IR 05000277/1993011

Insp Repts 50-277/93-11 & 50-278/93-11 on 930511-0614.No Violations Noted.Major Areas Inspected:Plant Operations, Engineering & Technical Support Activities,Surveillance & Maint & EP

ML20045E764
Person / Time
Site:	Peach Bottom, Reed College
Issue date:	06/25/1993
From:	Anderson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20045E760	List: IR 05000277/1993011
References
50-277-93-11, 50-278-93-11, NUDOCS 9307060021
Download: ML20045E764 (13)

Text

.

-

.

.

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report No.

50-277/93-11 License Nos. DPR-44 50-278/93-11 DPR-56 Licensee:

Philadelphia Electric Company Peach Bottom Atomic Power Station P. O. Box 195 Wayne, PA 19087-0195 Facility Name:

Peach Bottom Atomic Power Station Units 2 and 3 Dates:

May 11 - June 14,1993 Inspectors:

B. S. Norris, Acting Senior Resident Inspector F. P. Bonnett, Resident Inspector R. K.. Lorson, Resident Inspector

'

J. J. Lyash, Senior Resident Inspector f.

-

In b-Approved By:

C. J. An'ddrsonfChief Date Reactor Projects Section 2B Division of Reactor Projxts

,

9307060021 930628 PDR ADOCK 05000277 G

PDR

.

.

..

.

.

EXECUTIVE SUMMARY Peach Bottom Atomic Power Station Inspection Report 93-11

'

PLANT OPERATIONS Unit 2 operated at essentially 100% power for the entire period. Unit 3 began at 100% power; early in the period, an increase in off-gas activity led to the determination that an existing fuel leak had degraded. To minimize the effects of the fuel leak, adjacent rods were inserted,

'

thereby reducing the maximum allowable reactor power. During the report period, Unit 3 started a Technical Specification required shutdown when an emergency diesel generator and a redundant loop of low pressure coolant injection were both inoperable at the same time. The inspector determined that the licensee appropriately initiated the shutdown in accordance with the Technical Specification. The licensee acted promptly and effectively to troubleshoot and correct the deficiencies, thereby terminating the shutdown However, it was noted that the diesel generator deficiency had been previously identified on a nonconformance report. The inspector reviewed the NCR backlog and noted a large number awaiting disposition or completion of work. This is an Unresolved Item.

During the period, the Plant Manager announced his resignation from Philadelphia Electric

'

Company; his successor will be the previous Director, Site Engineering.

ENGINEERING AND TECIINICAL SUPPORT ACTIVITIES During this inspection, an existing fuel leak on Unit 3 degraded, causing offgas activity to increase. The reactor engineers immediately recognized the potential problem with the fuel.

The licensee inserted control rods in the immediate area of the fuel leak, minimizing the growth of the fuelleak, but establishing an asymmetric rod pattern. As of the end of the period, the licensee was still discussing contingency plans regarding whether to continue operating Unit 3 with the fuel leak or to shut down and replace the leaking fuel assemblies. The reactor engineering group effectively performs its role in assuring safe operation of the reactor. The group has an excellent understanding of core management and safety, and the training program is thorough and comprehensive.

SURVEILLANCE & MAINTENANCE During the period, a bus fast transfer caused the isolation of some of the Unit 3 feedwater heaters, resulting in a minor reactivity addition.

The inspector reviewed the licensee's investigation and was satisfied with the identified root causes, and the immediate and proposed corrective actions.

ii

.

.

EMERGENCY PREPAREDNESS The licensee conducted an emergency preparedness training exercise in preparation for the NRC graded exercise scheduled for late June. The appropriate facilities and personnel were activated and involved in the exercise.

.

b

n i

iii

..

.

.

.

.l

-

i TABLE OF CONTENTS Page EXECUTIVE SUMMARY

.......................................il 1.0 PLANT OPERATIONS REVIEW

.............................

2.0 FOLLOW-UP OF PLANT EVENTS............................ 1 3.0 ENGINEERING AND TECHNICAL SUPPORT ACTIVITIES............ 3 3.1 Reactor Engineer Activities............................. 3 3.2 PECo Response to GL 92-04 Concerning Reactor Vessel Water I.cVel...

4.0 SURVEILLANCE TESTING OBSERVATIONS..................... 6 5.0 MAINTENANCE ACTIVITY OBSERVATIONS.................... 6 6.0 EMERGENCY PREPAREDNESS

.............................

