IR 05000237/1988005

Insp Repts 50-237/88-05 & 50-249/88-05 on 880130-0317. Violations Noted.Major Areas Inspected:Previous Insp Findings,Operational Safety Verification,Followup of Events, LER Followup,Info Notice Followup & Mgt Meetings

ML17199W331
Person / Time
Site:	Dresden
Issue date:	04/05/1988
From:	Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
References
50-237-88-05, 50-237-88-5, 50-249-88-05, 50-249-88-5, IEIN-87-0415, IEIN-87-415, NUDOCS 8804120378
Download: ML17199W331 (14)

Text

U. S. NUCLEAR REGULATORY COMMISSION REGION I II Report Nos. 50-237/88005(DRP); 50-249/88005(DRP)

Docket Nos. 50-237; 50-249 Licensee:

Commonwealth Edison Company P. 0. Box 767 Chicago, IL 60690 License Nos. DPR-19; DPR-25 Facility Name:

Dresden Nuclear Power Station, Units 2 and 3 Inspection At:* Dresden Site; Morris, IL Inspection Conducted:

January 30 through March 17, 1988 Inspectors: S. G. DuPont P. D. Kaufman Approved By:

M. A. Ring~ Chi~f~~-,

Projects Section *lB Inspection Summary Ins ection durin eport o..

Areas Inspected:

Routine unannounced safety inspection by the resident inspectors on previous inspection findings; operational safety verification; followup of events; licensee event reports followup; NRC Information Notice

• followup; TI-2515/90; management meetings; Chairman Zech visit; and report revie *

Results:

Of the 9 areas inspected, no violations or deviations -were idehtified in 7 areas; one violation was identified in the area of operational safety verification (failure to ensure compliance with a Technical Specification -

Paragraph 3b).

Additionally, one violation was.also identified in the area of licensee event reports; h6wever, in accordance with 10 CFR 2, Appendix C, Section V.A, a Notice of Violation was not issued (failure to verify the APRM inoperative function while in the Startup mode-Paragraph 5}. *

8804120378 880405 SDR ADOCK 05000237 DCD

• DETAILS Persons Contacted Commonwealth Edison Company

• E. Eenigenburg, Station Manager

• J. Wujciga, Production Superintendent

• C. Schroeder, Services Superintendent L. Gerner, Superintendent of Performance Improvement T. Ciesla, Assistant Superintendent - Planning D. Van Pelt, Assistant Superintendent - Maintenance J. Brunner, Assistant Superintendent -

T~chnical Services J. Kotowski, Ass1stant Superintendent - Operations R. Christensen, Unit 1 Operating Engineer G. Smith, Unit 2 Operating Engineer

• E. Armstrong, Regulatory Assurance Supervisor W. Pietryga, Unit 3 Operating Engineer J. Achterberg, Technical Staff Supervisor R. Geier, Q.C. Supervisor D. Sharper, Waste Systems Engineer D. Adam, Radiation Chemistry Supervisor..

J. Mayer, Station Security Administrator D. Morey, Chemistry Supervisor D. Saccomando, Radiation.Protection Supervisor

• E. Netzel, Q.A. Superintendent

R. Stols, Q.A. Engineer. *

The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, and contract security personne *Denotes those attending one or more exit intervi~ws conducted on March 17, 1988, and informally at various times throughout the inspection perio.

Review of Previous Inspection Items (92701 and 92702)

(Closed) Open Items (237/83007-01 and 249/83006-01):

Potential generic issue pertaining to the failure of Rancho Seco's primary containment personnel access interlock doors manufactured by Chicago Bridge and Iro An engineering review of the Rancho Seco event revealed that the interlock door failure was due to the el ectri ca 1 interlock not functioning as designe The review also indicated that the failure was not applicable to Dresden because the interlock at Dresden was mechanical and not electrical, and that no similar failure had occurred with the mechanical interlock The inspector revi~wed the engineering evaluation and agreed with the assessment. This item is considered to be close ** (Closed) Open Items (237/83029-02):

Potential generic problem found at Brown's Ferry Nuclear Power Station (BFNPS) with pin holes located in Rosemont Model 1153 Series B Alphaline pressure transmitter diaphragm This open item pertained to a 1982 finding at BFNP The inspector reviewed the maintenance history, through 1987, of pressure transmitters used ~t Dresden and did not find any occurrences pertaining to pin holes in the transmitter diaphragms. Since this problem has not bccurred, and the specific model noted in the potential generic problem is not predominately used at Dresden, this item is considered to be close Operational Safety Verification (71710 and 71707)

