure Condition" dated May 6, 198 The licensee concluded, with the concurrence of General Electric, that the structural integrity of the reactor coolant system was unaffected by the out of limit temperature / pressure condition Because of the licensee's timely identification of the problem and the prompt cor-

{ rective action taken, the inspector determined that no further NRC action was required.

i On May 6, 1986, an operator detected unusual Reactor Coolant System (RCS) pressure indications during a RCS vessel hydrostatic

pressure tes During an investigation to determine the cause of the unusual indications, an inadvertent ESF actuation occurred due to a Channel "A" LOCA initiatio This occurred when a technician cracked

. open an instrument rack isolation valv It was determined that a number of these valves were not in the required open position which

, led to the observed erroneous RCS pressure indications. The licensee

subsequently determined that these (and other similar) valves had not been included in the valve lineups for the I&C department. The-licensee commenced walkdowns of the instruments to identify any other such valving errors. None of the adversely affected instru-ments that were found during the licensee's walkdowns were rcquired for operating conditions 4 or 5 and therefore no Technical Specifi-cation required equipment was found inoperabl The inspector will review the licensee's corrective actions to prevent future occur-rences during a future inspection after the licensee submits a report of the even ,

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On six separate occasions during this report period, a single

, channel Loss of Coolant Accident (LOCA) signal was generated l from a not always apparent cause. It appears likely that at least some of the LOCA signals resulted from valve operations on or around the reactor pressure and level instrument racks which feed the Reactor Protection and Emergency Core Cooling System logic schemes. However, because the exact cause for all of these LOCA initiations has not been positively determined, the licensee formed a task force to identify the root cause of these ESF actuations. The investigation included a review of all available data, including: process computer sequence of events printouts, control room and remote parameter charts, special

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.- 10 monitoring equipment outputs and interviews of operators and

, technicians. Special Test Procedure 86-05-29-112-5-1 was performed on May 31, 1986, in an attempt to examine the possible causes due to hydraulic, electrical or any RFI/EMI interferenc The results of this test procedure were still under evaluation at-

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the conclusion of this report period and will be reviewed during the next inspectio (86-26-01)

On May 8, 1986, OP-IS.ZZ-001(Q) " Inservice System Leakage Test of the Reactor Coolant. Pressure Boundary" was conducted. As a

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-result of this test, 12 control rod drive mechanisms were observed to be leaking and the licensee. decided to repair all i leaks no matter how mino The inspector performed an independent verification of selected portions of the valve lineups and witnessed -

portions of the reactor vessel heatup and pressurization. The 0-ring

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replacement on the control rod drive mechanisms was witnessed by the j inspector and is discussed in paragraph 5 of this repor t On May 12, 1986, the licensee made an ENS notification that both loops of Service Water (SW) were declared inoperable. The A and C pumps (A loop) were inoperable because of problems with the pump strainers, and the B loop was inoperable because the B pump had failed a surveillance test and the "D" Diesel Generator was

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! inoperable for preoperational testing. Technical. Specifications require a SW loop to have 2 operable pumps. Although, by Technical

Specifications, the "D" SW pump was inoperable it remained in service i and supplied plant cooling loads. Service water being inoperable requires the ECCS systems, Safety Auxiliary Cooling System (SACS) and the Emergency Diesel Generators to also be declared inoperable. .The plant was already in Mode 4 (shutdown, temperature less than 200

, degrees F) and secondary containment integrity was establishe Repairs were completed on the SW system and the system declared

operable prior to the end of the report perio .

