IR 05000309/1990025
| ML20029A226 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/28/1991 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20029A217 | List: |
| References | |
| 50-309-90-25, NUDOCS 9102050089 | |
| Download: ML20029A226 (15) | |
Text
,
.
U.S. NUCLEAR REGULATORY COhlhllSSION
REGION I
Inspection Report 50-309/90-25 License: DPR-36 Licensee:
hiaine Yankee Atomic Power Company 83 Edison Drive Augusta, hiaine 04336 Facility:
hiaine Yankee Atomic Power Plant inspection Period:
November 28,1990 through January 9,1991 Inspectors:
Charles S, hiarschall, Senior Resident Inspector Richard J. Freudenberger, Resident Inspector Approved by:
bO N.h iI2rrlH E. C. hicCabe, Chief, Reactor Project Section 311 Date Summary Resident inspection of plant operations, radiation protection, maintenance and surveillance, emergency preparedness, security, engineering and technical support, and safety assessment / quality verification.
For a summary of findings, see the following Overview.
91020500e9 910128 PDR ADOCK 05000309 G
_
.-.
.
.
OVERVIEW Maine Yankee Inspection Report No. 90-25 Elant Operatiap. The foresight demonstrated by the Manager of Operations and the skillful, controlled plant shutdown on December 17,1990 showed good planning and procedural controls, well-stated policies, appropriate decision making, compliance with procedures, and operator competence and proficiency.
Chemistry and Radiological Controls The safe shutdown of Maine Yankee on December 17, 1990 was enabled, to a large extent, by the alertness and safety consciousness of Chemistry personnel. The associated decision to extend the radiologically controlled area was conservative and timely.
Notwithstanding, radiological protection management involvement in steam generator work appeared to divert that management from appropriate review of implementation of radiological controls in the Turbine Hall. Also, the contamination control practices observed warrant attention.
Surveillance and Maintenancs. An unresolved item was opened pending further review of diesel load calculations and to address differences between the Technical Specification Emergency Diesel-Generator surveillance program and the program recommended by Draft Regulatory Guide 1.9.
In most cases, the testing conducted at Maine Yankee appears less rigorous than that recommended by the Regulatory Guide. An additional unresolved item was opened to monitor licensee and NRC review of repairs to Main Steam Valve MS-70.
Security. Security contractor transition from Hall Security to American Protection Services took place during December with no changes in security effectiveness noted. Maine Yankee took timely and appropriate actions in response to a Fitness For Duty matter.
Engineering and Technical Suppmt. Maine Yankee and Combustion Engineering personnel conducting eddy current testing and analysis were well-qualified, conducted a thorough inspection program, and developed improved analysis methods. Licensee efforts to identify and documem a generic safety concern resulted in a beneficial safety input to facilities that use GE AK-2A-25-1 Reactor Trip Breakers. The 10 CFR 50.59 review process at Maine Yankee was found to be adequate, with opportunities for improvoment in the administrative procedures. Plant Operations
[
Review Committee (PORC) reviews o 10 CFR 50.59 evaluations were rigorous.
Safety Assessment and Ouality Verifi:ation. An occasional tendency to question the accuracy of off-normal instrument indicators ndicates a lack of a questioning attitude and warrants i
immediate management attention. Tie lack of formal PORC review and approval of a Day i
Order which changed a procedure tequired by Technical Specification 5.8 violated NRC requireroents. Maine Yankee thoroughly inspected the steam generators; however, an apparent weakness in the understanding of the inspection basis and failure to consider design basis accidents indicated weakness in the resolution of this issue. A rewrite of the Corrective Action i
Request procedure has the potential to improve the effectiveness of corrective actions.
l kk
- - -
- -
-
____
___-
_-_
-___-
_ - _ -
-
.
TABLE OF CONTENTS OVERVIEW..............................................
ii i
TA B LE OF CONTENTS.......................................
iii 1.
P LA NT O P E R ATI O N S...................................
I 1.1 Unplanned Shutdown for Excessive RCS Leakage...........
,..
I 2.
R A DIOLOGIC A L CONTROLS
.............................
2.1 Leak Rate Monitoring.........
............
.........
2.2 Implementation of Radiological Control..................
..