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period from January 30 through March 17, 198 The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected component Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenanc The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security plan. *

The inspectors observed p1ant housek~eping/cleanliness conditions and

• verified implementation of radiation protection controls. During the inspection, the inspectors walked down the accessible portions of various ECCS systems to verify operability by comparing system 1 ineup with plant drawings, as-built configuration or present valve lineup lists; observing equipment conditions that coµld degrade performance; and verified that instrumentation was properly valved, functioning, and calibrate These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR~ and administrative procedure On March 12, 1988, Dresden Units 2 and 3 exceeded the previous Commonwealth Edison (CECo) record for continuous dual unit operation of 141 days set by Zion and the CECo dual unit BWR operation of 131 days set by Quad Cities (exceeded on March 2)~ Additionally, Dresden Unit 3 has exceeded its best continuous single unit run of 138 days on February 23 and its best continuous days without a reactor scram of 152 days on February 27, 198 Unit 3 shutdown on March 26, 1988, for a 13 week refueling outag On February 21, 1988, while performing the Operating Surveillance DOS 300-1, "Control Rod Drive (CRD) Weekly Exercising," the Unit 2 Nuclear Station Operator (NSO) discovered that CRD E-9 was not disarmed electrically as required by Technical Specification CRD

• E-9 had been made inoperable previously on May 8, 1987, when the CRD had failed to couple to its associated control blade. Technical Specification 3.3.B.1.b requires an uncoupled CRD to be electrically disarmed to prevent CRD withdrawa On May 8, 1987, CRD E-9 was disarmed per Outage 11-1047 and out-of-service (OOS) tags were hung locally on E- *

However, on February 21, 1988; CRD E-9 was found to be electrically armed and no OOS tag was foun The operators immediately took corrective actions by disarming the CRD and rehanging the OOS ta Subsequently, the.licensee formed an investigative team to determine the root cause of the eventi The team interviewed personnel involved with the initial outage (May 8, 1987) and recent activities in the associated area of E-9, and reviewed surveillances, operating logs, computer logs, alarm *

histories and security records.. This review concluded that the CRD was electrically rearmed sometime between January 26, 1988, at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> and January 28, 1988, at 1106 hour0.0128 days <br />0.307 hours <br />0.00183 weeks <br />4.20833e-4 months <br /> The cause has been attributed to personnel error, since CRD E-9 was returned-to-service without proper authorization, though no specific individuals or departments could be identified as responsibl As a result of not following procedure OAP 3-5, "Out-of-Service and Personnel Protection Cards, 11 CRD E-9 was improperly returned-to-service. This action r~sulted in a.violation of Technical Specifi-cation 3.3.B.1.b, which requires that any CRD that is not coupled to its associated control blade be declared inoperable, fully inserted and electrically disarmed (237/88005-01; 249/88005-01).

A contributing factor, which may have led to this event, was

• inadequate measures to control OOS taggin The licensee's veri-fication process utilized to ensure that OOS tags remain properly placed requires improvement to include a more timely method of verifying that all accessible safety related OOS tags remain hung as require *

Although CRD E-9 was uncoupled and_ improperly electrically armed, it remained fully inserted throughout this event. Thus,. the possibi.lity of.this control blade causing a Rod Drop Accident was eliminate Subsequent corrective actions were designed to ensure that this was an isolated inciden The other HCUs were inspected for abnormalitie None were foun Fifty out of fifty-one 1988 outages and twenty-one of thirty-eight 1987 outages were physically verified throughout the plant with no abnormalities being identifie Radiological conditions prohibited verification of the remaining outages.

I In order to prevent recurrence, the event was discussed with all plant personnel via the weekly tailgate meeting These discussions stressed the proper actions to be taken upon observing an abnormal condition and included a reminder to follow Dresden Administrative Procedure (OAP) 3-5, "Out-of-Service and Personnel Protection. Cards."

DOS 010-13, Qu~rterly OOS Card Verifitation, was revised to require an audit of all safety related and Technical Specification related outages older than 90 day The licensee submitted Licensee Event Report LER 237/88004, in accordance with 10 CFR 50.7 One violation was identified in this are.

