IR 05000354/1987024

Insp Rept 50-354/87-24 on 870929-1026.No Violations Noted. Major Areas Inspected:Followup on Outstanding Insp Items, Operational Safety Verification,Surveillance Testing,Maint Activities,Esf Sys Walkdown,Ler Followup & Monitors

ML20236T318
Person / Time
Site:	Hope Creek
Issue date:	11/20/1987
From:	Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20236T308	List: IR 05000354/1987024 ML20236T306
References
50-354-87-24, IEB-74-14, NUDOCS 8712010136
Download: ML20236T318 (15)

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L U. S. NUCLEAR REGULATORY COMMISSION

REGION I

050354-870213 050354-870829 Report ~N '50-354/87-24 Docket 5p354

. License NPF-57 Licensee: Public Service Electric and Gas Company Facility: Hope Creek Generating Station Conducted: September 29, 1987 - October 26, 1987 Inspectors: .R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector R. J. Summers, Project Engineer T. J. Kenny, Senior Resident Inspector, Salem Approved:

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[n P. Swetland, Chief, Projects Section 2B Date Inspection Summary:

Inspection on September 29, 1987 - October 26, 1987 (Inspection Report Number 50-354/87-24)

Routine onsite resident inspection of the foll.owing areas:

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Areas Inspected:

followup on outstanding inspection items, operational safety verification, surveillance testing, maintenance activities, engineered safety feature system walkdown, safety relief valve acoustic monitors, enforcement conference, safety review committees, and licensee event report followup. This inspection involved 166 hours0.00192 days <br />0.0461 hours <br />2.744709e-4 weeks <br />6.3163e-5 months <br /> by the inspectors'.

Results: During this inspection, three separate and unrelated instances relating to the lack of strict compliance with written procedures were identi-fied (paragraphs 3, 4 and 5). These examples indicate that a reduced level of I

, attention may exist among some technicians and operators in the area of strict procedural compliance and in the case of one example, the need to fully docu-t ment. activities. This perception is of concern to the NRC in that it may be-indicative of a developing relaxation in the level of attention to detail exhibited at the working leve It was also identified (paragraph 4) that the ECCS logic tester doesn't consistently give' an accurate indication of contact status. The licensee agreed to evaluate this situation and discuss any corrective actions with the inspecto PDR ADOCK 05000354:

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Within this report period, interviews and discussions were. conducted with 1 Nr. S. LaBruna and members of the. licensee management and staff and vari- i

.ous contractor personnel as necessary to support inspection activit ; Followup on Outstanding Inspection' Items' i

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- (Closed) Inspector Follow Item (86-36-01); Primary Containment IsolationLSystem (PCIS) automatic actuation. On August 5, 1986, -i an automatic PCIS isolation occurred during surveillance testing on the "A" channel. PCIS actuated when the "A" test signal was-combined with a "B" channel signal which was " sealed in" from a previous channel acuation. The "B" channel-signal had been gen-erated earlier in the day during backfilling of the "B" instrument rack. Contributing to this event was the fact that unlike the reactor protection system (RPS) there is no control room indication of " sealed in" PCIS actuation signals. The licensee has addressed this issue through operator and technician training, and the PCIS logic is verified reset following surveillance which actuate.the

" seal in" circuit. No further instances of spurious actuations due to unannunciated half-trip signals have occurre The licensee has initiated a design change to install control room annunciation for all initiation signal The modification will be reviewed under the routine NRC inspection program. This item is close . Operational Safety Verification 3.1 Inspection Activities On a daily basis throughout the report period, inspections were conducted to verify that the facility was operated safely and in conformance with regulatory requirements. The licensee's manage-ment control system was evaluated by direct observation of -

activities, totrs of the facility, interviews and discussions with i licensee personnel, independent verification of safety system status

.and limiting conditions for operation, and review of facility record The licensee's adherence to the radiological protection and security programs was also verified on a periodic basis. These inspection activities were conducted in accordance with NRC inspection proce-dures 71707, 71709 and 71881 and included weekend and backshift inspections conducted on October 1 (12:30 a.m. - 6:00 a.m.),

October 9 (12:00 a.m. - 6:00 a.m.) and October 12 (1:15 p.m. -

9:45 p.m.).