2.3 Control of Contamination..........................
...
3.
M AINTENANCE/ SURVEILLANCE..............
............
3.1 Diesel-Generator Surveillances...........................
3.2 Repair of Main Steam Non-Return Bypass Valve.....
..........
3.3 Maintenance Observations
................
.........
..
3.4 Surveillance Observations......
...
.....
..
..
...
4.
PHYSICAL SECURITY..................
....
.
........
4.1 Change of Security Contractor............
.
..........
4.2 Fi tness For D u ty...............................
.
5.
ENGINEERING / TECHNICAL SUPPORT........................
5.1 Steam Generator Eddy Current Testing
.....................
5.2 Reactor Trip Breaker Failure...................,.....
..
5,3 Governor Replacement
...............................
5.4 Review of 10 CFR 50.59, Changes, Tests and Experiments....
...
6.
SAFETY ASSESSMENT / QUALITY VERIFICATION....
.
....
...
6.1 Control of Safety-Related Activities
.....
.
...
,
......
6.2 Correction Action Procedure...................
I1
.
.
..
6.3 Management involvement in Assuring Quality..............
I1
.
6.4 Resolution of Technical Issues from a Safety Standpoint I1
.
......
.
7.
ADMINISTRATIVE.......
.
..
.............
.......
.
7.1 Person Contacted.....
......,,.............
.
...
7.2 Summary of Facility Activities
............
..
.........
7.3 Interface with the State of Maine..........
.
..
..
.
.
7.4 Exit Meeting
..........
............
.
.
iii
_ _ _ _ _ _ _ _ _ __ _ - _ - _ _ _ - _ _ _ _ _ - _
_ _ - _ _ _ __
_
f IlETMLS 1.
PLANT OPEllATIONS On a daily basis, during routine facility tours the following were checked: manning, access control, adherence to procedures and Limiting Conditions for Operation, instrumentation, recorder traces, protective systems, control room annunciators, radiation monitors, emergency power source operability, operability of the Safety Parameter Display System (SPDS), control room logs, shift supervisor logs, and operating orders. On a weekly basis, selected Engineered Safety Features (ESI') trains were verified to be operable. The condition of plant equipment, radiological controls, security and safety were assessed. Biweekly, the inspector reviewed a safety-related tagout, chemistry sample results, shift turnovers, portions of the containment isolation valve lineup, the posting of notices to workers. Also biweekly, selected ESF trains were verified to be operable. Plant housekeeping and cleanliness were evaluated. The following items were considered noteworthy.
1.1 Unplanned Shutdown for Excessive itCS Leaknge On December 17,1990 at 3:40 p.m., operators began to reduce reactor power when the calculated primary to secondary leak rate (based on air ejector activity) in steam generator SG 1 increased from 0.004 gpm to 0.01 gpm. The sequence of events is presented below:
lilllcll2i'de Event / Action 12-1-90 SG-1 leak rate based on chemical analysis is 0.0006 gpm. Samples are analyzed daily.
0722/12-13 SG-1 leak rate increases to 0.001 gpm. Samples are analyzed twice per shift.
0100/12-14 1.cak rate continues to increase to 0.003 gpm.
0057/12-15 Leak rate reaches 0.005 gpm.
0425/12-16 Leak rate increases to 0.007 gpm and stabilizes for approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
0234/12-17 Leak rate jumps to 0.017 gpm. Air ejector RhtS (Radiation blonitor) indicates 72,550 cpm.
0245 PSS (Plant Shift Supervisor) notitu. SCO (Duty Cad Officer), requests PORC (Plant Operation Review Committee) meeting per Procedure AOP 2-10.
0340 At management direction, PSS commences power reduction at 5% per hour.
0450 Air ejector Rh1S jumps to 360,000 epm. PSS estimates leak rate has exceeded 0.035 gpm, orders emergency shutdown per AOP 2-10, Step 5.5.10, informs I)CO and hianager of Operations.
j
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ __ _ _ _ _ _ _ _ _______________________
__
..-
--.
---
...
- _- -
.
-..
-.
.
