IR 05000250/1988026
| ML17345A488 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/24/1988 |
| From: | Butcher R, Mcelhinney T, Schnebli G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17345A487 | List: |
| References | |
| 50-250-88-26, 50-251-88-26, NUDOCS 8811030334 | |
| Download: ML17345A488 (19) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTAST., N.W..
ATLANTA,GEORGIA 30323 Report Nos.:
50-250/88-26 and 50-251/88-26 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33102 Docket Nos.:
50-250 and 50-251 Facility Name:
Turkey Point 3 and
License Nos.:
DPR-31 and DPR-41 Inspection Conducted:
Augu t 26, 1988 through September 25, 198S Inspectors:
R.
C. Butcher, Se ior sident Inspector Dat Signed T.
F. McElhinney, Reside Inspector lo gP't Signed G. A.
h bl
, Reside I spector ate Signed R.
.
C lenjak, e tion Chief Division of Reactor Project Dat Signed SUMMARY Scope:
This rou ine, unannounced inspection entailed direct inspection at the site, including backshift inspection, in the areas of annual and monthly surveillance, maintenance observations and reviews, engineered safety features, operational safety, facility modifications and plant events.
Results:
No violations or deviations were identified.
Two unresolved items and two inspector followup items ( IFIs)
were identified.
Additionally, the'esidents expressed concern about a
problem with the power operated relief valve (PORY) opening times and the related safety evaluation.
These concerns were previously brought to the licensee's attention in NRC Inspection Report 88-14, on June 25, 198S.
The licensee's corrective actions were delayed and NRC discretionary enforcement was necessary to allow Unit 4 shutdown for the refueling outage on September 19, 1988.
Another resident inspector concern was the excessive amount of time required to implement corrective actions to preven.
a possible spill from the spent fuel pool pump vent valves.
These valves have been shown x.o inadvertently open due to vibration.
8811030334 881025 PDR
@DOCK 05000250
PNU
The licensee committed to develop a program to periodically verify the proper oil level in all automatic oilers.
The program will ensure that controls are established to maintain proper oil levels, and that personnel are trained on the automatic oilers'
REPORT DETAILS Persons Contacted Licensee Employees B. A. Abrishami, System Performance Supervisor J.
W. Anderson, Quality Assurance (QA)
Superintendent'L.
W. Bladow, QA Superintendent J. Arias, Regulation and Compliance Supervisor D.
A Chancy, Engineering Manager
"J
~
E. Cross, Plant Manager-Nuclear J.
Donis, Engineering Department Supervisor
"R. J. Earl, Quality Control (QC) Supervisor T. A. Finn, Training Supervisor S. Hale, Engineering Project Supervisor
- R. D. Hart, Regulation and Compliance Engineer P. Higgins, Site Engineering-Licensing
~J.
W. Kappes, Maintenance Superintendent V. A. Kaminskas, Reactor Engineering Supervisor
- J. A. Labarraque, Senior Technical Advisor R.
G.
Mende, Operations Supervisor J.
S. Odom,'ite Vice President
"L. W. Pearce, Operations Superintendent
- G.
M. Smith, Services Manager-Nuclear
- F. H. Southworth, Technical Department Supervisor J.
C. Strong, Mechanical Department Supervisor D. Tomaszewski, Instrument and Control ( I&C) Department J.
D. Webb, Operations - Maintenance Coordinator W.
R. Williams, Assistant Superintendent Planned Mainte
"A. Zielonka, Engineering Supervisor Supervisor nance Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, mechanics, and electricians.
- Attended exit interview on September 26, 1988.
Unresolved Items (URI)
Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations of requirements or deviations from commitments.
Two unresolved-items are identified in this report.
See paragraphs 4 and 10.
Followup on Items of Noncompliance (92702)
A review was conducted of the following noncompliances to assure that corrective actions were implemented and results were in conformance with regulatory requirements.
Verification of corrective action was achieved through record reviews, observation, and discussions with licensee
personnel.