At 5:55 a.m. on May 19, 1986, a scram signal was initiated when j an I&C technician valved in a reactor pressure transmitter which

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caused a "B" RPS low reactor water level trip. In an unrelated

activity, the "A" RPS was already in a half scram condition I

because of surveillance testing of the "E" APRM. The plant was in Mode 4 (shutdown, RCS less than 140 degrees F) and no control rods were withdrawn. All systems responded as expected for the i plant condition On June 7,1986, the "A" Standby Liquid Control (SBLC) pump

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started, the squib valve fired and the reactor water cleanup system isolated coincident with the manual start ~of the'"A" Emergency Diesel Generator. The SBLC system ran for approximately 15 seconds before being tripped by an operator allowing the sodium -

pentaborate solution to flow to the reactor vessel injection line m o

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It appears that a very small amount of sodium pentaborate entered the  ;

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vessel since chemistry samples indicated 40 PPB compared to a minimum

detectable concentration of 20 PPB. ~By June 9, reactor water samples indicated less than minimum detectable concentration. The cause of the SBLC system initiation is unknown as of the end of this report period and efforts to duplicate the event have not been successful i to date'. The inspector is continuing to follow up on this event

during the next inspection perio No violations were identifie . Surveillance Testing I

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During this inspection period the inspector performed detailed technical-procedure reviews, reviewed in progress surveillance testing as well as completed surveillance packages. The inspector also verified that the surveillances were performed in accordance with licensee approved proce-dures and NRC regulations. The inspector also verified that the instru-j ments used were within calibration tolerances and that qualified techni-cians performed the surveillance ] The following surveillances were reviewed, with portions witnessed by

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OP-IS.ZZ-001(Q), Inservice System Leakage Test of the Reactor Coolant Pressure Boundary

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OP-ST.SF-001(Q), Reactor Mode Switch Functional Test - 18 Month

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No violations were identifie , Maintenance Activities

, During this inspection period the inspector observed selected maintenance activities on safety related equipment to ascertain that these activities-

' were conducted in accordance with approved procedures, Technical Speci- ,

fications, and appropriate industrial codes and standard Portions of the following activities were observed by the inspector:

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! Work Order Procedure Description i'

86-05-12-173-4 MD-CM.EA-003(Q) Service Water. Strainer 3 overhaul and repair

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Various Work Orders MD-CM.BF-008(Q) Control Rod Drive Shoot (see narrative below) out Steel Removal, Replacement and i

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i - 12 Work Order Procedure Description Various Work Orders MD-CM.BF-001(Q) Control Rod Drive (see narrative below) Removal and Replacement As a result of the " Inservice System Leakage Test of the Reactor Coolant Pressure Boundary" (OP-IS.ZZ-001(Q)) conducted on May 8,1986, a i

total of 12 control rod drive (CRD) mechanisms were identified to have leaks that the licensee determined needed'to be repaired. These repairs

. consisted of dropping the CRD from the bottom of the vessel, replacing the

0-ring seals and reinstalling the CRD. The leaking CRDs and the

corresponding work orders are listed below

CRD Work Order i

26-55

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86-05-06-010-7 42-31 86-05-06-012-3

10-43 86-05-12-021-5

? 30-35 86-05-06-011-5

42-07 86-05-06-013-1 02-27 86-05-15-178-1 38-43 86-05-15-179-0 26-59 86-05-15-177-3 50-27 86-05-06-015-8 22-51 86-05-15-176-5 54-27 86-05-15-174-9 30-11 86-05-15-175-1

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On May 22, while observing the work activities associated with the repairs to CRD 42-07 the inspector noticed that the work package and maintenance procedures were not being referred to by the maintenance

personnel. Although the mechanics were well into the CRD removal process, the procedure prerequisites had not all been satisfied nor had the supervisor made the required procedure signoff. Further i

review found that the work package contained a torque specification sheet that was not adhered to during disassembly and that the mechanics did not know was in the work package. The inspector discussed these concerns with a quality control (QC) inspector who had arrived at the work site. The QC inspector stopped the work activity so that the job supervisor could be located to make the prerequisite signoff. When the supervisor arrived at the job site he made the signoff, indicating that all of the prerequisites had been satisfied even though this was not true. Because there are no contamination / radiation problems associated with a non-irradiated'

reactor vessel and fuel, many of the procedure prerequisites were not-technically necessary, but because of the apparent disregard for l procedural compliance, the inspector immediately brought his concerns

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.- 13 to the attention of plant management. All CRD repair work was immediately stopped until the necessary procedure changes were made and the need for strict procedural adherence reinforced. Work recommenced approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later. At a later date, the inspector reviewed QA surveillance report HQC-85-628 which documented the QC inspector's findings of the same work activity. The closecut of this QC report will be reviewed in a future inspectio (86-26-02) .