3.2 Inspection Findings and Significant Plant Events The unit entered this report period in cold shutdown continuing a scheduled surveillance test outage that began on September 18, 198 i

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Althougn the major focus of the outage was intended to be the com-pletion of surveillance testing, inservice tests, and local leak rate tests, the discovery of 4 small diameter recirculation system instrument lines with fatigue cracks required a major repair effor Details of these line failures can be found in inspection report 354/87-22. Other unplanned work completed during the outage included replacement of the safety relief valve acoustic monitor accelerometers. Additional details on this work activity can be found in paragraph 7 of this report. Throughout the outage period the inspector conducted inspections of maintenance activities, paying particular attention to normally. inaccessible areas of the plant, such as the drywel Control of outage activities, radia-tion protection controls and plant housekeeping were found to be l generally effective. The licensee self-icSntified a number of areas in need of improvement, however neither the licensee nor the inspector identified any concerns of a nuclear safety or regulatory natur The assignment of shift outage manager responsibilities to the senior shift supervisor is a noteworthy initiative. His active involvement with all outage activities ensured a constant emphasis on plant safety, operational awareness, and compliance with technical specification At 8:44 a.m. on October 7, the licensee received a primary contain-ment isolation system (PCIS) channel "C" initiation while energizing panel 10C 652 after completion of a design change package (DCP).

This DCP de-energized panel 10C 652 to relocate electrical leads for installation of a new control room alarm. The PCIS actuation started the "C" service water pump, isolated the reactor building ventilation system, and shut several isolation valves. The system functioned as designed and was immediately reset and returned to normal lineu At 9:36 a.m. on October 10, the reactor was taken critical marking the end of the surveillance testing outage. The retest for replace- l ment of the acoustic monitors included performing safety ralief valve (SRV) lift tests at approximately 750 PSIG. At 7:30 p.m., '

with the reactor at 10*4 power, the retest procedure was commence ;

Eight SRVs were cycled without incident, however at 7:49 p.m. the l

"J" SRV stuck open. After it was determined that the valve could i not be shut, the reactor was manually scrammed and an unusual event {

declared at 7:50 p.m. During the resulting transient, reactor vessel water temperature decreased from 507 degrees to 427 degrees in 10 j minutes and to 324 degrees over an hour's time. Although this cool-down rate was in excess of technical specification limits (100 t degrees / hour) it is well within the nuclear steam system supplier's i design basis for a single relief valve blowdown. The vessel stress report indicates that the maximum stresses due to a cooldown of 151 degrees in 10 minutes, or a rate of 906 degrees / hour are acceptabl In relation to brittle fracture concerns, the vessel temperature throughout the transient was significantly higher than the vessel metal low temperature limits. All other plant systems operated as expected during and after the transient and the unit entered cold

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shutdown at 1:20 a.m. on October 11. The unusual event was then terminated. During the post event review the inspector questioned whether an internal and external inspection of the torus had been performed as required by operations department abnormal procedure OP-AB-ZZ-121, " Failed Open Safety Relief Valve." The inspector was informed that this requirement had been overlooked during restoration

of the stuck open SRV event. Further review by the licensee deter-mined that this inspection step was placed in the abnormal procedure in response to NRC Bulletin 74-14, "BWR Relieve Valve Discharge to Suppression Pool" dated November 14, 1974. It should be noted that Hope Creek was in the early phase of construction at that time and the bulletin was sent to the licensee for information onl In addition, technical specification 4.6.2.1 only requires an external visual examination of the suppression chamber after an SRV operation

with the torus average water temperature greater than or equal to 177 degrees and reactor coolant system pressure greater than 100 psig. Based upon the above and the fact that torus water tempera-ture did not exceed 100 degrees, the inspector agreed that the internal and external torus inspections were not necessary. The licensee has initiated a procedural change to delete this inspec-

. tion step from the abnormal procedure. This fact, however does n(t negate the concern over the failure to adequately follow the abnormal procedure to completion. (This is the first of three procedure com-pliance deficiencies noted during this inspection period). The SRV was replaced and the defective valve sent to Wylie for inspection and testing. The results of this inspection will be reviewed by the inspector when they become availabl The reactor was again taken critical at 7:57 a.m. on October 12. At 9:00 p.ni, the remaining SRVs, including the new "J" SRV were successfully teste At approximately 9:45 a.m. on October 13, an explosion and fire occurred in the phase "A" main transformer which resulted in a main turbine generator trip. Because the reactor was operating at 20% power, no reactor scram occurred nor were any ESF systems actuated. The licensee's fire department responded and the auto-matic water deluge system initiated. No personnel injuries occurred as a result of the incident. The transformer was destroyed by the fire and was replaced with an onsite spare. The licensee shut down the reactor during the main transformer replacement and entered cold shutdown at 10:35 Following the 13-day transformer replacement outage, the reactor was taken critical at 9:16 a.m. on October 26, and remained at less than 10% power while waiting for final transformer checks to be complete The unit remained in this condition until the conclusion of the report perio No violations were identifie ._