Licensee correspondence was evaluated to ensure that the responses were timely, and that corrective actions were implemented within the time periods specified in the reply.
(Open)
Violation 50-250,251/88-25-01.
This violation concerns an inadequate maintenance procedure for the 4A Spent Fuel Pool (SFP) cooling pump.
The root cause of this event was the improper installation of the oiler on the 4A SFP cooling pump.
See NRC Inspection Report ( IR)
50-250,251/88-25 for a discussion on the oiler.
The inspectors observed the
"as installed" condition of the oilers on safety related pumps througnout the plant.
Although the oilers appeared to be installed at the proper height to ensure an adequate oil level to the bearings, several deficiencies were noted and discussed with appropriate plant management.
The deficiencies noted included:
Loose oilers; oilers set slightly high or low; no oil level mark on some bearing housings to indicate proper oil level; 'different types of oilers installed, which caused confusion on the method of adding oil to the oiler reservoir; lack of understanding with maintenance and operations personnel in the principle of operation of the oiler; and setting the oiler to the proper height.
As a result of the discussions, the licensee performed a detailed inspection of all safety related oilers and corrected their noted deficiencies.
The licensee informed personnel, via their periodic employee news letter, as to the principle of operation of different type oilers installed at the plant.
Also additional training will be developed for maintenance and operations personnel in this area.
The licensee stated that a
program will be developed to ensure that the proper oil level for each bearing housing is periodically verified along with a method to control the proper setting of the oilers.
Another concern. identified during this event was the opening of the 4A SFP cooling pump vent valve, which resulted in the spill of contaminated fuel pool water.
The independent laboratory contracted to determine the cause of the vent valve opening concluded that vibration caused by pump failure could have opened the valve.
The inspectors witnessed several tests that verified this conclusion.
One area identified on this issue is that the pump vent discharge pipe was threaded and could be capped, thus preventing the spill of radioactive water if the vent valve inadvertently opened.
The Event Response Team (ERT),
formed to evaluate the spill, recognized this i-ssue shortly after the event on August 16, 1988, however, nc actions were taken until September 22, 1988, to prevent recurrence.
The licensee opted to lockwire the vent valves shut versus installing caps on the vents.
This option was chosen to allow rapid venting of the pump, if required, in that the operator could just remove the lockwire to the vent in lieu of removing the cap.
In addition, the licensee stated that installation of a
vent cap was hazardous from a personnel safety view point because if the valve leaked by the seat, the capped piping could become pressurized and blow the cap off during removal.
The licensee also stated that other pumps with the same type of vent valve would be lockwired shut if required.
The inspectors will continue to follow these issues prior to close out of the violatio (Closed)
Violation 50-251/87-06-02.
This violation concerns the inadequate analysis performed on the failed discharge check valve for intake cooling water ( ICW)
pump 4A.
The licensee addressed the concern in a letter to the NRC dated May 15, 1987.
A safety evaluation for the ICW system check valves was performed on May 5, 1987, File No.
PTP 287-18, which concluded that no unreviewed safety question existed.
The valves were subsequently replaced with a
valve of a different design less susceptib',e to the failures noted with the older type design.
Violation 50-251/87-06-02 is closed.
Followup on Unresolved Items (URIs),
Inspector Followup Items ( IFIs),
Inspection and Enforcement Information Notices ( IENs), IE Bulletins (IEBs)
(information only),
IE Circulars ( IECs),
and NRC Requests (92701);
(Closed)
URI 50-250,251/86-10-06, concerns the seismic adequacy of the metal supports adjacent to each'CW pump.
Nonconformance Report (NCR)
307-86 and NCR 524-86, dated March 9, and March 21, 1986, respectively, addressed significant corrosion concerns for the support frames and conduit for 3A, 3B, 3C and 4C ICW pumps.
Although 'the NCRs were unrelated to the referenced inspection concern, supports for the heater boxes were replaced with frames capable of supporting the attached electrical equipment for all Final Safety Analysis Report (FSAR) design load cases.