No violations were cite . Engineered Safety Feature (ESF) System Walkdown The inspectors verified the operability of the selected ESF systems by performing a walkdown of accessible portions of the system to confirm that system lineup procedures match plant drawings and the as-built configuratio This ESF system walkdown was also conducted to identify equipment conditions that: might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriat The Reactor Core Isolation Cooling and the "A" Loop of the Low Pressure Coolant Injection (Residual Heat Removal) systems were inspected and the Standby Liquid Control (SBLC) system walkdown was completed from the previous inspectio No violations were identifie . Preoperational/Startup Testing 7.1 Area Radiation Monitoring System / Process Radiation Monitor The inspector reviewed the preoperational testing of the Area Radiation Monitoring (ARM) Syste The following matters were reviewed:

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adequacy of calibration

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alarm point settings and verification

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local and remote alarm actuation

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review and approval of test results

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conformance with applicable Technical Specification Documents Reviewed

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Procedure PTP-SD-1, " Area Radiation Monitor to RM-11 Computer Preoperational Test,"

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Procedure DTP-SD-001, "GA Technologies Area Radiation ,

Monitors,"

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Procedure PTP-SP-1, " Peripheral Device Communication and Operations Test" .

Findings e

Within the scope of the review, no deficiencies or unacceptable conditions were identified. The licensee met commitments -

provided to the NRC relative to completion of selected ARM testing prior to fuel loa ;

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The following matter remains open:

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interface of ARMS with computer system (RM-11). This will !

be verified through performance of PTP-SD- '

Other matters brought to licensee attention:

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High range containment monitor found off-scale at RM-80

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Procedure cover sheets for refuel floor exhaust monitor calibrations incorrec The licensee initiated a eview of these matter .2 Radiation Surveys (Startup)

The inspector reviewed the status of the Startup Radiation Survey Program. The review was conducted per the following criteria:

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Final Safety Analysis Report (FSAR) Chapter 14, " Test Program"

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Procedure TE-SU.ZZ-021(Q), " Radiation Meas'rements" u

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General Electric Radiation Measurements Test . Number 2

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General Electric Field Deviation Disposition Request KT1-1743, Startup Test Specification Revision Findings The licensee is currently reviewing and revising the startup-tes The licensee has elected to use local site environmental survey data as background radiation survey data. No deficiencies or unacceptable conditions were identifie .

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7.3 Non-Safety Related Ventilation System Testing The inspector met with cognizant licensee personnel to review and discuss the status of testing of non-safety related ventilation system The review was conducted per the following criteria:

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Regulatory Guide 1.140, " Design, Testing, and Maintenance Criteria for Normal Ventilatien Exhaust System Air Filtration and Absorption Units of Light Water Cooled Nuclear Power Plants," Revision 1

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Final Safety Analysis Report Chapter 14, " Initial Test Program"

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IE Information Notice 82-43, Deficiency in LWR Air Filtration / Ventilation Systems, November 1982 Findings Within the scope of this review no violations or unacceptable conditions were identifie The following matter remains open (86-26-03)

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Place non-safety related ventilation systems on a schedule for periodic testin . Licensee Event Report Followup The licensee submitted the following event reports during the inspection period. All of the reports were reviewed for accuracy and timely submissio Certain designated reports were followed up by the insp,; tor for corrective action implementatio Inadvertent "B" Channel LOCA Signal During Surveillance Test Performance 86-003 Inadvertent RPS Initiation During Performance of NMS Component Troubleshooting Activities