-
,
0521 SG-1 leak rate reaches 0.105 gpm based on sample.
0607 Air ejector RMS jumps to 600,000 cpm.
l 0636 Based on air ejector sample, leak rate reaules 1.4 gpm, confirming Technical Specification primary to secondary leak rate of 0.15 gpm has been exceeded 0653 Reactor power <2%.
0659 Commenced emergency boration to cold shutdown boron concentration.
0710 Reactor suberitical.
L During observation of the shutdown, the inspector noted highly professional performance by the operating crew, The ROs (reactor operators) were calm, skillfully accomplished the shutdown in accordance with procedures, and were effectively guided and monitored by the SOS (shift operating supervisor). -The PSS (plant shift supervisor) remained aware of plant status and
,
'
actions taken by the ROs and SOS, prevented distractions of the operating crew, and performed operations management functions.
On Friday, December 14, the Manager of Operations had cautioned operators to anticipate the
.
possibility of a rapid increase in tube leakage. As a result, the operators reviewed AOP 2-10,
'
Loss of Pressure Control /RCS Leak. AOP 2-10 imposes administrative restrictions on plant operation for increasing steam generator leakage based on NRC Bulletin 88-01, Tube Rupture as Experienced at North Anna. SG-1 leakage increased more rapidly than the North Anna event, and the operators were prepared to take the steps required by AOP 2-10.
The inspector concluded that the foresight demonstrated by the Manager of Operations and the skillful, controlled plant shutdow n were indicative of good planning and procedural controls, well stated policies, appropriate decision-making, compliance with procedures, and competent and proficient operators.
2.
RADIOLOGICAL CONTROLS Radiological controls were observed on a routine basis. Areas reviewed included Organization and Management, external radiation exposure-control and contamination control. Standard l
. industry radiological work practices, conformance to radiological control procedures and 10 CFR l
' Part 20 requirements were observed. The following items were noteworthy.
l
{
.-
-.
._
.
.
.
2.1 leak Rate Monitoring hiaine Yankee has been aware of primary to secondary leakage in SG-1 since startup in July 1990, at the end of the 1990 refueling outage, initially, leakage was barely detectable; when fuel leakage increased in mid-July, quantification of primary to secondary leakage became more precise.
During much of period between July and December 1990, the air ejector Rhis was inoperable (refer to Detail 6.3). As a result, operators relied on Chemistry personnel for accurate leak rate information. When leakage began to increase more rapidly, Chemistry personnel recognized the safety signincance and alerted operators and managers. As the leak rate increased, Chemistry personnel remained vigilant, increasing the frequency of sampling to provide operators more time to respond.
Chemistry personnel demonstrated a clear understanding of the safety significance of increased tube leakage in SG-1. The safe shutdown of hiaine Yankee on December 17,1990, was enabled, to a large extent, by the alertness and safety consciousness of Chemistry personnel.
2.2 Implementation of Rndlological Controls When the tube leak in SG-1 exceeded administrative limits for primary to secondary leakage on December 17, Radiological Protection personnel identified two areas in the Turbine lluilding with higher than normal radiation levels: the blowdown demineralizer and the condenser air ejectors.
To monitor and limit radiation exposure resulting from the potential carryover of radioactivity into systems in the Turbine Hall, Radiological Protection h!anagement elected to extend the radiologically controlled area boundary to include the Turbine Hall. At the time the ra:hological controls were implemented in the Turbine lluilding, most of the radiological protection resources were focused on steam generator inspection and repair.
On December 21, 1990, the senior resident inspector, accompanied by the State of Maine Nuclear Safety inspector, reviewed the radiological controls for the Turbine Hall. Weaknesses noted included postings which appeared contradictory, unclear boundaries in the vicinity of the Primary Component Cooling and Secondary Component Cooling pumps, and inadequate provision for frisking items carried out of the Turbine Hall. When the inadequacies were reported to plant management, immediate and effective corrective actions were implemented.
The decision to extend the radiologically controlled area represented a conservative and timely approach to radiological protection.
Despite that positive initiative, weak implementation indicated that radiological protection management involvement in steam generator work diverted that management from appropriate review of implementation of radiological controls in the Turbine Hal.-
_
.
-.
.
.