7.0 RADIOLOGICAL CONTROLS............................... 7 8.0 PHYSICAL SECURITY..................................

'

9.0 LICENSEE EVENT REPORT UPDATE......................... 7 10.0 PREVIOUS INSPECTION ITEM UPDATE....................... 8 11.0 M ANAGEMENT MEETINGS............................... 9

,

I iV

. -.

--

.

-

- -

- -

- -

- - - - -

- -

- -

-

-

.

.

DETAILS 1.0 PLANT OPERATIONS REVIEW (71707)*

The inspectors completed NRC Inspection Procedure 71707, " Operational Safety Verification,"

by directly observing safety significant activities and equipment, touring the facility, and interviewing and discussing items with licensee personnel. The inspectors independently verified -

.

safety system status and Technical Specification (TS) Limiting Conditions for Operation (LCO),

reviewed corrective actions, and examined facility records and logs. The inspectors performed 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of deep backshift/ weekend tours of the facility.

The licensee operated Unit 2 at essentially 100% power for the entire period. While replacing the control switch for the E-22 breaker, an automatic bus fast-transfer caused the isolation of the

'B' feedwater heaters, resulting in a minor positive reactivity transient. The temporary loss of power also caused a primary containment isolation system (PCIS) Group II outboard isolation.

The operating staff responded promptly and in accordance with plant procedures (see Section 5.0).

Throughout the period, the inspectors observed that control room operators and supervision maintained very good oversight of activities and responded appropriately to equipment problems.

Unit 3 began the inspection period at 100% power. The unit operated without any major transients or engineered safeguard feature (ESF) system actuations. Early in the period, the licensee noted an increase in off-gas activity levels following rod pattern changes. The licensee performed flux tilt testing, verifying that an existing fuel leak had degraded. To minimize the effects of the fuel leak, the licensee fully inserted an adjacent control rod resulting in an asymmetric rod pattern and a reduction in the maximum allowable reactor power.

The inspectors are continuing to closely monitor this condition (see Section 3.1).

On June 8,1993, Ken Powers, the Plant Manager - Peach Bottom, announced his resignation from Philadelphia Electric Company (PECo) effective June 25,1993. His replacement will be Garrett Edwards, the current Director, Site Engineering.

2.0 FOLLOW-UP OF PLANT EVENTS (71707,93702)

During the report period, one plant event occurred. The inspectors evaluated licensee staff and management response to the event to verify that the licensee had identified the root cause, implemented appropriate corrective actions, and made the required notifications. On May 23, 1993, at 11:15 a.m., the licensee started to shutdown Unit 3, as required by TS 3.0.D.

Specifically, the E-2 emergency diesel generator (EDG) was inoperable (TS 3.9. A.2), due to low lube oil temperature; and the '3A' loop of the low pressure coolant injection (LPCI) system was inoperable (TS 3.5.A.3), due to the loss of normal electrical power supply to the swing bus for two LPCI system valves. The licensee terminated the shutdown at 12:28 p.m., after the lube The inspation procedure from NRC Manual Chapter 2515 that the insp& tors used as guidance is

parenthetically listed for each report sectio _

_

-_

_

.

_

m

.

P

oil temperature condition cleared and the E-2 EDG was declared operable. The Shift Manager notified the NRC via the Emergency Notification System. The licensee initiated a reportability evaluation / event investigation form (RE/EIF) to investigate this event, this investigation was not completed by the end of the reporting period.

The cause of the low lube oil temperature was attributed to the loss of the lube oil circulating pump as a result of the tripping of the pump's thermal overload circuit. The thermal overload contacts were reset and the lube oil temperature subsequently returned to the normal range. A failure of the voltage sensing relay for the swing bus resulted in the normal power supply being removed from the LPCI valves. The individual valves affected were the 'A' LPCI inboard injection valve (MO-3-10-25A) and the 'A' LPCI pump discharge valve (MO-3-02-53A). The licensee restored power to the valves by manually transferring the swing bus to the alternate power supply. The relay was replaced and the normal power supply to the LPCI swing bus was restored. The inspector determined that the licensee appropriately initiated the shutdown in accordance with the technical specification. The licensee acted promptly and effectively to troubleshoot and correct the deficiencies.