86-004 Non-coincident Scram Signal Resulting From Neutron Monitoring System Component Failure

86-005 FRVS Inoperability During Core Alterations86-006 Primary Containment Isolation Resulting From A Procedural Inadequacy s_

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86-007 "B" Channel ESF Actuation Due To Spurious Signal

86-008 Missed Surveillance During Initial Core Loading Due To Personnel Error 86-009 Inadvertent RPS Initiation During Surveillance Testin *

86-010 "A" Channel-ESF Actuation Due To Spurious Signal

86-011 Inadvertent Loss of Offsite Power

See narrative below for detail LER 86-004 At 2:49 a.m. on April 16, 1986, a non-coincident scram signal was initiated from the "B" Reactor Protection System when Local Power Range Monitor (LPRM) 4C-32-25 failed high resulting in a channel F APRM upscale trip signal. The plant was in operational Mode 5 with initial fuel loading activities in progress and the RPS shorting links removed as required by Technical Specifications. All rods were already fully inserted and therefore no rod movement occurred as a result of the scram. The failed LPRM was bypassed, the scram reset and fuel load activities recommenced. The licensee's investigation later revealed that a gain switch on an LPRM auxiliary card (GE part no. 136B2503AAG1) had failed and caused the LPRM to fail upscale. The inspector reviewed work order 86-04-18-030-3 which replaced and calibrated the LPPM auxiliary card. The inspector has no further cuestions LER 86-005 On April 16, the inspector identified a violation of Technical Specifications when the Filtration, Recirculation, Ventilation System (FRVS) fans were found to be in lockout and therefore inoperable during core alteration This violation is discussed in detail in Inspection Report 50-354/86-20 and the licensee's response to this violation will be reviewed in a future inspectio LER 86-008 On April 24, 1986, the inspector identified a violation of Technical Specifications when the Standby Liquid Control, System surveillances were not current during fuel handling and contFol rod testing. This violation is discussed in detail in Inspection Report 50-354/86-20 and the licensee's response to this violation will be rev'ewed in a future inspectio !

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These reports detail ESF actuations that occurred as a result of spurious reactor vessel low level signals due to an unknown cause(s).

At this time the licensee has not yet identified a definitive cause; however, they are still monitoring' the ESF instrumentation in order to do so. This effort is further detailed in paragraph 3.3 of this repor LER 86-011 This report detailed the inadvertent loss of offsite power while  !

conducting relay testing. The licensee determined that the root cause of this event was due to personnel error in failure to follow procedur This event resulted in the licensee declaring an alert in accordance with the Emergency Classification Guideline (ECG). The licensee's corrective actions include using this event in future technician training to prevent recurrence and also resulted in changing the ECG to recognize plant electrical systems requirements authorized by Technical Specifications for the operating condition that exists at the time of the even This would prevent declaring an alert for this type of event in the futur . Biscoseal Deficiency Report (Part 21)

In a letter dated April 28, 1986,- Bechtel Construction, Inc. notified the NRC of a defect found in a number of Biscoseal floor penetrations at Hope Creek. Certain of the Biscoseal penetrations had developed a Biscoseal material to sleove separation of approximately 1/32" to 1/16" around at least a portion of the sleeve perimeter. In some cases these defective seals are located directly over Class 1E equipment creating the potential for electrical damage due to a water leak from the above room. The observed condition does not maintain the required boundary for hydro and air pressure differentials, and therefore the design intent of the penetration seals is not me Corrective actions were completed as of June 6, in order to suppor the reactor's initial criticality testing. Throughout this report-period the inspector periodically witnessed the in process repair activitie The corrective action plan consisted of the following:

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.walkdown inspections to identify and document all deficiencies

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development and testing of a repair procedure

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application of a flexible caulking to correct the separation condition

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installation of vertical structural support where required The inspector will review the completed work packages in a future- -

inspection. (86-26-04)