Followup of Events (92700)

During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.7 The inspectors pursued the events onsite with licensee and/or other NRC official In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriate actions, that activities were conductedwithiri regulatory requirements and that corrective actions would prevent future recurrenc The specific events are as follows:

• At 12:15 p.m. (CST) on February 11, 1988, the licensee declared an Unusual Event on Unit 3 and commenced a unit shutdown from 93% powe A Shift Foreman and Mechanical Maintenance Foreman, while performing an inspection of Core Spray Valve 1402-4A, noticed that the valve operator-to-valve-body yoke on the minimum flow valve (LPCI-1501-13A) for the "A" loop of the Low Pressure Coolant Injection (LPCI) system had several cracks. These cracks were analyzed by the licens~e to be of a degree that would render the valve inoperable. Since the 2/3 (Swing) Emergency Diesel Generator was out of service (inoperable) for an annual.inspection, the inoperable LPCI loop placed the Unit in an Unusual Even The licensee terminated the Unusual Event at 10:40 p.m., the same day, when the 2/3 Diesel Generator was returned to service at 10:35 Subsequently, the yoke on the LPCI 1501-13A valve was repair~d with a full penetration weld, replactng the previous fillet wel Following the repair, the valve was cycled and signatures taken before returning it to servic Once repairs were completed, the LPCI system pump test (DOS 1500-6) was performed for the A, B, C, *

and D pump The LPCI tests were satisfactorily completed and returned to service at 1:00 p.m., on February 12, 198 Cause of the yoke failure is still under review by the licensee and is being followed by a NRC Region III Materials & Processes Section Specialis At 10:40 a.m. (CST) on February 11, 1988, the licensee notified*

the NRC Duty Officer that the 20 Low Pressure Coolant Injection Pump {LPCI) was considered to be inoperable and that Unit 2 was in an Unusual Even The Unusual Event was declared because the

• 2/3 (Swing) Diesel Generator was also inoperable due to the annual inspection (inoperable on February 8, 1988).

The decision to declare LPCI inoperable at 10:30 a.m., was based upon further review of the maintenance performed on the 20 LPCI pump 4KV Breaker at 02:18 a.m., after successfully completing the required surveillance. The review revealed that post maintenance testing had been conducted on the replaced breakerJ which was taken from the 2B Shutdown Cooling Pump Motor, but no additional LPCI pump test surveillance had been conducte Since the licensee was not certain that the techni~al Specification requirements had been satisfied with only post maintenance testing of the 4KV breaker, the licensee declared an Unusual Event and began a unit shutdown*

from 90% powe Further review by the licensee and the NRC

.

determined that the actions were conservative and not required by 10 CFR 50.72.* At 11:55 a.m., the Unusual Event was terminated after successful completion of the Technical Specification LPCI Surveillance. Failure of the 20 LPCI pump breaker has been attributed to the trip latch roller which would not rotate. Since the trip latch roller did not rotate properly, the trip latch an roller were not in the necessary position to permit the trip latch to latch in and hold the open springs in compression, thus causing the mechanism linkage to collapse and the br~aker to re-op~n. The trip latch, trip latch roller, and linkage were thoroughly cleaned and lubricate The breaker was then tested successfully and installed in the 28 Shutdown Cooling Pump breaker cubicle on February 17, 198 No violations or deviations were identified in this are.

Liceniee Event Reports Followup (93702) *

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifi cations.:

{Closed) LER 249/88002-00:

Unit 3 Diesel Generator Auto Start During Work on Auto Start Relay Due to Management Deficienc During the installation of a missing cover on the Unit 3 Emergency Diesel Generator (EOG) output breaker cubicle by the Electrical Maintenance Department, on January 13, 1988, at 1940 hours0.0225 days <br />0.539 hours <br />0.00321 weeks <br />7.3817e-4 months <br />, the electricians inadvertently vibrated the relay mechanism causing the contacts to momentarily close and the EOG tQ start. Cause of this event was attributed to a management deficiency concerning work activity on in"."service equipmen As correc-tive action to prevent recurrence, the licensee's maintenance staff will issue a station policy statement to provide guidance for all working

• departments when work is to be performed on equipment that is in servic The policy will require that a line of communication exist between working departments and the operating shift on all equipment that is in service and suggest taking the equipment out-of-service or consider de 1 ayi ng the work activity until a unit outag Unit 2

. (Closed) LER 237/88002-00:. Fa{lure to Test the APRM Inoperative Function in the Startup Mode Due to Procedural Deficiency. * During the course of a Technical Specification adherence audit on February 1, 1988, the site Quality Assurance group identified a deficiency in the weekly*Average Power Range Monitor (APRM) surveillance procedure, Dresden Operating

.. Surveillance (DOS) 500-3, 11APRM Rod Block and Scram Functional Test.