During the review of the EDG circulating lubricating oil pump trip, the inspector noted that the licensee issued a nonconformance report (NCR) in March 1991 to investigate a recurring problem involving tripping of the lubricating oil circulating pump thermal overload contacts for the E-3 EDG. The initial disposition of this NCR was to replace the thermal overload heater elements with larger size elements. This change was in accordance with the thermal heater vendor's catalog, and designed to compensate for the ambient temperature differential between the pump motor and the heater elements. The scope of this NCR was expanded in July 1992, to include the lubricating oil circulating, pre-lubrication oil and jacket coolant pumps for all.four i

EDG's. This revision resized the thermal overload heater elements for the above pumps. The repair disposition of this NCR authorized replacement of the existing thermal overload heater elements with larger size elements. The licensee utilized this NCR disposition to replace the E-2 EDG thermal overload heater elen'ents with larger elements following the May 23 trip.

With respect-to the LPCI swing bus failure, the inspector interviewed an electrical system manager who indicated that this type of voltage sensing relay had previously failed three time, in the LPCI swing bus application. The first relay failure occurred in October 1991, and ' ae licensee initiated a NCR to evaluate this failure. The disposition of this NCR was based e i an earlier engineering calculation and concluded that use of this relay was acceptable. A 5 rond relay failure occurred in January 1992 and the licensee initiated a RE/EIF and conducti ; a test study to determine the root cause of the relay failure. This test study was unable to 6termine the exact root cause for the failure. The failed relay was examined and the licensee's review indicated that the relay failure was due to excessive long term heating. This relay failed a third time (October 1992) and the licensee issued a NCR to correct this recurring problem. The NCR identified the need to replace this relay with a higher environmentally rated relay. The electrical

system manager indicated that these relays were to be replaced during system outages and after failure. Following this recent failure (May 1993), the licensee replaced this relay with a higher environmentally rated as planned.

l

,

l

_

.-

-.

--

.

In both cases, the inspector concluded that the licensee's short term corrective actions were adequate. However, the Inspector is concerned with the length of time to complete the repairs associated with the EDG NCR. Additionally, the inspector will evaluate the licensee's RE/EIF findings.

The inspector reviewed the NCR backlog to determine if the concern discussed above was unique or indicative of a general weakness in the licensee's ability to process and complete NCR's. The inspector noted that the backlog of NCR's awaiting disposition or requiring work is relatively large. The inspector interviewed managers in both the Site Engineering and Maintenance Departments and concluded that the licensee was taking action to reduce the backlog through management awareness, increased engineering resources, a pending enhance-ment to the NCR procedure, and an improved NCR tracking capability. This item will remain open pending further review of the licensee's effectiveness at reducing the NCR backlog (Unresolved Item 50-277/93-11-01 and 50-278/93-11-01).

3.0 ENGINEERING AND TECIINICAL SUPPORT ACTIVITIES (37700, 40500, TI-2515/119)

The inspectors routinely monitor and assess licensee support staff activities. During this inspection period, the inspectors focused on activities performed by the Reactor Engineers and the licensce's response to Generic Letter 92-04. The results of these reviews are discussed in detail below.

3.1 Reactor Engineer Activities Over the course of this inspection period, the inspector evaluated the effectiveness of licensee's Reactor Engineers (REs). The purpose of the inspection was to assess the affect that the recent reorganization had on the effectiveness of the REs; i.e., the training program, their technical knowledge, and their ability to monitor and trend reactor performance.

In March 1993, the licensee reorganized the Technical Section into the Site Engineering Section.

The purpose of the reorganization was part of the Nuclear Effectiveness and Efficiency Design

'

Study (NEEDS). The Site Engineering Section is lead by the Director Site Engineering who reports to the Site Vice President. The Site Engineering Section is composed of four branches, each having a Senior Manager who reports to the Director Site Engineering. The Reactor Engineering Group is one of four groups that is a part of the Plant Engineering Branch.

.

The RE Group has a staff of nine engineers. Two REs are fully qualified to perform all of the duties of an RE and the remaining seven are partially qualified. In addition to the REs assigned to the group, there are four fully qualified REs that were reassigned to other organizations at Peach Bottom as part of NEEDS, and five Shift Engineers. Of the Shift Engineers, two are fully qualified, two are partially qualified, and one is newly assigned. There are effectively eight fully qualified REs and ten partially qualified REs on site. The experience level for the_

REs varied from one week to over ten years, with most being over five years. The RE

l

-_

-

-

.

-

-

-

-

- -- - - -

- L

- -. -

.

.