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, , 5 4. Surveillance Testing 4.1 Inspection Activity

During this. inspection period the inspectorfperformed-detailed tech-l ~n ical procedure reviews, witnessed in progress surveillance testing,-

and reviewed completed surveillance-packages. The inspector veri-fied~.that the surveillance tests were performed in accordance with-Technical Specifications, licensee approved-procedures, and NRC'

regulations. These inspection activities were conducted _in. accord-ance with NRC inspection procedure 61726.

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OP-IS.BC-005 Residual Heat Removal Inservice Test IC-FT.SE-012 Functional Test of "H" Intermediate Range Monitor

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IC-FT.SP-052 Functional Test of "B" Main Steam Line Radiation Monitor

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OP-ST.BH-002 . Standby Liquid Control System Flow Test

- IC-FT.BB-020 Drywell High Pressure B1 Functional Test

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IC-FT.BB-006 Reactor Vessel Water Level Instrument B21-N691G 4. During observation of the functional test on. reactor' vessel

water . level instrument B21-N691G, the inspector witnessed that step 5.2.1.9.e. could not be performed as writte After establishing certain test conditions,'this step instructs _the technician to verify that the B-C liquid crystal display (LCD) on the emergency core cooling system (ECCS). logic tester indicates an open contact. When.this s_tep was actually performed the B-C LCD on the ECCS logi tester indicated that the contact was closed. The tech-nicians were not surprisec by this indication and state'd i that there were' situations when the ECCS logic tester did .

-not give an accurate indication of contact status. The 'I technicians then proceeded to check the condition of the :J contact using a volt-ohm' meter and verified that the con- _

tact was indeed open as required. Although the actions by )

the technicians were technically correct, the inspector 1 expressed concern over two issues. The first concern i related to the potential design flaw of the ECCS logic tester in that on certain occasions it provides erroneous

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indication of contact status. Of even more concern than the easily recognizable situation where an unexpected status indication is shown is the situation where an erroneous indication provides the expected result In the latter situation a malfunctioning contact status would be masked by the faulty teste The licensee has been requested to provide an explanation for the observed occur-rence and evaluate any appropriate corrective actions. The l

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second concern relates'to the_ ready. acceptance'ofLthis

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deviationffrom.the. procedure.by the technicians.'(This is the'second of.three procedure c'ompliance-deficiencies dis-cussedJin this inspection. report). The' inspector was informed that all of the technicians were aware of this-occasional abnormality and-that since it was so easy t . verify the' contact status with~a volt-chm meter.no proce-dure change or technical ' review was necessary. These items will remain unresolved pending further_ licensee management and NRC review. (354/87-24-01) Maintenance Activities

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. Inspection' Activit 'Ou' ring this inspection period the inspector observed selected mainte-nance activities-on' safety related equipment to ascertain.thatLthese'

activities were conducted in.accordance with approved procedures, Technical Specifications, and' appropriate industrial codes and standards. These. inspections were conducted in accordance with NRC inspection procedure 6270 .2c Inspection Findings Portion.s of the~following activities were observed by'the l inspector: 'l

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-Work Order Procedure Descriptio DCR 4HC-0143 Recirculation i double valve remova '

870910116 DCP 4TMJ-86-264 ~ Install breech seals on H202 heat I trace panel j i

, 870831113 1C-GP.ZZ-008 Troubleshoot and l repair Al scram 1 relay !

