IR 05000271/1989017
| ML19324C112 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/03/1989 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19324C111 | List: |
| References | |
| 50-271-89-17, NUDOCS 8911150004 | |
| Download: ML19324C112 (17) | |
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U.S. NUCLEAR REGULATORY COMMIS$10N
REGION I
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Report No.
50-271/89-17
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Docket No.
50-271 License No. OPR-28 i
Licensee:
Vermont Yankee Nuclear Power Corporation
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RD 5. Box 169 Brattlebero, Vermont 05301 Facility:
Veement Yankee Nuclear Power Station f
Inspection At: Vernon, Vermont Inspection Conducted:
September 6 - October 16, 1989 Inspectors:
Geoffrey E. Grant, Senior Resident Inspector John B. Macdonald, Resident Inspector Approved by:
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//-19 7 A. Randy Bis 6gh, Chief, Reactor Projects Section 3A Date
Inspection Summary: Inspection on September 6 - October 16,1989 (Report No.
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h0-271/89-17)
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Areas Inspected:
Routine inspection on daytime and backshifts by two resident
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inspectors of: actions on previous inspection findings; operational safety; security; plant operations; maintenance and surveillance; engineering support;
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radiological controls; licensee event reports; licensee response to NRC initi-l atives; and, periodic reports.
i Results:
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General Conclusions on Adequacy, Strength or Weakness in Licensee Programs
t The licensee demonstrated excellent planning and interdepartmental com-
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munications and coordination in the preparation and execution of the ex-
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tensive activities conducted during the September 23 power reduction and
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drywell inspection entry (Section 6.1).
The September 11 discovery of an intentionally defeated reactor water cleanup pump room door indicates plant management efforts to insure all plant personnel respect and comply with established radiation protection measures have not been fully successful.
This event, in addition to pre-viously identified disengaged doors to locked high radiation areas located in the turbine building, is evidence that further management attention is required in this area (Section 8.1),
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8911150004 891107 PDR ADOCK 05000271 Q
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Violations
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One licensee-identified violation was identified involving three examples l
of missed or improperly implemented Technical Specification surveillances.
J Prompt and effective corrective actions were taken. No Notice of Viola-t tionisissued(Section7.1).
3.
Unresolved Items An unresolved item was identified this inspection period concerning assur-ance that coitsistent comprehensive procedure reviews are conducted (Sec-tion 7.1).
An unresolved item was identified this inspection period concerning the effectiveness of management corrective actions regsrding control of radio-active area barriers (Section 8.1).
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TABLE OF CONTENTS
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PAGE 1.
P e r s o n s C o n t a c t ed....................................................
2.
S umma ry o f T 4 c i l i ty Ac t i v i t i e s......................................
3.
Status of Previous Inspection Findings (IP 92701,92702*)............
3.1 (Closed) Unresolved Item 86-25-02: Licensee Review and Evaluation of Generic RHR Minimum Flow Requirements...........
3.2 (Closed) Unresolved Item 87-12-02: Evaluation of CO2 Intrusion Into the Control Room Ventilation System Following Fire Suppression System Actuations.................................
3.3 (Closed) Unresolved Item 88-08-04: Review of Licensee Actions to Control the Use of Tempora ry Tubi ng...........................
3.4 (Closed) Violation 88-14-05: Failure to Perform Adequate Reviews of Design Changes to Several Fire Protection System Control Panels.........................................................
4.
Operational Safety (IP 71707,71710).................................
4.1 Plant Operations Review.........................................
4.2 Safety System Review............................................
4.3 Inoperable Equipment............................................
4.4 Review o f Tempora ry Modi f ica ti on s...............................
4.5 Review of Switching and Tagging Operations......................
4.6 Operational Safety Findings.....................................
5.
Security (IP 71707)..................................................
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5.1 Observations of Physical Security...............................
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Plant Operations (IP 71707,93702)...................................
6.1 Tempo ra ry Powe r Red uc ti on.......................................
7.
Maintenance / Surveillance (IP 61726,62703)...........................
l 7.1 Missed Technical Specification Required Surve111ances...........
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Radiol ogi cal Cont rol s ( I P 71707).....................................
8.1 Compromised Administrative Control of Radiation Area Boundaries.
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Table of Contents L
PAGE 9.