An engineering calculation dated October 10, 1986, demonstrated that the support frames for 4A and 4B ICW pump motor heater indicator junction box and attached electrical components were adequate to sustain FSAR postulated seismic acceleration load cases.
UNR 50-250,251/86-10-06 is closed.
(Closed)
Inspector Followup Item 50-250,251/88-14-03 identified a
discrepancy between the actual opening times for the PORVs and the opening
.
times assumed in the Safety Evaluation Report issued for License Amend-ments
and 47.
By letter dated March 14, 1980, the NRC issued Amend-ments 55 and 47 to the Turkey Point Operating Licenses for Units 3 and 4, respectively.
These amendments incorporated the limiting conditions for operation and surveillance requirements associated with the low tempera-ture reactor vessel Overpressure Mitigation System (OMS) that is installed for both units.
The design basis events analyzed for the OMS include an inadvertent injection of a safety injection pump for the mass input case,
.and a start of a reactor coolant pump with a
50 degrees F temperature 5ifference between the water in the reactor vessel and the water in the steam generator for the heat input case.
The Westinghouse analysis to determine the Power Operated Relief Valve (PORV)
OMS assumed a relief valve opening time of 2.0 seconds for the mass input case and 3.0 seconds opening time for the heat input case.
A review of the Inservice Test ( IST)
program indicated that the valves were not being tested to the design basis requirements, but to a
15 second opening time.
The actual stroke times from 1984 to present were reviewed.
This review revealed that the valves were not meeting the design basis opening time requirements in most tests performed.
The times achieved ranged from below 2.0 seconds to a maximum of 6.41 second The consequence of the degraded stroke times was to place the plant in an unanalyzed condition while in water solid plant operations.
The licensee requested Westinghouse to perform an evaluation to determine the magnitude of the setpoint overshoot with the degraded stroke time for the most probable overpressure event, and to recommend compensatory measures to be taken until the PORYs can be worked.
This analysis determined'hat the most probable overpressure event was a letdown isolation with the positive displacement charging pumps running.
The case where two charging pumps are running with a loss of letdown and a
PORV opening stroke time of 6.0 seconds revealed that a setpoint overshoot of 18 psi would occur.
This is well below the
CFR 50, Appendix G limits.
The basis for Technical Specification (TS) 3. 15, Overpressure Mitiqating System, states that either of the PORVs has an adequate relieving capability to,protect the Reactor Cooling System (RCS)
from overpres-surization when the transient is limited to either:
( 1) the start of an idle Reactor Coolant Pump (RCP) with secondary water temperature of the steam generator a
50 degrees F above the RCS cold leg temperature or (2) the start of a
High Pressure Safety Injection (HPSI)
pump and its injection into a water solid RCS.
These transients were not evaluated by Westinghouse in their analysis of the degraded stroke times.
The inspectors questioned the operability of the OMS based on the uncertainty of the OMS to prevent a
design basis transient from exceeding the
CFR 50, Appendix G limits.
The inspectors also questioned the licensee concerning potential reportability since this condition had existed for at least four years, therefore, TS could have been violated.
On September 13, 1988, the licensee reported a significant event to the NRC per
CFR 50.72(b)( l)(i) (8).
This event is not an immediate opepabi lity concern.
TS 3. 15, OMS, requires two PORVs to be operable with a setpoint of 415 psig
+ 15 psi at RCS temperatures less than or equal to 275 degrees F and RCS pressure boundary established.
Therefore, this situation would become a problem when the licensee decided to shut the unit down and cool down to less than 275 degrees F.
The licensee determined that, since a Unit 4 refueling outage was to commence shortly, relief from TS 3. 15 was required.
On September 16, 1988, per FPL letter ( L-88-420),
the licensee requested discretionary enforcement to enable the plant to cool down to less than 275 degrees F.