871201057 MD-PM.PB-001 4.16 KV breaker maintenance and i cleanin . During the observation of preventative maintenance on 4.16 KV breakers the inspectors noted that a number of I supervisor hold points had been passed over. (This is the third procedure compliance deficiency discussed in this inspection report). When the technicians were questioned, j they stated that the supervisor had instructed them to

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T ignore these supervisor hold points. 'Upon further? discus-l sion it was determined that there was a misunderstanding-between the supervisor.and the technicians and that the supervisor hold points were required to be completed, cThe necessary: portions of the preventative maintenance activity were subsequently re performed and the' supervisor hold points correctly completed. The inspector was informed -

that-the maintenance supervisors were-counseled on the importance of properly completing supervisor hold point inspections and procedural complianc No violations were identifie . Engineered Safety Feature (ESF) System Walkdown 6.1 Inspection Activity The inspectors independent'y verified the operability of selected.ESF. ~

systems by performing a walkdown of accessible portions of the system to confirm that system lineup.pr.ocedures match' plant drawings and the as-built configuration. 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 ~

This inspection was condurted in accordance with NRC' inspection procedure 7171 .2 Inspection Findings The f'iltration, recirculation, and ventilation system (FRVS) was inspected and in plant conditions were.found to be acceptable. FRVS consists of two subsystems that are required to perform post-accident, safety-related functions simultaneously. The first subsystem, recir-culation system, filtera reactor building atmosphere to reduce offsite doses below 10CFR100 gJidelines in the event of high radioactivity in the reactor building. The second subsystem, ventilation system, performs a second filter process on reactor building atmosphere and maintains the reactor building at a negative pressure with respect to atmospheric condition The inspector verified that certain technical specification surveillance test requirements are satisfied through a review of the following procedures:

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OP-ST.GU-001 FRVS Monthly Operability Test

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OP-ST.GV-004 FRVS Initiation Test

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1C-FT.GV-001 FRVS Functional Test

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IC-CC.GU-008 FRVS - Division 4 Channel Calibration

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{ The inspector performed in plant valve position spotchecks to verify the accuracy of the FRVS tagging request and information system valve alignment worksheets. No valve position discrepancies were identifie On July 29, 1987, an unusual event was declared and a plant shutdown commenced after it was determined that the FRVS flow transmitters were previously calibrated using incorrect calibration data. The ;

incorrect and non-conservative calibrations would have prevented the FRVS recirculation fans from reaching the design flow rate of 30,000 -

SCFM per fan. The actual flow rates would have been limited to approximately 23,000 SCFM per fan. The root cause of the miscali-bration was that subsequent to startup testing, a Bechtel generated instrument calibration dcta (ICD) card was inadequately reviewed and substituted for the correct data determined during the startup test program. The FRVS ventilation fans were similarly affected although ,

their design flow rate is 9000 SCFM/ fan. During this inspection, t the inspector verified that all the recirculation and ventilation fan ICD cards had been updated in accordance with Field Directive N l H-1-GUXX-CFD-0487-0 which reflects the correct start-up calibration dat No violations were identifie . Safety Relief Valve Acoustic Monitors Since September, 1986, Hope Creek has experienced three separate failures of accelerometers used in the reactor vessel safety relief valve (SRV)

acoustic monitoring system. This system is designed to alert the control room to an open SRV by sensing the SRV tailpipe vibration which would be caused by the resulting steam flow. During the startup test program two accelerometers failed as evidenced by an erratic output and a higher than normal signal level. The vendor, NDT International Inc., reported that analysis of these accelerometers " .. revealed no internal component failures. Failure can only be determined to be due to a random error in the assembly process." During May, 1987, a third channel (accelerometer)

was suspected to have failed based upon its erratic and elevated noise level readings. The reactor was operating at power and in order to pre-vent a forced shutdown the licensee requested and received an Emergency Technical Specification change that allowed continued operation with one acoustic monitor inoperable until September 21, 1987. Other indications of an open SRV, that continued to be available in the control room, include tailpipe temperature, suppression pool temperature, main steam line flow, and turbine generator outpu The third accelerometer was removed during the September 1987 surveillance test outage and was found to rattle when shaken. After discussions with NDT it was determined that the entire lot supplied to the licensee could be subject to an inadequate assembly process. The possible deficiency related to the torque requirements of a set screw that holds two ceramic elements in place. If the set screw torque is inadequate the ceramic

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elements'will become loose-and result in an erratic and elevated outpu NDT also informed the licensee that their accelerometer supplier, Colum- 1 bia Research Laboratories,thad' revised the assembly procedure to 1nclude- l

insp'ection steps of proper torque values. -Based.upon the above findings,

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the licensee replaced the affected accelerometers with ones assembled-under the revised assembly procedure. The retest for replacement of th accelerometers involved lifting.each SRV individually in order to estab-lish a new valve-open baseline noise level. Although problems were encountered with a stuck open SRV during this test (See paragraph 3 of this report),ithe acoustic monitors operated properly and all setpoint calibrations were' complete '