EngineeringSupport(71707,35502)...................................
Il 9.1 Cable Vault Roon Fire Watch.....................................
Il 10. Licensee Event Reporti ng ( LER) (IP 93702)............................
Il 10.1 LER 89-20,......................................................
Il 10.2 LER 89-21.......................................................
10.3 LER 89-22.......................................................
10.4 LER 89-23........................................................
10.5 LER 89-24.......................................................
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Review of Licensee Response to NkC Initiatives (IP 35502)............
11.1 Installation of a Hardened Wetwell Vent.........................
12. Review of Periodic and Special Reports (IP 71707)....................
13. Ma n a g eme n t Me e t i n g s ( I P 307 0 3 ).......................................
- The NRC Inspection Manual Inspection Procedure (IP) that was used as inspsc-tion guidance is listed for each applicable report section, t
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DETAILS 1.
Persons Contacted Interviews and discussions were conducted with members of the licensee staff and management during the report period to obtain information per-tinent to the areas inspected.
Inspection findings were discussed peri-odically with the ma.1agement and supervisory personnel listed below.
Mr. P. Donnelly, Maintenance Superintendent
- Mr. R. Grippardi, Quality Assurance Supervisor
- Mr. 5. Jefferson, Assistant to Plant Manager Mr. J. Herron, Operations Supervisor Mr. R. Lopriore, Maintenance Supervisor Mr. M. Mete 11. Engineering Support Supervisor
'Mr. R. Pagodin, Technical Services Superintendent Mr. J. Pelletier, Plant Manager
'Mr. R. Wancryk, Operations Superintendent Mr. T. Watson, I&C Supervisor
- Attendee at post-inspection exit meeting 2.
Summary of Facility Activities Vermont Yankee Nuclear Power Station (VYNPS or the plant) continued full power operations during this report period.
Throughout the period, short term scheduled power reductions to 80-95% of full power were conducted weekly to perform routine surveillances of control rod drives, main tur-bine and bypass valves. On September 21, power was reduced to 89% in re-sponse to off-site distribution system perturbations.
Full power opera-tions resumed September 22. On September 23, power was reduced to 60% to accomplish a rod pattern exchange, main steamline isolation valve (MSIV)
full closure testing, single control rod scram testing and to conduct a drywell entry inspection.
Return to full power was achieved on September 24.
On September 12, the licensee informed the NRC Operations Center via the Emergency Notification System (ENS) when the High Pressure Coolant Injec-tion system (HPCI) was removed from service to perform periodic and cor-rective maintenance. This notification was in accordance with 10 CFR 50.72. Notifications were also made on September 23 when the Reactor Core Isolation Cooling system (RCIC) was removed from service to facilitate corrective maintenance, and on September 28 for the spurious actuation of a civil defense siren.
3.
Status of Previous Inspection Findings
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3.1 (Closed) Unresolved Item 86-25-02: Licensee Review and Evaluation of Generic RHR Minimum Flow Requirements.
In November 1986, the vendor for the RHR and core spray pumps (Bingham-Willamette) at Vermont Yankee issued service information communications that recommended
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minimum flow for the pumps be established at 2700 gpm and 1500 gpm respectively. Continuous operation was defined as more than two
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hours of minimum flowpath operations in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. The
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VYNPS plant design limits RHR and core spray flow path to 350 gpm per pump.
However, routine minimum flowpath operations are limited to
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approximately one to two minutes during surveillance testing. The
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licensee performed extensive evaluation of this condition and in re-sponse to USNRC Bulletin 88-04, " Potential Safety Related Pump Loss."
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Ultimately, the licensee concluded a significant safety hazard does
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not exist at VYNPS relative to this issue.
This position was based on empirical data accumulated in conjunction with GE (General Elec-tric) and the pump vendor, as well as, extensive reliable operation.
In addition, metallurgical examination of the four original instal-l 1ation RHR impe11ers replaced during the 1987 refueling indicated no
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operationally induced degradation. However, because the potential exists for this scenario to present a safety concern elsewhere, the
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licensee submitted a report in accordance with 10 CFR 21 requirements
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on March 20, 1987. Any further action with respect to this item will
be addressed with regard to the generic issue.