This request is only through the next refueling outage for each unit, which allows the licensee time to resolve this issue (i.e.
repairing/modifying PORVs to meet the 2.0 seconds stroke requirement).
In order to prevent a design basis overpressure transient, the licensee committed to the following administrative restrictions prior to cooldown below 275 degrees F:
a.
HPSI hot leg and cold leg injection valves will be closed and their breakers racked out.
The specific valves isolated will provide double isolation to ensure that a single failure will not allow the HPSI to inject into the solid RC b.
Anytime a reactor coolant pump is secured below 275 degrees F, its breaker is to be racked out to prevent an inadvertent start.
The licensee's.
evaluation of the effect of PORV opening times on the operability of the OMS during previous shutdowns and cooldowns below 275 degrees F,
is still ongoing and will be tracked as unresolved item 50-250,251/88-26-01.
(Closed)
TI 2515/98, Information of High Temperature Inside Containment/
Drywell in PWR and BWR Plants.
The inspectors collected the required information under this Temporary Instruction (TI)
a'nd forwarded it to the designated contact in Region II.
This item is closed.
Onsite Followup and In-Office Review of Nonroutine Events (92700/92712)
The Licensee Event Reports'LERs)
discussed below were reviewed and closed.
The inspectors verified that reporting requirements had been met, root cause analysis was performed, corrective actions appeared appropriate, and generic applicability had been considered.
Additionally, the inspectors verified that the licensee had reviewed each event, corrective actions were implemented, responsibility for corrective actions not fully completed was clearly assigned, safety questions had been evaluated and resolved, and violations of regulations or TS conditions had been identified.
(Closed)
Concerning the loss of flowpath from the Boric Acid Storage Tanks (BAST) to the Reactor Coolant.System due to coupling failure of the 3B Boric Acid Transfer Pump (BATP) and seal failure of the 3A BATP.
At the time of the event the 4A BATP was realigned to provide the required flowpath to Unit 3.
The coupling was repaired on the 3B BATP and the pump was tested and returned to service at 5: 15 pm, on April 15, 1988.
The mechanical seal ior the 3A BATP was replaced, and the pump was tested and returned to service at 4:00 pm on April 18, 1988.
This LER is closed.
(Closed)
Concerning a misinterpretation of TS 3.6.6.3 and 3.6.c.3 which allowed a
BAST to be classified in service when it was below the TS limit. Upon discovery the licensee placed the unit in TS 3.0. 1.
Adequate tank level and proper pump lineup were immediately verified for the unit.
A letter dated September 30, 1987, was issued by the Regulation and Compliance Section to clarify the interpretation of the TS require-ments for the BASTs.
The procedure for the RCO Daily Logs, 3/4-OSP-201. 1, was modified to verify the requirements of TS.
This LER is closed.
Monthly Surveillance Observation (61726)
The inspectors observed TS required surveillance testing and verified:
That the test procedure conformed to the requirements of. the TS; that testing was performed in accordance with adeauate procedures; that test instrumentation was calibrated;
'that limiting conditions for operation (LCO) were met; that test results met acceptance criteria requirements
and were reviewed by personnel other than the individual directing the test; that deficiencies were identified, as appropriate and were properly reviewed and resolved by management personnel; and that system restoration was adequate.
For completed tests, the inspectors verified that testing frequencies were met and tests were performed by qualified individuals.
The inspectors witnessed/reviewed portions of the following tests:
O-OSP-016.26 Electric Driven Fire Pump Operability Test 3-OP-. 1604. 1 Full Length Rod Cluster Control (RCC) Periodic Exercise 4-PMI-028.2 Axial Flux, Rod Deviation and Rod Position Indication Monthly Test No violations or deviations were identified in the areas inspected.
7.
Maintenance Observations (62703)
Station maintenance activities of safety related systems and components were observed and reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with TS.