The deficiency resulted in a. failure to verify the APRM inoperative function test whenever the procedure was.performed with the reactor

• in.the Startup mod Review df surveillance records for DOS 500-3 disclosed that the most recent instance of performing the surveillance in the Startu~ mode was on October 22, 1987, (Unit 2) and on July 16, 1987 (Unit 3).

The APRM inoperative function was tested satisfactorily for both reactors on January 28, 198 The cause of failure to test the APRM inoperative function while in the Startup mode was a procedural deficiency with DOS 500- The licensee issued*a temporary change to station procedure DOS 500-3 on March 3, 1988, to ensure that the APRM inoperative function is tested on a weekly basis independent of

_!(hether the reactor is in Startup or Run mod *

The failure to verify the APRM inoperative function while in the Startup mode is a violation of Technical Specification Table 4. (237/88005-02(DRP);* 249/88005-02(DRP)).

This violation meets the -

tests of 10 CFR 2, Appendix C, Section V.A; consequently, no Notice of Violation will be issued and this matter is considered close *(Closed) LER 237/88004-00:

Inoperable Control Rod Drive {tRD) E-9 Found Electrically Armed Due to Personnel Erro Review of this event is documented in Paragraph 3.b of this repor Two violations were identified in this are The violation associated with _LER 237/88004-00 is documented in paragraph 3.b of this repor The other violation associated with LER 237/88005 met the criteria of 10 CFR 2, Appendix C a~d no Notice of Violation was issue.

IE Information Notice Followup (92701)

The following IE Information Notice was reviewed by the Resident Inspectors to verify (1) that the Information Notice was received by licensee management, (2) that a review for applicability was performed, and (3) that if the Information Notice was applicable to the facility, applicable actions were taken or were scheduled to be take (Closed) IN_F87-41:

"Failures of Certain Brown Boveri Electric (BBE)

Circuit Breakers." Dresden does not utilize BBE circuit breakers; General Electric and Westinghouse circuit breakers are currently used at the station. The licensee review of NPRDS (Nuclear Plant Reliability Data System) failure record, regarding these breakers for Dresden Units 2 and 3 did not disclose any instances of circuit breaker malfunction due to loose or displaced mounting bolts. Based on this review, no further action is necessitated~ TI2515/90 Scram Discharge Volume Capability The numbering (04.01) of sections in this paragraph relates to applic-able sections of TI 2515/9.01 Scram Discharge Header Size (SDV) Criterion The scram discharge headers shall be sized in accordance with GE OER-54 and shall be hydraulically coupled to the instrumented volume(s) in a manner to permit operability of the scram level instrumentation before loss of system functio Actions taken at Dresden On March 18, 1982, Dresden responded to Mr. Harold R. Denton. *

(then Director, Office of Nuclear Re~ctor Regulation)

addressing the Long Term Program Criteria of the Generic SER of BWR SD The response addressed the adequacy of size of the SDV and instrument volumes (IV).

The SDV, including 41 gallons volume of the instrument volumes (IV), has the capability of receiving a maximum scram discharge of 800 gallons of water or 4.5 g~llons per drive. This analysi~ did not take credit fo~ the entire volume of the IVs (in excess of 90 gallons) to provide conservative assurance of scram capabilit _The calculated free volume also does not include any 1 inch vent line piping or 2 inch drain line piping.

.. The SDV sizing requirement of 3.34 gal ions per contro rod drive, as stated in GE DER 54 dated March 14, 1972, is based u~on th~ following:

control rod drive stroke time ten (10) seconds of scram breakage flow of 10 gpm (10 gpm X 10 sec/60sec/m = 1.667 gal/drive).

added volume in the SDV to limit the pressure rise to 50 psig during a scra The Dresden current volume of 4.5 gallons per drive exceeds the GE OER 54 3.34 gallons criteri.

• • In addition, the March 18, 1982, response also addressed the instrumentation redundanc The current Dresden design incorporated diverse and redundant instrumentation utilizing safety related Magnetrol float type water level switches and ITT Barton differential pressure water level transmitters used in conjunction with Rochester Instrument Systems current/voltage alarm Each of the two scram discharge volume~ has its own instrument volum Each IV has six water level sensing instrument One float type instrument annunciates that water is present in the IV; one float type prevents further control rod withdrawal (Rod Block); two safety related float switches and two safety related differential pressure water level transmitters automatically scram the reactor while sufficient water volume still exists in the SDV to ensure safe shutdow The IV scram instrumenta~ion incorporates *a 1 out of 2 twice logic with each of the two IV having the ability to independently scram the reacto.