The inspectors had no further questions at this time.
This item is closed.
3.2 (Closed) Unresolved Itein 87-12-02: Evaluation of CO2 Intrusion Into the Control Room Ventilation System Following Fire Suppression System Actuations. On several occasions since 1983, the CO2 fire suppres-
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sion systems in the switchgecr rooms have spuriously actuated.
Fol-lowing actuation, CO2 has exhausted from the affected switchgear room, migrated into the control room ventilation system inlet plenum, and caused the toxic gas monitoring systLm (TGM) to initiate.
Initi-ation of the TGM system requires the control room operators to don self-contained breathing apparatus (SCBA) until a safe control room atmosphere can be verified.
Following each event, the licensee con-ducted an adequate root cause evaluation of the CO2 system actuation.
However, corrective actions to preclude TGM system initiations we re not effective.
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On March 3, 1989, the TGM again initiated after the spurious actu-ation of the west switchgear room C02 fire suppression system.
Fol-lowing this event, the licensee performed an extensive analysis of the TGM system initiation and enacted actions which appear appropri-ate to preclude similar events. The licensee conducted a design re-view of the TGM system which verified the system is operating as de-signed.
The review also concluded that future TGM system initiations could occur following CO2 fire suppression system actuations.
There-fore, the licensee revised applicable operating procedures (ops 3020 and 2186) to require the control room ventilation system and the TGM system to be placed in the recirculation mode immediately following a CO2 system discharge.
In addition, the licensee inspected and, as necessary, repaired components necessary to prevent outside air in-leakage while the control room ventilation and TGM systems are in
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the recirculation mode. These actions in conjunction with generally improving fire protection program training appear adequate to pre-clude similar TGM initiations. This item is closed.
3.3 (Closed)UnresolvedItem 88-08-04: Review of Licensee Actions to Control the Use of Temporary Tubing. On~ April 14, 1988, the licensee identified that an uncontrolled length of temporary tubing had been installed on the standby gas treatment system which adversely im-pacted system operability. The licensee reported this event in LER 88-05. As stated, the licensee identified all temporary tubing applications, performed appropriate evaluations of their operational impact, and issued mechanical bypass orders to ensure proper admini-strative control of the tubing is maintained. Additionally, in re-sponse to inspector concerns, the licensee recently implemented an enhanced temporary modification program which includes comprehensive reviews and controls of all temporary plant modifications (see In-spection Reports 50-271/88-14 and 89-12 for more detail). The in-spector had no further questions. This itam is closed.
3.4 (Closed) Violation 88-14-05: Failure to Perform Adequate Reviews of Design Changes to Several Fire protection System Control Panels.
In September 1988, the licensee identified that the fire protection panels in the east and west switchgear rooms, the cable vault, and the control room contained a module whose actual current flow could exceed design capacity, Upon discovery of this condition, each panel was declared inoperable and firewatches were posted until temporary jumpers were installed to provide sufficient power sources and cur-rent protection for the affected circuits.
Inspector review of the event indicated that inadequate engineering evaluation of design changes resulted in increased electrical loads being placed on the modules until design capacities were exceeded.
In addition to the immediate actions described above, the licensee revised the appropriate fire protection control system drawings to include original system parameters and has proposed permanent panel modifications. The flawed design change packages and reviews were executed from the late 1970's through 1982.
The increased quality control and oversight of the preparation, review and implementation of the design changes program since that time further decrease the likelihood of similar events.
The inspectors have no further ques-tions. This item is closed.
4.
Operational Safety 4.1 Plant Operations Review The inspector observed plant operations during regular and backshif t tours of the following areas:
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Control Room Cable Vault i
Reactor Building fence Line (Protected Area)
Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building Control room instruments were observed for correlatian between chan-nels, proper functioning, and conformance with technical specifica-
tions. Alarm conditions in effect and alarms received in the control
room were reviewed and discussed with the operators. Operator aware-
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ness and response to these conditions were reviewed. Operators were
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found cognizant of board and plant conditions.
Control room and shift manning were compared with technical specification require-ments.
Posting and control of radiation, contaminated, and high radi-ation areas were inspected. Use of and compliance with radiation work permits and use of required petsonnel monitoring devices were
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checked.
Plant housekeeping controls were observed including control of flammable and other hazardous materials.