Procedures defines the roles and responsibilities of the REs at Peach Bottom. The RE is responsible for specifying control rod movements and reactivity management, to maintain the fuel within thermal limita. They monitor core performance, maintain the process computer software, and are the Sy stem Managers (SM) for the primary reactor systems that directly affect core reactivity.

The inspector reviewed the training and qualification manual used by the partially qualified REs.

The training starts with five weeks of classroom training, based on the General Electric " Station Nuclear Engineers Manual (NEDO-24810C)." The trainee then demonstrates, to a qualified RE, the ability to perform the tasks listed in the qualification manual. After the task is completed satisfactorily, the trainee is permitted to perform that function alone in the control room. The trainee is given an oral examination when the qualification manual is completed.

The inspector interviewed the Director Site Engineering and the Senior Reactor Engineer regarding management's e-xpectations of the RE. Licensee management has communicated that the REs must conservatively apply the principles for safe reactivity management, be able to recognize adverse situations quickly that could cause a transient, and have a questioning attitude and good judgement. Their philosophy is that the SM is the individual responsible for oversight of the system, establishing priorities, and providing direction to others with respect to these priorities.

The inspector interviewed five REs to determine how they accomplish the responsibilities outlined in the RE procedures and to evaluate their effectiveness. During these interviews, the inspector focused on workload, general system knowledge, performance of job function, and training. In addition, the inspector observed a number rod pattern adjustments, and TIP system operations. The REs are knowledgeable and capable of performing their assigned tasks. They demonstrated their abilities during their assessment of the increasing offgas levels on Ur'.t 3 (see following paragraph).

During this inspection, an existing fuel leak on Unit 3 degraded, causing offgas activity to increase. The REs immediately recognized the potential problem with the fuel. The Senior RE contacted the Fuel Management Group, at Chesterbrook, and General Electric (GE). The REs and Fuel Management group have been assessing the fuel condition daily, via coolant samples and offgas calculations. The licensee inserted control rods in the immediate area of the fuel Icak, thus minimizing the growth of the fuel leak. Due to early detection and mitigating actions, the licensee has been able to maintain the leak small. Although no safety issue presently exists, the fuel is continuing to degrade. The inspector discussed the core condition, the failure mechanism, and the strategy to continue plant operations with the licensee. The inspector -

attended various management meetings where the licensee discussed contingency plans regarding whether to allow the unit to continue operation until the next planned outage, or to shutdown early and replace the leaking fuel assemblies. The licensee is actively addressing this problem.

The inspector will continue monitoring licensee actions and developing information concerning reactor conditions.

-

-

-

-

_ _ _ _ _ _ _

.

Based on this inspection, the inspector concluded that the RE Group effectively performs its role in assuring safe operation of the reactor. The REs provice valuable technical and operational support. The REs and Fuel Management Group have an excellent understanding of core management and safety. The group has not been adversely affected by the recent reorganization within PECo. The training program is thorough and comprehensive.

3.2 PECo Response to GL 92-04 Concerning Reactor Vessel Water level The inspector reviewed the licensee's implementation of actions taken in response to NRC Information Notice (IN) 92-54, " Level Instrumentation Inaccuracies Caused By Rapid Depressurization," and NRC Generic Letter (GL) 92-04, " Resolution of the Issues Related to Reactor Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)." During accident conditions, a sudden uncontrolled depressurization of the primary coolant system could cause a loss of reference leg inventory; this would result in a false indication of high water level in the reactor sessel. This inspection focused on verifying the adequacy of operator training in this area, ensuring that the required actions could be implemented; and ensuring the training was consistent with the plant's Emergency Operating Procedures (EOPs).

The inspector reviewed lesson plans, training records, required reading material, observed a shift crew during a simulator training scenarios, and conducted interviews. In the respective lesson plan, the inspector noted that appropriate industry and NRC guidance was incorporated. The inspector also determined that the simulator exercise guide satisfied all of the objectives contained in the Boiling Water Reactor Owner's Group letter, dated October 16, 1992. A required reading package, issued in August 1992, alerted operators to the problem; the package incorporated site speciic and generic guidance. During the simulator training scenario, a rapid depressurization resulted in an indeterminate reactor water level indication. The operating crew recognized the problem and took the appropriate actions, as directed by the EOPs. The inspector concluded that the training conducted regarding this issue had been effective and that the operators were capable of recognizing and responding to this type of event. The inspector reviewed the EOPs and determined that they were consistent with industry guidelines. The EOPs clearly directed the operators to implement the reactor flood procedure if the reactor

!

vessel level was unknown.