On' September 28, 1987, NDT International, Inc. submitted a Part 21 report to NRC Region I. In this report, NDT states.that 14 accelerometers with -

specifically identified serial numbers purchased on purchase order number-10855-J-800 (Q)-AC~and shipped in October 1983, were possibly, subject;to'

improper torquing of the ceramic stack set screw. NDT is_not aware of any other similar failures and has concluded that the problem is-isolat'ed to the. specific lot supplied to Hope Creek. NDT has stated _that the utilities which have been supplied accelerometers will'be notified of the problems encountered atiHope Creek. Based upon the replacement of the subject accelerometers and successful verification of operability on October 10

.and 12, 1987, this Part 21 report is considered closed at Hope Cree No violations were identifie . Safety Review Committees Technical specifications require that the following review committees function to advise licensee management and provide'an independent safety -

review'for activities related to nuclear safety. The review' committees; are!the Station Operations Review' Committee (SORC) and the Nuclear Safety Review (NSR) group which consists of the onsite safety review group (SRG)

and the offsite safety review group (OSR).

The inspectors performed an inspection of the above review groups to evaluate implementation of their functions in accordance with Technical Specifications, ANS 3.1 (1981) and ANSI N18.7 (1976). The following paragraphs give an overview of the review groups functions and the inspectors finding .1 Station Operations Review Committee (SORC)

The'SORC is responsible for advising the General Manager - Hope. Creek Operations by written approval or disapproval of matters concerning nuclear safety. The inspectors periodically attend SORC meetings and review minutes of all SORC meetings. As a result of the above, '

the inspectors have verified that:

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The composition of the SORC conforms to Sections 6.5. and 6.5.1.3 of the Technical Specifications (T.S.).

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The 50RC meets a minimum of once per month as required by A minimum quorum of the Chairman and four members (including alternates) is present at the SORC meeting The SORC reviews required. documents and changes thereto and'

recommends approval or disapproval in writing to the General Manager - Hope Creek Operation The SORC maintains written minutes of each meeting and provides copies to the Vice President - Nuclear, General Manager - NSR, and Manager, OS The activities and responsibilities of SORC are implemented through Station Administrative Procedure (AP)- The ccmmittee reviews and concurs in all matters relating to the cause of a reactor trip and subsequent startu .2 Nuclear Safety Review (NSR)

The NSR is two review groups: one that consists of a General Manager and at least three dedicated, full time engineers located onsite, and the other includes a Manager offsite that is supported by at least four dedicated, full time engineer These organizations perform continuous evaluations and serve to advise the Vice President on matters of nuclear safet . Onsite Safety Review Group The Onsite Safety Review Group (SRG) functions to provide ,

a review of plant design and' operating experience for i potential opportunities to improve plant safety, to eval-uate plant operation; and maintenance activities, and to advise management on the overall quality and safety of plant operation The SRG makes recommendations for revised procedures, aquipment modifications, or other means of improving plant safety to appropriate station / corporate managemen The inspectors review the activities being performed by this group on a continuous basis and to date have reviewed all onsite safety review recommendations for 1987. The inspectors note that procedure changes and design changes have been instituted as a result of SRG findings.

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8. Offsite Safety Review Group (OSR)

c The offsite review group performs reviews of the ,

following:

Safety evaluations for changes to procedures, equip-

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ment, or systems and tests or experiments completed under the provisions of 10 CFR 50.59,'.to verify that such actions did not constitute an unreviewed safety questio Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in 10 CFR 50.5 Proposed tests or experiments that involve an unreviewed safety question as. defined in 10 CFR 50.5 Proposed changes to Technical Specifications or to the Operating Licens Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significanc Significant operating abnormalities or' deviations from normal and expected performance of plant equipment that affect nuclear safet All reportable event All recognized indications of an unanticipated deficiency in scme aspect of design or operation of safety-related structures, systems or component Reports and meeting minutes of the Station Operations Review Committe The offsite review group is also cognizant of the perform-ance of plant audits :onducted by the Nuclear Quality Assurance Department or an independent consultant. Audit results and recommendations are reviewed by OS The inspectors reviewed the documentation related to OSR activities itemized in Attachment I to this report to ascertain conformance with the abov l

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.As a result Uf the above reviews, the inspectors have deterdined d '