During plant tours, logs
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ar.d records were reviewed to ensure compliance with station proce-dures, to determine if entries were made correctly, and to verify correct communication of equipment status.
These records included various operating logs, turnover sheets, tagout and jumper logs, and r
potential reportable occurrence reports.
Inspections were performed on weekends and backshif ts including Septerr.ber 6, 7, 12, 13, 19, 20 26, 27,28, and October 4, and 5, 1989.
" Deep backshift" inspections
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were conducted as follows:
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Time Date
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9/23/89 10:00 a.m. - 3:00 p.m.
Operators and shift supervisors were alert, attentive and responded appropriately to annunciators and plant conditions.
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4.2 Safety System Review
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The emergency diesel generators, reactor core isolation cooling, core
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spray, residual heat removal, standby gas treatment, residual heat l
removal service water, safety related electrical, and high pressure
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coolant injection systems were reviewed to verify proper alignment and operational status in the standby mode.
The review included I
verification that: (1) accessible major flow path valves were posi-tioned correctly, (ii) power supplies were energized (ifi) lubri-cation and corrponent cooling was proper, and (iv) components were operable based on a visual inspection of equipment for leakage and
general conditions. No violations or safety concerns were identi-fied.
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o 4.3 inoperable Equipment
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Actions taken by plant personnel during periods when equipment was inoperable were reviewed to verify: technical specification limits were met; alternate surveillance testing was completed satisfac-torily; and, equipment returned to service upon completion of repairs was proper.
This review was completed for the following items:
Date Out Date In System 97f 9/7
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9/12 9/14 HPCI 9/23 9/24 RCIC 9/29 10/5
"A" Toxic Gas Monitor 10/3 10/5
"A" EDG 7/30 ROOS *
Drywell HVAC-RRU-1**
- Remained out-of-service at the conclusion of the inspection period.
- RRV-1 was found to have a broken shaft and will remain rut-of-ser-Vice until the next outage of sufficient duration to fr.cilitate repairs.
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4.4 Review of Temporary Modifications Temporary modifications were reviewed to verify that controls estab-lished by Ap 0020 were met, no conflicts with technical specifica-tions were created, safety evaluations, as required by 10 CFR 50.59, were prepared, and requests were reviewed and approved prior to in-sta11ation.
Implementation of the requests was reviewed on a samp-ling basis.
The following request was reviewed:
89-50 -- Implemented on September 13 to relocate the power supply for RBCCW-117, the reactor building closed cooling water system re-turn line outboard isolation valve, to a motor control center (MCC)
supplied by an emergency bus, to return the system to the configura-tion described by the FSAR. The power supply was relocated from non-safety related MCC-7A to safety related MCC-8B (see Section 10.3 in this report and IR 89-12 for more detail).
Additionally, several temporary modifications were closed out during the report period.
These were reviewed for completeness and adequacy of system restoration.
4.5 Review of Switching & Tagging Operations The switching and tagging log was reviewed and tagging activities were inspected to verify plant equipment was controlled in accordance with the requirements of AP 0140, Vermont Local Control Switching Rules, a
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l 4.6 Operational Safety Findings
Licensee administrative control of off-normal system configurations by the use of temporary modifications and twitching and tagging pro-cedures, as reviewed in Sections 4.5 and 4.6, was in compliance with procedural instructions and was consistent with plant safety.
Back-shift inspections have consistently found operators to be alert and attentive.
Operations are routinely conducted in a professional raanner in an atmosphere of quiet control and competence. With the exception of isolated instances, overall plant cleanliness and mate-rial condition continue to be good. No deficiencies were identified in licensee operations associated with the reviews covered in Section 4.
5.
Security 5.1 Observations of physical Security Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accord-ance with the security plan and approved procedures.
This review included the following security measures: guard staffing; vital and protected area barrier integrity; maintenance of isolation zones; and, implementation of access controls, including authorization, badging, escorting, and searches.
No inadequacies were identified.
6.
Plant Operations 6.1 Temporary Power Reduction On September 23, the licensee reduced power to approximately 60% in order to perform a rod pattern exchange and to conduct a drywell entry and inspection.
In addition to the rod pattern exchange, scram testing of 30 control rods, the MSIV full closure testing, and weekly bypass valve and turbine testing were completed successfully.