2.3 Control of Contaminntion During the inspection period, the inspectors identified poor material control practices for work in contaminated aleas. In several instances, cleaning materials which might have been used on contaminated equipment appeared on both sides of the contamination control boundary. Such poor material control could cause contamination to spread into clean areas, increase low level radioactive waste, and require increased expenditure of radiation protection manhours. This dencient contamination control practice warrants attention.
,
3.
. MAINTENANCE / SURVEILLANCE The inspectors observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radiological controls for worker protection, retest requirements, and reportability per Technical Speci0 cations. Noteworthy examples follow.
3.1 Diesel-Generator Surveillances The Emergency Diescl-Generator Surveillance Test Program was reviewed to determine the acceptability of diesel-generator loading during monthly surveillance tests and to compare the surveillance testing program required by Technical Specification 4.5 to that outlined in Proposed Revision 3 to Regulatory Guide 1.9, " Selection, Design, Qualification, Testing and Reliability of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants,"
dated November 1988.
The Technical Specification surveillance test program at Maine Yankee and the program recommended by the Regulatory Guide differ significantly. In most cases the testing currently conducted at Maine Yankee appears less rigorous than that recommended by the Regulatory l
Guide. For instance, the Maine Yankee Technical Specifications do not require refueling interval 24-hour endurance run tests, load reject tests or ten year interval concurrent start tests. Also, the Regulatory Guide recommends more frequent fast start tests than the refueling interval tests currently performed by Maine Yankee.
The Plant Engineering Department had previously identified and documented the differences
!
between the Regulatory Guide and the Technical Specifications. As part of establishing a diesel-generator reliability program, the licensee currently plans to revise the emergency diesel-generator surveillance test program prior to the 1991 refueling outage. This will allow enhanced refueling interval testing to be initiated during that outage. Progress toward developing
,
the reliability program was delayed due to recent unplanned engineering projects. Therefore, Plant Engineering Department management plans to dedicate additional resources to the emergency diesel-generator reliability program to maintain the above mentioned schedule.
.
-
-
-
-
.
.
._
.
.
The Plant Engineering Department's assessment of Proposed Regulatory Guide 1.9 demonstrated a good initiative to maintain emergency diesel-generator surveillance testing up to date with current industry practices. Dedication of additional resources in an attempt to maintain the scheduled implementation demonstrated the licensee's understanding of the significance of the reliability of emergency diesel-generators with regard to station blackout scenarios; however, adequacy of emergency diesel-generator surveillances is unresolved pending further review of load calculations and of the fulfillment of General Design Criterion 18 (UNR 50-309/90-025-001).
3.2 Repair of Main Steam Non Return Bypass Valse At approximately 4:15 p.m. on January 8, operators declared Valve MS-70 inoperable for its containment isolation function. MS-70 is a manual valve in the bypass line around the Steam Generator SG-2 isolation non-return valve and excess flow check valve. The valve is located at the boundary of Safety Class 2 to Safety Class 0 (non-safety) piping and serves as a containment integrity boundary valve.
During plant startup operators opened MS-70 to heat up the main steam lines and equalize pressure across the larger isolation non return and excess flow check valves. After the plant was on-line, operators attempted to close MS-70. The valve could not be closed and the valve stem threads were observed to be galled in the area of the yoke bushing.
Maine Yankee attempted to repair the valve by cutting off a portion of the valve yoke and replacing it with a longer section from a yoke from a spare valve. When the yoke was cut from the installed valve, the yoke and part of the valve stem were blown free from the valve. This caused a small unisolable steam leak from the main steam line. A plant shutdown was necessary to replace the valve.
Region 1 management met with Maine Yankee management on January 23,1991 to discuss this incident, the appropriateness of the activities, the licensee's evaluation and planned corrective actions. The adequacy of the licensee's personnel and plant safety precautions as well as safety Evaluation reviews (50.59) will remain unresolved pending further NRC review of the incident and the results of the management meeting. (UNR 50-309/90-025-002)
3.3 Maintenance Observations The inspectors observed portions of the following maintenance evolutions:
DR (Deficiency Report) 5765-90, Preventive Maintenance on PCC (Primary Component
-
Cooling) Pump P-9B and Discharge Check Valve PCC-13,
--
DR 6212-90, Resin Replacement in Steam Generator Blowdown Demineralizer (I-6),
_
Code analysis of damaged steam generator primary manway stud hole,
--
._________
.
l l
'
The MWCC (Maintenance Work Control Center) provided planning for the resin replacement in the steam generator blowdown demineralizer.