The following items were considered during this review, as appropriate:
That Limiting Conditions for Operations (LCOs) were met while components or systems were removed from service; that approvals were obtained prior to initiating work; that activities were accomplished using approved procedures and were inspected as applicable; that procedures used were adequate to control the activity; that troubleshooting activities were controlled and repair records accurately reflected the maintenance performed; that functional testing and/or calibrations were performed prior to returning components or systems to service; that Quality Control (QC)
records were maintained; that activities were accomplished by qualified personnel; that parts and materials used were properly certified; that radiological controls were properly implemented; that QC hold points were established and observed where required; that fire prevention controls were implemented; that outside contractor force activities were controlled in accordance with the approved QA program; and that housekeeping was actively pursued.
The inspectors witnessed/reviewed portions of the following maintenance activities in progress:
Repair of 3B Main Circulating Water Pump and Motor.
Chemical Cleaning of the ICW/CCW Heat Exchangers.
Troubleshooting the Failure of 4A SFP Cooling Pump Vent Valve.
Troubleshooting Control Rod K-12 Dro Acid Flush of the Emergency Diesel Generator Cooling Systems.
Repair of "B" Emergency Diesel Generator Radiator Leaks.
Replacement of "B" Emergency Diesel Generator Radiator.
No violations or deviations were identified in the areas inspected.
8.
Engineered Safety Features Walkdown (71710)
'The inspectors performed an inspection designed to verify the operability of the Unit 4 Safety Injection System outside tne containment.
This was accompli shed by performing a
complete walkdown of all accessible equipment.
The following criteria were used, as appropriate, during this inspection:
a.
Systems lineup procedures match plant drawings and as built configura-tions~
b.
Housekeeping was adequate and appropriate levels of cleanliness were being maintained.
c.
Valves in the system are correctly installed and did not exhibit signs of gross packing leakage, bent stems, missing handwheels or improper labeling.
d.
Hangers and supports were made up properly and aligned correctly.
e.
Valves in the flow paths were in correct position as required by the applicable procedures with power available, and valves were locked/
lockwired as required.
f.
Local and remote position indications were compared and remote instrumentation was functional.
g.
Major system components were properly labeled.
The inspectors reviewed the following documents during the course of the inspection:
4-0P-062, revision dated March 17, 1988, Safety Injection System; 4-0SP-202.1, revision dated July 14, 1988, Safety Irjection/
Residual Heat Removal Flowpath Verification; and Operating Diagram 5610-T-E-4510, sheet one, revision 80, Safety Injection and Residual Heat Removal Systems Outside Containment.
The following deficiencies were noted and discussed with appropriate plant management:
The following valves are not included in the valve alignment procedure 4-OP-062:
4A and 4B SI Pump Vents 948 E,F,G and H.
4A and 4B SI Pump Seal Water Vents, 3477 through 348 RWST LT-4-6583A Root Valve 4-897A Root Valves 4-899P and N for FI-4-942.
Valve 4-940W, Spent Fuel Pit Cooling System to SI Line Vent, labeled 4-940M in plant.
Valve 4-941T, Hose connection drain valve not labeled.
Valve 4-898T, RWST LT-4-6583B Isolation mislabeled as 4-897B.
Different component description for MOV-4-843A and B in 4-OP-062 even though both valves perform the same function.
In 4-0P-062, MOV-878B, SI Pump Header Tie, listed as component number 40708 located at MCC 4C, should be component number 30708 located at MCC 3C.
In 4-OSP-202.
1 the required position for MOV-4-843B, HPSI Cold Leg Injection, states:
"LOCKED OFF (N/A if RCS Temp is greater than 380 degrees F)".
Should also include a
statement for required position if less than 380 degrees F for clarity.
Correction of the noted discrepancies will be tracked as Inspector Follow-up Item 50-250,251/88-26-02.
9.
Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs, conducted discussions with control room operators, observed shift turn-overs and confirmed operability of instrumentation.
The inspectors verified the operability of selected emergency systems, verified that maintenance work orders had been submitted as required and that followup and prioritization of work were accomplished.