The drywell entry was conducted to inspect for suspected steam leaks, to evaluate the condition of drywell air handling unit RRU-1, and to investigate the cause of an "A" recirculation pun.p low oil level alarm. As previously documented in inspection report 50-271/89-12, RRU-1 had been secured following indications of a failed fan belt.
Visual inspection of RRU-1 revealed that the motor shaft had actually broken near the belt pulleys.
The RRU will remain out of service until a shutdown of sufficient duration to facilitate repairs.
In addition to the loss of RRU-1, the licenses had believed a packing steam leak from valves in the HPCI or RCIC steam supply lines may have been contributing to increases in local and average drywell tem-peratures.
However, no steam leaks were identified during the dry-well inspection.
The "A" recirculation pump oil level was within specifications and the low level alarm was determined to be spurlous.
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The licensee demonstrated excellent planning prior to the power re-duction. Activity objectives were clearly stated, proper training was conducted and appropriate radiation protection measures were in place. While at reduced power, good communications and cooperation between departments was evident. All activities were conducted in a well controlled manner. The inspectors had no concerns regarding this evolution.
7.
Maintenance / Surveillance 7.1 Missed Technical Specification Required Surve111ances On September 11, the licensee identified that the TS required daily surveillance of the Low Pressure Coolant Injection (LPCI) crosstie monitor was not being performed.
The LPCI crosstie monitor is simply the main control room position indicating lights for the LPCI cross-tie valve RHR-20.
Since the LPCI loop select logic was removed in 1974, RHR-20 he,s oeen required to be maintained locked closed with its motor power leads disconnected. Only the RHR 20 motor starter remained energized to provide power for control room position indi-cation. The TS Table 4.2.1 was amended in 1975 to require a daily check of the LPCI crosstie monitor to ensure RHR-20 remained closed.
However, on March 21, during tne 1989 refueling outage, LPCI crosstie monitor capability was lost when the RHR-20 circuit breaker was re-moved from its motor control center cubicle for use in another appli-cation.
On September 13, when senior operations department personnel were reviewing a request to use the vacated RHR-20 motor control center cubicle for another application, it was determined the TS required daily instrument check could not be accomplished.
Two factors contributed to this event.
Firstly, the intent of the TS requirement lacks amplifying discussion in the TS bases. Absent definitive bases documentation, the involved operators believed erroneously that, because the RHR-20 was chain locked closed with operational power removed and its keylock switch defeated, the intent of the TS requirement was met.
Additionally, in 1980 a revision to the procedure which specifically directed the completion of this surveillance droppea the check from the control room operator operational round sheet. The shutdown round sheet of the procedure AP 0150, " Responsibilities and Authori-ties of Operations Department Personnel," maintained the requirement to perform the daily check. This procedural inconsistency was not identified during any of the ensuing biennial reviews of the proce-dure.
A recent independent audit of the surveillance program identi-fied the TS Table 4.2.1 crosstie monitor instrument check among many other surveillances as having potential implementation problems.
The operations department response to the audit did not include surveil-lances with periodicities of a day or less, therefore, the LPCI
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t crosstie monitor was not addressed. Upon identification of this event the operations department reviewed the surveillances with daily n
or shorter periodicities and found no additional discrepancies.
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It should be noted that failure to perform the daily crosstie monitor instrument check was of negligible safety significance because the RHR-20 valve is essentially retired in place in the closed position.
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The licensee returned the monitor to service, revised AP 0150 to in-
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l clude the instrument check in the CR0 operational round sheet, and is t
currently evaluating further licensing action to address this TS.
Additionally, on September 13, the licinsee identified that IST pro-
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gram leak rate testing of the inboard RHR shutdown cooling suction
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line containment isolation valve, RHR-18, was not accomplished during the 1989 refueling outage as required.
Revision 9 of the IST pro-gram, implemented on January 17, 1989, included a commitment to leak
test RHR-18 during each refueling outage. However, this commitment l
was not incorporated into the implementing RHR surveillance proce-
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dure.
Consequently, the leak test was not performed during the 1989
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outage.
It was not until September 13, while performing procedure
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reviews in order to implement revision 10 to the IST program, that the licensee identified this discrepancy.