The planning process included an interdepartmental meeting which significantly aided the successful completion of an infrequently performed evolution that required support from several departments within the organization.
3.4 Surveillance Observations The inspectors observed portions of the following wrveillance evolutions:
--
Emergency and Auxiliary Feedwater Pump Test, Procedure 3.1.5, Revision 33, effective 6-27-90, for Pump P-25B.
The surveillance was adequate to assure operability of the pump.
4.
PIIYSICAL SECURITY Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those checks included security staffing, protected and vital area barriers, vehicle searches and personnel identification, access control, badging, and compensatory measures when required. No discrepancies were identified.
4.1 Chnnge of Security Contractor On January 1,1991, the contractor security services were changed from Hall Security to American Protection Services. The transition took place during December 1990. This change primarily effected management personnel and resulted in little turnover of the security staff. No changes in security effectiveness were noted by the resident inspectors during the transition period.
4.2 Fitness For Duty On Friday, December 7, the MRO (Medical Review Officer) was notified that a staff member at Maine Yankee had tested positive for marijuana in random sample urinalysis. When contacted by the MRO, the person requested verification of the results by testing of the remainder of the split sample. On Saturday, December 8, the individual admitted use of marijuana to a supervisor and resigned; site access was immediately revoked for this person. The Resident Inspectors were notified by licensee management on Monday, December 10. Licensee actions regarding this matter were assessed as timely and acceptable.
- _ _ _ - ___ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _
__
_
.
5.
ENGINEEIUNG/TECIINICAL SUPPORT 5.1 Steam Generator Eddy Current Testing In response to the primary to secondary leak in Steam Generator SG-1, Maine Yankee embarked on a thorough program of eddy current testing. The leaky tube was k)cated in the SG steam blanket region, an area of relatively short radius U-bends subjected to flow reduction. The flow reduction, caused by shadowing from internal support structures, resulted in impurities being deposited on the upper and lower surfaces of the U bend, Maine Yankee and industry experience indicates that residual stresses associated with the narrow radius U-bends, in combination with the deposited impurities, causes corrosion cracking. The steam blanket region includes rows 5 through 8 of the Maine Yankee steam generators.
.
Technical Specification TS-4.10, Steam Generator Tube Surveillances, requires inspection of steam generator tubes in areas where experience has indicated potential problems and all non-plugged tubes that previously had detectable wall penetrations. In addition, TS-4.10 required Maine Yankee to inspect a minimum of 514 tubes in SG-1, As a result, Maine Yankee elected to perform eddy current testing of 525 open tubes in Rows I through 11, and an additional 9 tubes with previously identified degradations. Experts from CE (Combustion Engineering)
performed eddy current testing and analysis.
Eddy current test results revealed 4 questionable tubes in SG 1. These were plugged at the recommendation of CE. One of these, also in the steam blanket region, exhibited a precursor of degradation similar to the leaking tube. Two pit-like degradations identified during the previous refueling outage were also plugged.
Based on the results of the testing in SG-1, TS-4.10 required inspection of an additional 1028 tubes in SG-1. Maine Yankee senior managers requested relief from the requirements for additional inspection in SG-1 (refer to Detail 6.3). The NRC granted a one-time emergency change to TS 4.10, substituting requirements for inspection of the steam blanket area of SG-2 and possibly SG-3 for additional inspection in SG-1.
Since Maine Yankee and industry experience indicates that pit-like degradations experience limited growth and have never resulted in a through wall degradation, the NRC based the change to TS-4.10 on a net safety benefit l
stemming from the additional inspection of tubes in steam blanketed regions, where potential L
problems were indicated, rather than of tubes outside the steam blanket region in SG-1, where l-the potential for additional degradation was assessed as lower.
As a result of the additional eddy current testing, Maine Yankee discovered three U-bend flaws and one pit in SG-2 and five U-bend flaws and 5 pits in SG-3. These required tube plugging.