,,, that the licensee's review groups performance meet or pgce d Tech h '

nical Specification requirements. The review groups perform an f '

effective independent safety review and evaluation p t:Jon, and a produce useful corrective action recommendations. nnrougn) their g published meeting minutes, reports, open item tracking systems, api ,

verbal communication these groups keep the General Manager and Vice s ,

President apprised of matters]vertairing to nuclear safet >

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The 1.icensee submitted the following event' reports during the in wection' .

b period. Then event reports and periodical reports were reviewed for 1 accuracy and timeliness of submissio The asterisked ' report received additional followup by the inspector for corrective action ,it%1ementatio /i j-

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Monthly Operating Report for Septembe;L 1987 '

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LER 87-015-01 Auto-Isolation of the ' Control Room Ventilation System Caused by Spurious S.?nal from Radiation Monitodng System , '

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LER 87-039-00 Reactor Scram While Performing Tuitine Overspeed . oM[

OperabiDty Test Due to Pressure innsient in Turbine cp Electrohydraulic Control System -

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LER87-39describesareactorscramwhichoccurredwhilephformingahh . ;; ,

turbine overspeed protection operability tes '" !g Combined f ritermediate valve (CIV) No. 5 was being stroked closed when a transie d 'of undeter-mined origin in the turbine emeryncy trip system (ETS) csused the turbine control valves to fast-close, resulting in the reactor scram. The subject

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test was repeated at least six timee Aubsequent to the scram in an attempt to determine the failure mechan h i that caused thz transient in the ETS e header. The licensee was unable to recreate the tr%nsient, and thus based their corrective actions on industry experience withithe turbine electro-hydraulic control syste ,

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ETS pressure transients can be caused by an improperly seated disk dump valve or air entrapment in the disk dump cavity. To minimize the possi- J bility of this type of pressure transient, the licensee's corrective l action included implementation of a design change which installed

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restricting orifices in the feed ports.t.o the disk dump valve for all '

20 turbine valves. This restricting orifice tend.; to reduce pressure fluctuations in the ETS supply oil pressure, ,This design change was recommended by the Institute of Nuclear Power Operation for a similar problem at the James A. Fitzpatrick Electric Generating Statio t t

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1 Exit Interview The inspectors met with Mr. S. LaBruna and other licensee personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspection activitie ' '

g Based on Region I review and discussions with the licensee, it was

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determined that this report does not contain information subject to

, 10 CFR 2 restriction ,. Infcccement Conference On Oitober 19, 1987, an enforcement conference was held with the licensee at t.be ARC Region I Headquarters in King of Prussia, PA. The purpose of this inceting was to discuss an apparent Technical Specification viola-tien documented in NRC Inspection Report 50-354/87-19, and the licensee's

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management control of instrumentation isolation valves and equipment switches. The list of attendees and the handout provided by the licen-

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I: see are included as Attachment 2 to this report. At the conclusion of the meeting, Region I management stated that an enforcement action

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i decision would be transmitted by separate correspondenc E

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ATTACHMENT 1 .

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Offsite Review Group Records Reviewed (/

L OSR Procedures

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-M4l0-90P-01 1 Organization and

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M40-LEP-01 0 Technical Qualifications and Resources

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M40-LEP-02 0 Independent Safety Review Program

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M40-LEP-03 0 Audits

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M40-MSP-01 1 Offsite bufety Review -

Manual Preparation and Control Experience and Competence Matrix, Revision 10, May 1, 1987 f

, Staff Resumes t OSR Biweekly Reports to the Vice President ' OSR Review of the following documents: Procedures Changes

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- M6A / Design Changes ,

- 2EC-P.241

- t-HMJ-86-1256 '

- 4EC-1081

- 2EC-1680  :

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< Deficiency Report  ;

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- H1C-87-0126 1'

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I ATTACHMENT 2 October 19, 1987 Enforcement Conference Region I, King of Prussia, PA q List of Attendees NRC James Allan, Deputy Regional Administrator  ;

Samuel Collins, Deputy Director, Division of Reactor Projects Edward Wenzinger, Chief, Projects Branch No. 2, DRP Jay Gutierrez, Regional Attorney Paul Swetland, Chief, Projects Section 2B Richard Borchardt, Senior Resident Inspector, Hope Creek David Allsopp, Resident Inspector, Hope Creek Robert Summers, Project Engineer PSE&G Steven Miltenberger, Vice President, Nuclear Operations Stanley LaBruna, General Manager, Operations George Connors, Jr., Operations Manager Bruce Preston, Manager, Licensing and Regulation N.J. B. David Scott Dennis Zannoni

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