Although the valve was not leak rate tested, required valve stroke
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timing was conducted which did not indicate any valve degradation.
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System pressure downstream of RHR-18, which would also indicate valve degradation, has been normal throughout the current operating cycle.
Additionally, if RHR-18 was determined to be inoperable, indefinite i
continued plant operations is permitted by TS if the first downstream valve is closed. The downstream valve, RHR-17, is normally closed.
Temperature and pressure interlocks prevent opening of RHR-18 and RHR-17 at normal operating conditions. Additionally, a leakage de-
tection system which alarms in the control room would indicate gross leakage past these valves.
If gross leckage did in fact occur, over-pressure protection is provided by a safety valve.
It should be
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noted that RHR-18 is not identified by TS as a primary containment valve requiring leak rate testing.
j The licensee performed a complete review of IST program implementing
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procedures and no further discrepancies were identified.
The appro-priate procedure revision to incorporate RHR-18 leak rate testing will be in place prior to the next refueling outinge.
Further, future implementing procedure revisions involving the IST program will re-
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quire the review and concurrence of the IST program coordinator.
Failures to perform the LPCI crosstie monitor daily instrument check,
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to perform the IST program leak rate test of RHR-18, and to conduct
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battery surveillances in accordance with TS as documented in Section 10.1, are identified as three examples of violation of TS require-l ments. However, these events were identified by the licensee; were l
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L of minor safety significance; could not have been reasonably pre-cluded from occurring as a result of previous corrective actions; were properly reported; and appropriate c^rrective actions to prevent i
recurrence have or will be implemented. Therefore, in accordance with discretion afforded by 10 CFR 2, Appendix C, a Notice of Viola-i tion will not be issued for these licensee identified examples of TS
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noncompliance (50-271/89-17-01).
Licensee corrective actions were
comprehensive and adequately addressed these issues.
Consequently,
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this item is closed.
Causal analyses of these events identified procedural deficiencies as either primary or major contributors.
In each instance, several opportunities existed to identify the deficiencies during administra-
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tively required procedure reviews within the implementing department and at the Plant Operations Review Committee (PORC) quorum. The in-spectors expressed concern that procedure review, whether minor in-tent change, major revision or biennial review, fully address license bases in addition to technical adequacy. Although, the licensee re-view process is generally very effective, these events illustrate the potential for existing procedural inadequacies to go unnoticed through skveral independent revie..s.
Licensee action to ensure con-sistently comprehensive procedural reviews are performed is identi-fied as an unresolved item (50-271/89-17-02).
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R_adiological Controls 8.1 Compromised Administrative Control of Radiation Area Boundaries On September 11 at approximately 4:40 p.m., an auxiliary operator
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(AO) on routine reactor building rounds discovered the "B" reactor water clean-up (RCU) pump room door locking mechanism defeated by the
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application of tape to the locking bolt, thus preventing it from en-gaging.
The A0 also observed that the "A" RCU pump room door was ajar. He removed the tape and properly secured both doors prior to continuing his round. The A0 concluded his round at 5:00 p.m. and
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notified the RCA control point radiation protection technician (RPT)
of the condition in which he found the RCU pump room doors.
The RPT appropriately logged the event, independently verified the doors to
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be properly secured, and initiated a radiation protection incident report (RPIR) 89-71.
The RCU pump rooms are posted as high radiation areas (potential general area dose rates of 100 mR/ hour to 1000 mR/ hour to the whole body). As posted, the doors are not required by technical specifica-tions to be locked closed.
However, the licensee conservatively pro-vides administrative control of physical access to selected plant high radiation areas to further enhance personnel protection from inadvertent exposure, as well as, to reduce the possibility of vio-lating regulatory exposure limitations.
In addition to requiring a radiation work permit (RWP) and a survey meter to enter these areas, personnel must be issued a high radiation area key which is under the
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administrative control of the plant health physicist and the shift
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supervisor.. Procedure OP 0532, "High Radiation Area Key Control,"
provides explicit instruction for the control of keys issued to plant
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personnel by the radiation protection department.
Implicit in the instruction of OP 0532 is that the access doors to high radiation areas be similarly controlled at all times.
The licensee has aggressively investigated this event.