The pit degradations were previously identified.
l-
.
Maine Yankee and CE personnel conducting eddy current testing and analysis were well-
!
qualified, conducted a thorough inspection program, and developed improved analysis methods during the Maine 'iankee inspection.
l
.
.
.
.
5.2 Reactor Trip Breaker Failure During surveillance testing of the reactor trip breaker logic trip relays on November 27, Trip i
Circuit Breaker TCB-3 was tripped open but was unable to be reclosed when actuated from the
]
main control board.
l Licensee investigation revealed that the circuit breaker's closing coil failed as a result of the disengagement of both of the operating springs.
Maine Yankee's reactor trip breakers (RTBs) are type AK-2A-25-1 Low Voltage Switchgear Breakers manufactured by GE (General Electric). There are eight breakers currently used in the Reactor Protective System and one breaker installed as a cross connect (TCB-9). Maine Yankee also has two spare breakers on site. The RTBs use two operating springs in a spring-actuated over-center toggle type mechanism. These operating springs are required for both the closing and tripping functions. The safety function of the reactor trip breakers is to trip open upon receipt of a trip signal from the RPS (Reactor Protection System). That interrupts power to the control rod drive mechanisms, allowing the rods to fall into the core, shutting it down.
Immediate licensee actions included the replacement of the failed breaker with a spare and the inspection of all installed reactor trip breakers to ensure that both operating springs were in place. Procedure changes were initiated to inspect the operating springs subsequent to operation of the reactor trip breakers.
l The Plant Engineering Department was assigned the lead to develop a closecut plan to ensure I
the issues identified as a result of the failure were adequately resolved.
The breaker manufacturer was contacted and requested to aid in the failure analysis. With manufacturer representatives on site, the breaker was tested in an attempt to recreate the disengagement of the operating springs. The failure was not re-created. The breaker was then transported to the l
manufacturer for a complete dimensional check and rebuild. The results of the dimensional l
check were not available at the time of this report.
The licensee's initial search of their OEDB (Operational Event Database) identified that a similar failure had occurred at Maine Yankee on January 12, 1990. Further investigation identified i
l similar failures at other Combustion Engineering plants, based on NPRDS (Nuclear Plant
!
Reliability Data System) and contacts with those nuclear power facilities. No instances were identified in which a disengaged operating spring has prevented a reactor trip breaker from performing its safety function, tripping open. However, it was postulated that one of the two
,
'
operating springs could become disengaged, with the other providing for apparently normal operation of the breaker until the breaker was called upon to trip. At this time the disengaged operating spring might jam the operating mechanism, preventing the breaker from opening.
Based on this hypothesis, the licensee reported the potential generic safety concern to the NRC in a letter dated 12/28/90 (MN-90-130) in accordance with 10 CFR 21.
l l
-
.
-.
-.
.
9 PED (Plant Engineering Department) aggressively pursued identiGeation of the failure mode.
PED informed plant operators of the failure mode and took appropriate immediate corrective actions to alleviate any safety concerns at Maine Yankee. The licensee's thorough review of the failure mode and postulated impact on the safety function of the reactor trip breakers identified a generic safety concern which other licensees, with similar failure experience, evidently did not recognize. Maine Yankee's efforts to identify and document the generic safety concern resulted in a beneficial safety input to facilities that use GE AK-2A-25-1 Reactor Trip Breakers.
At Maine Yankee, the RTB springs are now required to be checked after opening and closing evolutions. The inspector had no further questions.
5.3 Governor Replacement On December 11,1990, Maine Yankee personnel replaced the governor on P-25B, the turbine-driven AFW (Auxiliary Feedwater) pump, to achieve improved reliability. Work, controlled under DR (Deficiency Report) 2515-89, included engineering review and oversight.
Inspectors reviewed the engineering package, observed portions of maintenance and testing, and concluded that the engineering package was thorough and technically sound, and that the responsible engineer's close attention assured careful control of the activity.