The inspectors reviewed tagout records, verified compliance with TS LCOs and verified the return to service of affected components.
By observation and direct interviews, verification was made that the physical security plan was being implemen ed.
Plant housekeeping/cleanliness conditions and implementation of radio-logical controls were observed.
Tours of the intake structure and diesel, auxiliary, control and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks and excessive vibrations.
The inspectors walked down accessible portions of the following safety related systems to verify opera'bility and proper valve/swi'ch alignment:
A and B Emergency Diesel Generators Control Room Vertical Panels and Safeguards Racks Intake Cooling Mater Structure 4160 Volt Buses and 480 Volt Load and Motor Control Centers Unit 3 and 4 Feedwater Platforms Unit 3 and 4 Condensate Storage Tank Area Auxiliary Feedwater Area Unit 3 and 4 Main Steam Platforms While reviewing Operating Procedure (OP)
3-0P-094, Containment Post Accident Monitoring Systems, dated August 25, 1988, the following minor discrepancies were noted:
Containment high radiation monitors RI-3-6311A and B were called out as RAD-3-6311A and BE Containment pressure monitors PI-3-6306A and B were called out as PR-3-6303A and B.
Containment pressure monitors PI-3-6425A and B were called out as PR-3-6425A and B.
Containment water leve'1 monitors LI-3-6308A and B and LI-3-6309A and B were called out as LT-3-6308A and B and LT-3-6309A and B.
'he equivalent procedure for Unit 4, 4-0P-094, had similar nomenclature errors.
Correction of the noted errors will be tracked -as Inspector Followup Item 50-250,251/88-26-03.
No violations or deviations were identified in the areas inspected.
10.
Physical Security (71881)
Station security activities were observed during this inspection period to ascertain that they were conducted in compliance with the approved Physical Security Plan (PSP).
The following attributes were considered during these observations, as appropriate:
That the minimum number of armed guards is on site for each shift; that search equipment such as x-ray machines, metal detectors and explosives detectors are operational; that the protected area (PA) barrier is well maintained and is not compromised by erosion, opening in the fence or walls, or proximity of vehicles or other objects that could be used to scale the barrier; that illumination in the PA is adequate to allow patrolling guards to observe the area at night and permit the use of closed circuit monitors by alarm station operators; that the vital area (VA) barriers are well maintained:
that persons granted access to the site are badged to indicate whether they have unescorted or escorted access
authorization; that there are no obstructions in the isolation zone that
- could conceal an individual attempting an unauthorized entry or inter-ference with the, detection/assessment system; and that when scarc equipment or alarm systems are inoperable, or when there is a breach of the PA or VA barrier, the licensee implements appropriate compensatory measures.
At 3:30 pm, on September 22, 1988, while touring the Emergency Diesel'enerator (EDG) area, the inspectors found that the security barrier at the B
EDG radiator had been removed, allowing access to the A
EDG vital area with no security guard in place.
At 11:00 a.m.,
on September 20, 1988, a guard was posted a
the front of the B
EDG barrier at the radiator to control access during replacement of the B
EDG radiator.
Approximately 1 1/2 hours later, the, guard was moved o another EDG area to cover a
related access door, but this left the open radiator area uncovered.
The B
EDG had not been declared a devitalized area and the A EDG was still in service.
This left uncovered several access paths into the A EDG radiator room.
After the inspectors notified the guard that a vital area was not secure, another guard was immediately posted until the licensee could review the event.
The licensee determined that a vital area had been left uncovered and made notification to the NRC.
Also, a telecon was held on September 23, with Region II security personnel.
The regional. security personnel will follow up on this event and this will be identified as unresolved item 50-250,251/88-26-04.
No violations or deviations were identified within the areas inspecte d.
Plant Events (93702)
.The following plant events were reviewed to determine facility status an d the need for further fol 1owup action.
Plant parameters were evaluated during transient response.
The significance of the event was evaluated along with the performance of the appropriate safety systems and the actions taken by the licensee.