Based on in-terviews with involved personnel and review of ongoing plant activi-ties, the licensee has concluded the doors were improperly secured for approximately two hours.
The "B" RCU pump room door lock was most probably neutralized to facilitate ease of entry while plant staff prepared to support corrective maintenance on the "B" RCU pump.
No overexposures occurred as a result of this event.
The inspectors independently reviewed the radiation protection (RP) key control log, radiation survey maps, and reactor building egress lists; interviewed various licensee personnel; and verified the initial conclusions of the licensee investigation to be accurate.
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In September 1988 the licensee identified an apparent pervasive plant staff disregard for adn. nistrative and physical measures established to control access to posted high radiation areas (see Inspection Re-port 50-271/88-18). At that time, the licensee took immediate and comprehensive actions to identify and correct radiation protection programmatic deficiencies.
In addition, extensive staff retraining, including staff and departmental meetings with the plant manager, was conducted to ensure individual responsibilities and requirements for entry into posted radiation areas were clearly defined and under-stood.
The September 11, 1989 compromise of the "B" RCU pump room door lock has been the first reported event of this nature in the past year. However, in April 1989, on two occasions a turbine build-ing high radiation area door designated to be locked was found to be ajar thereby permitting potential uncontrolled entry into these areas.
Licensee review determined that the primary cause of these events was personnel failure to ensure door closure after passage.
Licensee corrective actions included surveillance of all doors and locking mechanisms, supervisor discussion of the events with depart-ment personnel, and training program revisions.
The April and Sop-tember 1989 events indicate that plant management has not been fully successful in assuring all personnel comply with approved procedures and policies established to enhance radiation protection measures.
Notwithstanding a conservative safety perspective toward the resolu-tion of previous events, additional management attention to address continuing compromise of radiological protection controls appears warranted. The inspectors will continue to review licensee admini-strative control of radiation areas.
Review of the effectiveness of management corrective actions is an unresolved issue (50-271/89-17-03).
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9.
Engineering Support I
9.1 Cable Vault Room Fire Watch On April 21, 1989, the licensee posted a continuous firewatch in the cable vault room following a determination that the supporting CO2 fire suppression system was inoperable (see Inspection Report 50-217/89-04 for more detail).
Routine inspector tours of the cable vault room in the ensuing mont.hs have found the watches to be alert, aware of the basis for the watch and cognizant of the process to sum-mon relief.
Inspector review of the watch scheduling practices in-dicated that for non-holiday weekdays an eight hour watch with an onsite dedicated relief was scheduled. On weekends and holidays the watch period was reduced to four hours, such that a dedicated relief was not necessary.
However, specific instruction has been provided to ensure a relief can be dispatched promptly as necessary during these periods of time.
The firewatch will remain in effect until a CO2. discharge test or its equivalent is performed which would demon-strate the ability of the suppression system to maintain a 50*4 CO2 concentration throughout the cable vault room for a ten minute mini-mum duration.
The inspec' ors had no further questions.
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10.
Licensee Event Reporting (LER)
The inspector reviewed the licensee event reports (LERs) listed below to determine that with respect to the general aspects of the events: (1) the report was submitted in a timely manner; (2) description of the events was accurate; (3) root cau:e analysis was performed; (4) safety impi(cations were considered; and (5) corrective actions implemented or planned were sufficient to preclude recurrence of a similar event.
10.1 LER 89-20
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The LER 89-20, " Removal of a TS Surveillance Requirement From a Pro-cedure Due to an Inadequate TS Review," addresses the August 11, 1989 licensee discovery that battery surveillance testing was not being conducted in accordance with TS. The TS 4.10.2.A requires that elec-trolyte temperature be obtained from the cells adjacent to the pilot cell. The TS requirement was previously fulfilled properly by pro-cedure OP 4210. " Maintenance and Surveillance of Lead Acid Storage Batteries." However, in March 1987, OP 4210 was revised and the re-quirement to monitor adjacent cell temperatures was erroneously re-placed by a requirement to monitor pilot cell temperature.
Upon dis-covery of this error, the licensee immediately issued a Department Instruction to correct OP 4210 and require the temperature of cells adjacent to the pilot cell be monitored. The licensee is currently addressing a proposed TS change to revise battery surveillances to be l
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consistent with applicable IEEE standards. The LER event descrip-tion, causal analysis, and corrective actions were well documented.