Adequate hydraulic coupling was also assured by positively venting each SDV and IV to assure drainage in prepar~tion for scram reset. Additionally, because the IVs are located remotely from the SDV, each SDV has its own independent

• vent line and each of the IVs has its own independent drain lin *sased upon the Dresden response dated March 18, 1982, and closure of Supplements 1 through 3 of Bulletin 80~17 in inspection repbrts 237/80024 and 249/80028, this criterion (04.01) is considered close.02. Automatic Scram on High SDV Level Criterion Level instrumentation shall b~ provided for automatic scram initiation while sufficient volume exists in the SD Actions taken at Dresden In early 1982, Dresden modified the IV instrumentation to

~rovide diverse redundanc The scram logic was modified to be a 1 out of 2 twice logic with either of the IV having the capability to independently scram the reactor automatically on a water level less than 50 gallon The IV volume is capable of holding in excess of 90 gallons and the analysis, to ensure adequate scram volume, assumes only 40 gallons for the IVs in the calculated total volume of SDVs plus IVs at 800 gallons or 4.5 gallons per*driv **

Diversity is accomplished by using two float type sensors and two differential pressure sensors on each I.03 Instrument Taps Not on Connected Piping a~

Criterion Instrument taps shall be provided on the vertical 1V and not on the conne.cted piping.* Actions taken by Dresden The inspector reviewed the design drawings, Technical Specific-cation amendment requests, FSAR (updated) and visually verified that the IV instrument taps are connected to the 20 inch vertical Instrument Volume and not to the SDV/IV piping.*

04.04 Detection of Water in the IV Criterion The scram instrumentation shall be capable of detecting water accumulation i~ the IVs assuming a single active failure* in the instrumentation system or the plugging of an instrument *

line Actions taken at Dresden The scram instrumentation is set up with 1 out of 2 twice logi The *logic is divided into two channels (A & B)

containing two relays in each (Channel A - relays A & C, Channel B - relays B & D).

Each relay has two associated diverse sensors (one float and differential pressure type).

One relay from each of the channels (etther relay A or C in Channel A and either relay B or D in Channel B) will initiate an automatic scram if de-energize Each relay is de-energized if either of its sensors indicate water level in the IVs at or greatei thari 50 gallon The sensors are arranged such that, if one of the 2 inch instrument lines associated with an IV is blocked, only one of the float and one of the differential pressure type sensors will be inoperative. This would not degrade the protection logic since the other sensors associated with the affected relays are from the opposite I Additionally, if a generic failure of either type of sensor (float or differential preisure) would make all four associated sensors inoperative (on both IVs), automatic scram capability would still be provided by the other type of sensor. This combination of protection logic redundancy and diversity of type of sensors would still provide automatic scram capability with a complex 1 failure of complete blockage of one IV and a generic failure of either type of sensor, in that the two remaining sensors on the operable IV would still provide an automatic scra.05 Vent and Drain Valve System Interfaces Criterion Vent and drain functions shall not be adversely affected by other system interfaces. The objective of this requirement is to preclude water backup in the scram IV, which would cause a spurious scra Actions taken by Dresden*

The Dresden March 18, 1982, response ~ddressed the design specifications for the vent and drain functio The SDV header piping is pitched in such a manner as to continuously gravity drain water to the instrument volum All. vent and drain lines have been installed free of any loop seals to ensure venting and drainag The analysis was also verified to have addressed water backup into the IV through the drain line Each of the IVs have an independent drain line feeding to the reactor building equipment drain tank located approximately 15 feet below the IV elevation. The reactor building equipment tank

• overflow hatch is open to the atmosphere and would empty into the surrounding room rather than feedback to each I.06 Vent and Drain Valves Close on Loss of Air Criterion The power-operated vent and drain valves shall close under loss of air and/or electric powe Valve position indication shall be provided in the control roo Action taken at Dresden All vent and drain valves on the SDV and IVs at Dresden are air operated such that they will fail closed on loss of air. Each valve has control and position indication on the control room 90x-5 pane.07 Operator Aid

• . Criterion Instrumentation shall be provided to aid the operator in the detection of water accumulation in the IVs before scra Actions taken at Dresden The inspector verified that each of the IVs has two Magnetrol water level switches to detect water accumulation prior to scram initiation. One instrument annunciates in the control room to indicate water accumulatio If water accumulation continues, the second instrument will block further control rod withdrawa *