The inspectors verified that required notifications were made to the NRC.
Evaluations were performed relative to the need for additional NRC response to the event.
Additionally, the following issues were examined, as appropriate:
details regarding the cause of the event; event chronology; safety system performance; licensee compliance with approved procedures; radiological consequences, if any; and proposed corrective actions.
The licensee plans to issue Licensee Event Report (LERs)
on each event within 30 days following the date of nce.
occurre
J
29 1988 the licensee's Quality Assurance (QA) group identified a potential procedural problem with Emergency Operating Procedure (E
)
OP)
3-EOP-E-O, Reactor.
Trip or Safety injection, dated June 8, 1988.
Attach-ment A to the procedure
, which lists the hydrogen analyzer start criteria d d t posi ion open manual isolation valve PAHM 4-008B shown on drawing 5610-T-E-4534, revision 15, Containment Ventilation 5-
. ost Acc'd i
no posii'ontainment Ventilation and Sampling System Outside Containment.
Further investigation by the licensee revealed that, although PAHN 4-008B existed
in the plant and on the drawing, no official design change had added the valve.
Also, an equivalent valve on Unit 3, PAHM 3-008B and vent valve PAHM 3-009, existed in the plant and on the drawing but no design change could be found that added these valves.
The licensee determined that the noted valves were physically located such that they could be in an inaccessible post accident radiation area.
The licensee then issued Plant Change/Modification (PC/M)88-310 and PC/M 88-311 to remove PAHM 3-008B and PAHM 3-009 from Unit 3 and PAHM 4-008B from Unit 4, respectively.
The valves noted above were installed only in the B train of the post accident hydrogen monitors on Units 3 and 4.
The A train post accident hydrogen monitors were not affected.
TS table 3.5-5, Accident Monitoring Instrumentation, item 12 for the containment hydrogen monitors, action item 6, states that with one hydrogen monitor inoperable, restore the inoperable monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> s.
The licensee modified the B trains of the post accident hydrogen monitors and returned them to service within the allowed 30 days.
On August 31, 1988, at ll:10 p.m.,
the control room ventilation system automatically tripped to the recirculation mode.
This control room isolation was reported to'he NRC in accordance with
CFR 50.72(b)(2)(ii).
The operators were performing a check of control room ventilation inlet air radiation monitor RAI-6642, when the isolation occurred.
This check is performed each shift.
A Plant Work Order (PWO)
was initiated to troubleshoot the control switch which should have interrupted the system initiation signal during the check.
The licensee recognized, during pre-operational testing, the possibility of inadvertent trips of the control room ventilation system due to the sensitivity of the control switch.
On June 6,
1988, a caution tag was placed on the control switch with the following caution:
When testing, after the alarm actuates, depress and hold the trip pushbutton 1, while slowly returning the test/operate switch io operate, in accordance with 4-OSP-201. 1.
In this event the operator did not have the opportunity to follow the caution tag.
The licensee is considering a
design change to prevent future inadvertant actuations.
12.
Exit Interview The inspection scope and findings were summarized during management interviews held throughout the reporting period with the Plant Manager-Huclear and selected members of his staff.
An exit meeting was conducted on September 26, 1988.
The areas requiring management attention were reviewed.
Yo proprietary information was provided to the inspectors during the reporting period.
The inspectors had the following findings:
50-250,251/88-26-01, Unresolved Item:
Review licensee's evaluation of the effect of PORV opening time on operability of the OMS during previous shutdowns.
(paragraph 4)
50-250,251/88-26-02, Inspector Followup Item:
Correction of noted discrepancies during walkdown of the Safety Injection System.
(paragraph 8)
50-250,251/88-26-03, Inspector Followup Item:
Correction of nomenclature errors in Operating Procedure 3/4-OP-094.
(paragraph 9)
50-250,251/88-26-04, Unresolved Item:
Followup on personnel access control to the Emergency Diesel Generator area.
(paragraph 10)