The LER fulfills the above criteria.
The inspectors had no further questions at this time.
10.2 LER 89-21 The LER 89-21, " Failure of RM-16-19-1B Primary Containment High Range Radiation Monitor," is a special report submittal to the Commission as required by TS Table 3.2,6 whenever one of the two high range con-tainment radiation monitors is inoperable in excess of 30 days. On July 25, 1989, the "B" containment high range radiation monitor failed downscale following erratic indications and was declared in-operable. Troubleshooting has identified the failure to be an open circuit in a detector signal cable inside primary containment.
Therefore, furt.her troubleshooting and corrective maintenance will not be possible until the next reactor shutdown.
The licensee is procuring the r.ecessary parts to facilitate the anticipated component replacements. This special L6R fulfills the requirements of TS Table 3.2.6, Note 6.
The inspectors had no further questions.
10.3 LER 89-22 The LER 89-22, "RBCCW Return Motor Operated Valve 70-117 Not Powered From Emergency Bus As Required by FSAR," addresses a design defi-ciency in the power supply to the motor operator of the RBCCW system return line containment isolation valve. The FSAR describes the valve, RBCCW-117, as being powered from one of the AC emergency busses. However, the valve was actually powered from a non-safety bus via motor control center 7A.
Power supplies were subsequently rerouted to MCC-8B which is powered by emergency bus 4.
This event was documented in detail in Section 8.4 of IR 89-12.
This LER pro-vides extensive causal analysis for the design deficiency and appro-priately identifies a similar occurrence reported in LER 89-09 (IR 89-12, Section 9.2).
No deficiencies were identified.
10.4 LER 89-23 The LER 89-23, " Failure to Perform Daily Instrument Checks of the LPCI Crosstie Monitor Due to Interpretation of TS Requirements,"
addresses the improper release of the LPCI crosstie valve circuit breaker for use in other application such that power was lost to the LpCI crosstie monitor. The details of this event are documented in l
Section 7.1 of this report. The event description developed an l
excellent history of the licensing bases for the LPCI crosstie moni-l tor as well as the evolution of procedures which implemented the in-
strument check. The causal analysis and corrective actions were well l
documented.
In addition, appropriate similar events were identified.
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The inspectors had no further question.,
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10.5 LER 89-24 The LER 89-24, " Missed RHR System Valve Leakage Surveillance Due To
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Incomplete Procedure Review," addresses the September 13, 1989, lic-ensee discovery that RNR-18 was not leak rate tested during the 1989 refueling outage, as required by the IST program. The details of the event are documented in Section 7.1 of this report.
This was a com-prehensive LER which provided good event analysis and appropriate corrective actions. No deficiencies were identified.
11.
Review of Licensee Response to NRC Initiatives 11.1 Installation of a Hardened Wetwell Vent
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In response to the recent Commission action regarding the Mark 1 Con-tainment Improvement Program, the licensee indicated in a September
1, 1989 letter to the NRC a voluntary commitment to install a hardened vent to provide containment overpressure protection.
The commitment was based upon the assumption that this protection would not reduce the overall margin to safety and would be of a passive det,ign which would not adversely affect the existing design basis.
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Installation would most likely be accomplished during the 1992 re-
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fueling outage.
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12.
R_eview of Periodic and Special Reports e
Upon receipt, the inspector reviewed periodic and special reports sub-i mitted pursuant to Technical Specifications.
This review verified, as
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applicable: (1) that the reported information was valid and included the NRC-required data; (2) that test results and supporting information were consistent with design predictions and performance specification; and (3) that planned corrective actions were adequate for resolution of the
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problem. The inspector also ascertained whether any reported information
should be classified as an abnormal occurrence.
The following reports
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were reviewed:
Monthly Statistical Report for plant operations for the month of
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September 1989.
I Feedwater leakage detection system monthly performance summary for
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the month of September 1989.
13.
Management Meetings At periodic intervals during this inspection, meetings were held with
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senior plant management to discuss the findings. A summary of findings
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for the report period was also discussed at the conclusion of the inspec-tion and prior to report issuance.
No proprietary information was identi-fied as being included in the report.
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