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. 18 10. Allegation Followup 10.1 Hiring Impropriety On March 24, 1986, the NRC resident inspector received a copy of an anonymous letter which alleged that a PSE&G employee had accepted a payment of money for getting a contractor a job position at Hono Crook, The original letter had been sent to the PSE&G Vice President - Nuclea The inspector reviewed the licensee's investigation into this allegation and independently reviewed the qualifications of the contractor involve Based upon the information available to the inspector this allegation is unsubstantiate .2 Tagging Procedures The NRC received an anonymous allegation dated October 14, 1985, regarding the firing of contractor personnel over a valve tagging issue. Region I requested information from the licensee and received their response dated December 16, 198 The inspector reviewed the licensee's response and. during a site inspection, confirmed its accuracy. This inspection verified that those individuals fired had been trained tc use the tagging procedure and had, in fact, violated the tagging procedure. This created a potential personnel risk that was unacceptable to plant management. It was also confirmed that other employees had been previously fired for the same type of procedure violation Therefore, the subject allegation is unsubstantiated and is considered close No violations were identifie . Safety Issue Survey The inspector surveyed the licensee's programs for a selected sample of safety issues in accordance with temporary instruction 2515/7 The inspector reviewed P& ids, operating procedures, maintenance procedures, conducted in plant inspections, and discussed issues with the system engineers to determine the system's status, and information on the reliability of the High Pressure Coolant Injection / Reactor Core Isolation Cooling systems. The licensee's progran for monitoring biofouling of cooling water heat exchangers was also reviewe No violations were identifie .

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12. Management Meeting On June 5, 1986, a meeting was held between Public Service Electric

and Gas Company and the NRC Region I staff in King of Prussia, Pennsylvania. The purpose of the meeting was to discuss operating

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experience and lessons learned since receipt of the low power

license. The licensee was requested to address each of the following

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Discuss extent to which operational problems at sister plants were reviewed, and the benefits, if any, of such reviews ( design modifications, procedure improvements, etc.).

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Discuss measures used to preclude unacceptable work practices in the balance of plant which could affect safe operation of the facilit Include any lessons learned from Salem operational-history (e.g. recent event involving use of nylon line to secure limit switch in SGFP valve).

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Discuss results of any analysis performed to assess reasons for operational events (related causes, etc.). Of particular concern are the events attributed to design feature Discuss the status of determining the cause(s) for spurious ECCS actuations, and corrective actions identifie Discuss management personnel changes in the station organization (rationale, qualifications,etc.).

The list of attendees and a copy of a handout provided by PSE&G during the meeting is provided as enclosure (1) to this repor . Exit Interview The inspectors met with licensee and contractor personnel periodically and at the end of the inspecticn report to summarize the scope and findings of their inspection activities. Written material was not provided to the licensee during the exi ,

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During the course of this inspection, the licensee was provided a written listing of NRC open items from previously issued inspection reports. All of the information provided in the open item list was obtained from publicly available issued inspection reports and was i provided to the licensee in order to more effectively address l outstanding NRC concern '

Based on Regi_on I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restriction l

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Enclosure 1 June 5, 1986 Meeting Between PSE&G and NRC Region I List of Attendees Name Title Organization W. Kane Dep. Dir. Div. Reactor Projects NRC Regior. I T. Murley Regional Administrator NRC Region I R. Starostecki Director DRP '

NRC Region I S. Collins Chief, Reactor Projects Branch 2 NRC Region I W. Bauer Principal Engineer PSE&G/Offsite

Safety G. Peet Lead I&C System Engineer PSE&G/ System Engineer B. A. Preston Manager - Licensing & Regulation PSE&G C. A. McNeill Vice President - Nuclear PSE&G R. S. Salvesen General Manager - Hope Creek PSE&G Operations S. LaBruna Assistant General Manager - PSE&G Hope Creek J. Nichols Technical Manager - Hope Creek PSE&G W. Merritt Senior Technical Supervisor - PSE&G Hope Creek L. Bettenhausen Chief, Operations Branch, DRS NRC Region I R. Borchardt SRI, Hope Creek NRC Region I