04.08 Active Failure in Vent and Drain Lines Criterion Vent and drain line valves shall be provided to contain the scram discharge water with a single active failure and to minimize operational exposur Actions taken at Dresden The inspectors verified that operational exposure is minimized by. posting and restricting access to the vent and drain line Additionally, redundancy is pro~ided to assure the single active failure criterion is satisfied, with independent vents on both SDVs and independent drains on both IV.0~ Periodic Testing of Vent and Drain Valves Criterion Vent and drain valves shall be periodically tested:.Actions taken at Dresden The inspector verified that the Scram Discharge System vent and drain valves are periodically tested by reviewing completed Dresden Operating Surveillances DOS 1600-18, Cold Shutdown Valve Testing and DOS 1600-1, Quarterly Valve Timin Addi-ti6nal ly, the inspector verified acceptable stroke time criteria of 30 seconds or less per Technical Specification 4.3..10 Periodic Testing of Level Detection Instrumentation Criterion Level detection instrumentation ano verifying level detection instrumentation shall be periodically tested in plac Actions taken at Dresden The inspector verified that the IV water.level detection instrumentation is functionally tested quarterly and calibrated

once per refueling outage, in accordance with Technical Specification Tables 3.1.1, 4~1.1 and 4.1.2, by reviewing completed Dresden Instrumentation Surveillances DIS 500-5,

"Unit 2 Scram Discharge Volume Level Sensor Calibration and Functional Tests," and DIS 500~10, "Unit 3 Scram Discharge Volume Level Sensor Calibration and Functional Tests.

Included in the review of DIS 500-5 and DIS 500-10, was the verification that proper restoration of the system configura-tion is clearly prescribed by the procedure.11 Periodic Testing Operability of the Entire System Criterion*

The operability of the entire system as an integrated whole shall be demonstrated periodically and during each operatin cycle by demonstrating scram instrument response and valve function at pressure and temperature at approximately 50%

control rod densit Actions taken at Dresden The inspectors verified that an integrated test involving a scram at 50% contro.l rod density at pressure and temperatur~

does not exist, nor is such a test required by Technical Specifications. A review of Technical Specifications revealed that a surveillance.does exist for Dresden and LaSalle which verifies vent and drain va 1 ve operability, but there is no associated Limiting Condition for Operation (LCO).

This issue is being discusse9 with NRR and is considered an open item (237/88005-03; 249/88005-03). Management Meetings (30702)

On March 3, 1988, an Enforcement Conference was held with Cormnonwealth Edison Company staff at the NRC Region III Office in Glen Ellyn, Illinois. The meeting was to discuss 10 CFR 50 Appendix J Local Lea Rate Testing (LLRT) ~xceeding the 2 year testing interval at Dresden and applicability to other Cormnonwealth Station Inspection details *

of the Unit 3 Dresden event are addressed in Region III Inspection Reports 50-Z37/87040; 50-249/8703.

Cormnissioner Visit On February 4, 1988, Chairman Lando W. Zech, Region III Regional Administrator, members of the regional staff and resident inspectors toured the Dresden Nuclear Station. While on site, the Chairman met with the NRC resident*staff, as well as with licensee management and supervisory personne The facility tour included the Units 2 and 3 reactor buildings, turbine buildings, and control room with observation of shift operating crew performanc The Chairman emphasized that he believes in a clean, quiet, and taut plan He also stressed that the control room was in need of improvements in sound reduction and recognized that a program is in place to upgrade the control roo.

Report Review (90712)

During the inspection period, the inspectors reviewed the licensee's Monthly Operating Report for January and February, 198 The inspectors confirmed th.at the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.1.

Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or bot An open item disclosed during this inspection is discussed in Paragraph.

Violations Foi Which a Notice of Violation" Will Not Be Issued The NRC us~s the Notice of Violation as a standard method for iormalizing the existence of a violation of a legally binding requiremen However, because the NRC wants to encourage and support a licensee's initiatives for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V. These tests area:

(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level *iv or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reason~ble time period; and (5)

it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violatio A violation of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued is discussed in paragraph.

Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1)

on March 17, 1988, and informally throughout the inspection period, and summarized the scope and findings of the inspection activitie The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspectio The licensee did not identify any such*

documents/processes as proprietary. The licensee acknowledged th findings of the inspectio