The inspector also reviewed design features associated with the safety parameter display system (SPDS) to determine if the system would provide additional information to assist the operator in recognizing this event. The SPDS displays an average of the two wide range water level instrument readings. A redundant point check compares each of the indicated level readings to ensure that the difference is within a specified tolerance. If the tolerance is exceeded, the system output is color coded to afst the operator that a problem exists with the reading. As stated in GL 92-04, "an abrupt dep;essurization event resulting in a common mode, common magnitude level indication error is unlikely," thus the erroneous level readings following a depressurization event would be unlikely to satisfy the redundant point check. The SPDS provides another means of alerting the operator to a divergent reactor water level condition.

- _ _ _ - _ _ _ _ _ _ _ -

..

..

. - _ - - _

.

.

<

Based on the discussions with the licensed operators, reviews oflesson plans and procedures, and the design features of the SPDS, the inspectors are satisfied that the operating shift can effectively mitigate the consequences of a reactor water level instrumentation inaccuracy.

4.0 SURVEILLANCE TESTING OBSERVATIONS (61726,71707)

The inspectors observed conduct of surveillance tests to determine if approved procedures were used, test instrumentation was calibrated, qualified personnel were performing the tests, and test acceptance criteria were met. The inspecors verified that the surveillance tests had been prop-erly scheduled and approved by shift supervision prior to performance, control room operators were knowledgeable about testing in progress, and redundant systems or components were available for service, as required. The inspectors routinely verified adequate performance of

~

daily surveillance tests including instrument channel checks, and jet pump and control rod opera-bility. The inspectors found the licensee's activities to be acceptable.

5.0 MAINTENANCE ACTIVITY OBSERVATIONS (62703)

The inspectors observed ongoing maintenance activities to verify proper implementation of procedures and controls. The licensee adequately implemented administrative controls for blocking permits, fire watches, ignition sources, and radiological controls. The inspectors reviewed maintenance procedures, action requests, work orders, item handling reports, radiation work permits, material certifications, and receipt inspections.

During observation of maintenance work, the inspectors verified appropriate Quality Verification involvement, plant conditions, TS LCOs, equipment alignment and turnover, post-maintenance testing, and reportability review. The inspectors found the licensee's activities to be acceptable.

On May 19,1992, while replacing the control switch for the E-22 breaker, a bus fast transfer occurred due to a trip of the normal power supply breaker (E-322). This resulted in a momentary loss of power to the E-22 bus, which caused Unit 2's 3B, 4B, and 5B feedwater heaters to isolate. Isolation of these heaters resulted in a minor positive reactivity addition transient. The operating staff's response to mitigate the transient was prompt and in accordance with plant procedures. This loss of power caused a PCIS Group II outboard isolation. The inspector concluded that the licensee's response was appropriate.

The licensee conducted an investigation of this event and determined that the probable cause of the event was an inadvertent shorting of the terminal lugs for the indicating light for the E-322 breaker. Shorting of these lugs would insert a trip signal to the E-322 breaker and cause a fast transfer to initiate. The E-322 breaker is physically close to the E-22 breaker control switch, and the licensee determined that this may have contributed to this event. The inspector interviewed the maintenance electrician who performed this activity, and the licensee personnel who performed the event investigation, and reviewed the associated electrical schematic diagram, and concluded that the licensee's analysis of this event was sound. The inspector also reviewed the licensee's draft investigation report and was satisfied with the immediate and proposed corrective actions.

_ _ _ _ _ _ - _ _

.

.

6.0 EMERGENCY PREPAREDNESS (80301)

The inspectors observed an emergency preparedness training exercise conducted by the licensee on June 2. The exercise utilized the simulator control room, technical support center (TSC),

operations support center (OSC), and emergency operations facility (EOF). The inspectors noted that the appropriate license personnel were actively involved in the exercise. The licensee's post-drill critique was critical of the actions taken during the exercise and noted areas where improvement was needed.

7.0 RADIOLOGICAL CONTROLS (71707)

The inspectors examined work in progress in both units to verify roper implementation of health physics (HP) procedures and controls. The inspectors monitored the ALARA (As Low As Reasonably Achievable) program implementation, dosimetry and badging, protective clothing use, radiation surveys, radiation protection instrument use, handling of potentially contaminated equipment and materials, and compliance with RWP requirements.. The inspectors observed that personnel working in the radiologically controlled areas were meeting applicable requirements and were frisking in accordance with HP procedures. During routine tours of the units, the inspectors verified a sampling of high radiation area doors to be locked, as required. All activities monitored by the inspectors were four.d to be acceptable.