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REVIEW OF OPERATIONAL EXPERIENCE AT SISTER PLANTS

l 0 AGGRESSIVE APPROACH o Review began during Hope Creek Construction o Over 3000 items reviewed to date

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NRC

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INPO

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GE o Review continuing o REVIEW HAS RESULTED IN PROCEDURE AND DESIGN CHANGES o TIL 965 - Copper Dusting operating procedure changes to minimize time on turning gear. Design change to remove high speed operation o SIL 250 - SRV Torus Loading Operating procedure changes to assure even distribution of SRV use o SOER 82-2 - Inadvertent Reactor Pressure Vessel Pressurization Operating procedure changes to specify minimum natural recirculation reactor water level .

o SOER 8 2-04 - RHR System Leakage Paths Design change requested to modify RHR valve interlocks to prevent the inadvertent draining of the RPV o OE 1556 - Dropped Fuel Assembly Hope Creek reviewed the OE and determined the rigging components at the station were properly matched. Th is review was complete prior to the issuance of the INPO SER and GE SI o OE 1706 - Limerick Unit 1 SJAE Problems Similar drainage problem noted in the OE was discovered by Hope Creek Operators before the OE was received. Design change has been implemente o REVIEW CONTINUES AFTER CLOSEOUT o Incident Report "Look Rack" (i.e. is this a problem we should have known about)

- Majority of incidents are plant specific

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. o DISSEMINATION OF KEY ITEMS TO MANAGEMENT FOR INFORMATION o Increases visibility o Maintains awareness of industry trends o Examples:

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Nuclear Network

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Monthly Operating Experience Summary

o USE OF OPERATIONAL EXPERIENCE IN HOUSE o Over 300 Recommendations from Scram Reduction Walkdowns

Examples

- Relocation of 1" sensing line for high moisture separator level turbine trip

- Replacement of solid doorc with louvered doors on control rod drive instrumentation cabinet to reduce the potential of electronic equipment within f ailing due to overheatin Flow element in mechanical vacuum pump discharge piping o Operating events at Hope Creek are screened for generic applicability and distribution to INPO I

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o Component f ailures into Nuclear Plant Reliability. Data System (NPRDS)

o Lessons learned included for training review i

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BOP 1:ORK PRACTICES AND LESSONS LEARNED PROM SALEM o ADMINISTRATIVE PROCEDURES TREAT BOP SYSTEMS THE SAME AS NSS SYSTEMS o Require pro'cedures to perform work o Incident reporting similar o Industry experience feedback includes BOP systems o SALEM OPERATIONAL EXPERIENCE WAS INITIALLY FACTORED INTO HOPE CREEK PROGRAMS o Initial department manning An 1980 to develop department procedures and practices were Salem. experienced supervisors o Station AP's patterned af ter Salem's o HOPE CREEK PROGRAMS ARE CONTINUALLY MODIFIED TO REFLECT SALEM EXPERIENCE o VPN Procedures Manual regn. ras consistency between plants o The Vice President - Nuclear personally directs appropriate Salem LER's to Hope Creek for consideration o The Nuclear Training Department incorporates both Salem and Hope Creek operating experience into training programs o TO ENSURE APPROPRIATE S ALEM EXPERIENCE I'S REING INCORPORATED AT HOPE CREEK, THE FOLLOWING IS BEING DONE o LER's and NRC insnection reports for 1984 to present are being screened f or applicability. Appropriate " lessons" will be reviewed with the Hope Creek work force in

" tailgate" sessions. This ef fort will be complete by

.Tu ly o The Nuclear Department RCT will routinely review both Salem and Hope Creek incident reports and f eedback information as appropriate o Hope Creek will institute INPO observer training and establish an observation program similar to Salem's before the end of 1986

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ANALYSIS OF OPERATIONAL EVENTS o APPROXIMATELY 4 8 OPERATIONAL EVENTS (INCIDENT REPORTS)

SINCE THE RECEIPT OF A LOW POWER LICENSE o 24 Require Licensee Event Reports o ROOT CAUSE CATEGORIZATION o Personnel Frror o Procedural Inadequacy o Equipment Failure o Design Inadequacy o Other o CATEGORIZATION OF PERSONNEL ERROR o Considered a plant specific issue, not inadequate training o Predominant indication is learning curve on procedures o DESIGN CRANGES

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o Few required to correct root cause o Design changes made even if design was not the root cause if deened necessary for defense in depth

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' ROOT CAUSE SUMMARY

EQUIPMENT OTHER P ROCEDURE DESIGN PERSONNEL INADEQUACY FAILURE d ERROR INADEQUACY 9 5 4 2 IR 23 l

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4 5 3 0 LER 11

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DESIGN CHANGE SUMMARY o LER 86-00 3,009 INADVERTENT RPS TRIP DURING TROUBLESHOOTING OF NUCLEAR MONITORING SYSTEM Root Cause: Personnel error Corrective Action: Trainin An additional action will be to cushion the IRM cabl *

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o LER 86-006 ESP ACTUATION, B CHANNEL Root Cause: Procedural error Corrective Action: Procedures will be revised to require resetting of the trip logic memor Also, the Incident Report which generated the LER requires a design change to provide memory status in the control roo o LER 86-007,010,016 ESF ACTUATIONS - LOCA SIGNALS Varies between personnel error and unknown i

Root Cause:

Corrective Action: Training and procedure changes. In aadition, a number of design changes are being considered or implemented to preclude testing induced transient Additional Action: As a result of the review of the R Channel LOCA, RRCS instrumentation is being relocated and the EPAs for RPS alternate power will be set with a time dela _--

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e EVALUATION OF ESP ACTUATIONS o ESP ACTUATIONS WITH UNKNOWN ROOT CAUSES DATE DESCRIPTION EVENT INCIDENT REPORT N /20/86 ESF ACTUATION ON R CHANNEL

86-048 4/26/86 ESF ACTUATION ON A CHANNEL

- 86-057 5/6/86 ESF ACTUATION ON D CHANNEL

86-064 5/13/86 ESF ACTUATION ON D CHANNEL

86-065 5/13/86 ESF ACTUATION ON D CHANNEL

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o MOST PROBABLE ROOT CAUSES FOR ESF ACTUATIONS A PERTURBATION IN COMMON SENSOR LINES OF LOCA INSTRUME i

' A PERTURBATION IN THE COMMON POWER SUPPLIES FOR LOCA INSTRUMENTATIO . SOME TYPE OF REACTOR VESSEL TRANSIEN l

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EVALUATION OF ESF ACTUATION o CONCLUSION / RECOMMENDATIONS o Prior to initial criticality

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Blow back instrument lines

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Install identification tags on LOCA/ECCS instrument sensor lines

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Install quick disconnects on LOCA/ECCS instruments

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Review incidents with all I&C Technicians o By July 15, 1986

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Complete installation of identification tags on RPS/NSSS instrument sensor lines

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Review LOCA/ECCS surveillance procedures

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Include instrument sensor valves in status log

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o By September 1, 1986

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Complete review of RPS/NSSS surveillance procedures

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Complete installation of cages around LOCA instrument racks

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Q MANAGEMENT ORGANIZATIONAL CHANGES o Management responsiveness can be enhanced by minimizing the

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layers of management between the plant manager and the worke i o We propose to reduce the levels of management in the near term by taking the Department Engineers out of the line position and utilizing them as staf f support. In the longer term, we will also eliminate the Assistant General Manager positio t o To assure this move is effective, we are considering rotating personnel to assure the most qualified individuals are in the line position .

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