8.0 PHYSICAL SECURITY (71707)

The inspectors monitored security activities for compliance with the approved Security Plan and associated implementing procedures. The inspectors observed security staffing, operation of the Central and Secondary Access Systems, and licensee checks of vehicles, detection and assessment aids, and vital area access to verify proper control. On each shift, the inspectors observed protected area access control and badging procedures. In addition, the inspectors routinely inspected protected and vital area barriers, compensatory measures, and escort procedures. The inspectors found the licensee's activities to be acceptable.

9.0 LICENSEE EVENT REPORT UPDATE (92701,92702)

In addition to the events discussed in section 2.0; the inspectors reviewed Licensee Event Reports (LERs) submitted by the licensee during the period for events which were less safety significant, and did not warrant immediate review and evaluation by the inspector at the time of the event. The inspector reviewed the following LERs and found that the licensee had identified the root causes, implemented appropriate corrective actions, and made the required notifications.

LER No.

LER Date LER Title 2-93-008 4/3/93 Technical Specification Violation - Fire Suppression Carbon Dioxide Storage Tank Pressure Below Minimum Limit.

i

=

J

i

.

-

,

2-93-009 4/22/93 Technical Specification Violation - Containment Sump Data was not Recorded during Surveillance Test.

10.0 PREVIOUS INSPECTION ITEM UPDATE (92701,92702)

(Update) URI 92-32-02, Operability Determinations During Surveillance Testing

,

The inspector identified that the loss-of-coolant-accident (LOCA) automatic initiation signal for two EDGs was inoperable during the performance of the Core Spray Logic System Functional Test (LSFT). At the close of inspection period 93-01, the licensee had committed to analyzing

,

the operability of the EDGs themselves with respect to the design basis function of the EDGs.

On March 12,1993, the licensee adopted Plant Operations Review Committee -(PORC) position 61. PORC-61 stated: during surveillance tests that cause a system to be inoperable as a result

,

of a system lineup, the system shall be considered inoperable if the system cannot meet its intended safety function. The PORC position further stated that, if a related system is affected by a particular surveillance test, but is still able to perform its intended safety function, then the related system need not be considered inoperable.

In the PORC position, the licensee specifically addressed the Core Spray LSFT affect on the EDGs. The licensee adopted the position that, because the EDG loss of bus power start feature was unaffected by the LSFT, then the EDG was capable of meeting its design basis function.

According to the licensee, the loss of power start feature represented diversity and that it was not necessary to consider a single failure to the diverse loss'of power start feature at the time the LOCA start signal was blocked for test.

The inspectors, with assistance from the NRR technical staff, reviewed the design of the EDG start circuitry and the Peach Bottom Updated Final Safety Analysis Report for information on the facility design basis. The staff noted that it is the MCA relay sequence that is blocked by the Core Spray LSFr. The MCA relay is designed to start the EDG on a LOCA as an anticipatory function. A loss of offsite power (LOOP) initiates the actual start signal and sequence relays that cause the EDG output breaker to close onto the dead bus. The Peach Bottom design basis accident assumes a concurrent LOCA and LOOP. The staff recognized that a simultaneous LOOP /LOCA is the design basis accident for numerous BWRs and that within the confmes of the dermition, the EDG design function was met during the Core Spray LSFT.

The licensee informed the NRC of their intention to modify the core spray /EDG interface logic by installing an additional relay in parallel with the existing relay. The new relay will interface with companion logic channel and ensure the LOCA signal is not blocked to any EDG during an LSFT of either CS loop. The staff recognized this as a prudent initiative.

I

.

\\

i

'

11.0 MANAGEMENT MEETINGS (30702,71707)

The Resident Inspectors provided a verbal summary of preliminary findings to the station management at the conclusion of the inspection. During the inspection, the Resident Inspectors verbally notified licensee management concerning preliminary findings. The inspectors did not provide any written inspection material to the licensee during the inspection. The licensee did not express any disagreement with the inspection findings. This report does not contain propri-etary information. The following specialist inspection also occurred during the report period:

D2 Subject Report No.

Inspector 5/12-19/93 Environmental Monitoring 93-12 Strickmeyer

,.

l-

--

-

--