5.4 Review of 10 CFR 50.59, Changes, Tests and Experiments The inspector reviewed Plant Procedure 0-06-4,10 CFR 50.59 Determination, Revision 3, effective 3-7-90. The procedure was found unclear, internally inconsistent and at variance with 10 CFR 50.59. Paragraph 2 of the discussion section states that a safety evaluation that should (emphasis added) be completed, whoe paragraph 4 refers to the evaluation required (emphasis added) by the procedure. Further, Paragraph I discusses tests and experiments involving an
" abnormal" mode of operation not explicitly described in the FSAR.10 CFR 50.59 applies to tests and experiments not described in the FSAR regardless of mode of operation. No tests or experiments in violation of 10 CFR 50.59 were, however, identified, and this matter appears to be a case of improper application of the word " abnormal."
The inspector also observed that a handwritten change had been made to the approved original copy of Procedure ES 1.1, with no evidence of review and approval. Maine Yankee determined that the handwritten change was reviewed and approved by PORC on July 20,1989, and that the failure to annotate the change was an isolated occurrence. Inspector review found no safety signincance to this item.
The inspector reviewed Engineering Design Change Request EDCR 88-505, Charging Pump Suction Vent Valves.
Page 7 (of 7) of the 10 CFR 50.59 determination for this plant modification indicates that the modification is bounded by the existing analysis. This appears to be an error that escaped the review and approval process for this modi 0 cation package until the package reached the PORC (Plant Operations Review Committee). During their review
_
_
.
..
.
process, PORC noted that an evaluation had not been performed for a failure associated with the closed mode for this valve. Accordingly, the cognizant engineer for this modification completed a Minor ECN (Engineering Change Notice) to address this (and other, less significant) comments on the package. In this case, the overall review function was found to nicet NRC requirements.
The inspector found the 10 CFR 50.59 review process at Maine Yankee to tm acceptable. Also, this inspection clearly identined the rigorous review and attention to detail that PORC provided.
There are, however, opportunities for improvement in the administrative procedure governing such reviews, and in the reviews these packages receive before reaching the PORC approval stage.
6.
SAFETY ASSFSSMENT/ QUALITY VERIFICATION 6.1 Control of Safety-Related Activities On December 18, 1990, operators were preparing to isolate and drain RCS (Reactor Coolant System) Loop One in preparation for SG-1 tube testing and repair. The Day Orders (DO-90-029) for December 18 instructed the operators to stroke the loop stop valves open and closed three times to clean the seats prior to isolating Loop One in accordance with OP l-10-6, Reactor Coolant Loop Draining.
Operators stroked the loop stop valves as directed by the Day Orders, leaving them closed in anticipation of the requirements of OP l-10-6. While performing the manipulations of OP l-10-6, the operators noted that Loop One RCS pressure was increasing between the stop valves.
They quickly deduced that the cause of increasing pressure was seal water being supplied to the reactor coolant pump seals in Loop One. Operators secured seal water to reactor coMant pump P-1-1, and the pressure increase was terminated.
Operations management concluded that the operators should have known that all sources of loop overpressurization should be isolated prior to shutting the loop stops, in addition, a caution in OP l-10-6 requires completion of steps in the specified sequential order, and the order requires isolation of seal water to the loop prior to closing the loop stops.
In this case, operations management assessment identified two credible factors which contributed to inadvertently pressurizing the RCS loop.
Operations management failed to recognize, however, that the loop stops were not closed in the sequence required by OP l-10-6 as a result of direction contained in the Day Orders. The instructions contained in the Day Orders for cyding the loop stops became, in effect, a change to the loop draining procedure which was not reviewed and approved prior to use. This is a violation (VIO 50-309/90-25-03).
To prevent future inadvertent pressurization of isolated loops, the licensee plans to incorporate the requirement to stroke the loop stops prior to closure into OP l-10-6. In addition, operations management plans to review the use of Day Orders for impact on plant operations.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ -
___-_
-
\\
-
.
6.2 Correction Action Procedure On January 4,1991, a hiaine Yankee Quality Programs section head previewed a draft revision of Procedure 0-16-1, Corrective Action Procedure, for the senior resident inspector. The procedure revision is a major rewrite which shifts the focus of de0ciency tracking to performance and reliability from ex crnal and internal requirements. The inspector concluded that the procedure change has the potential to focus Quality Program resources on observation of activities and away from document review, and could result in more effectise corrective action.
The procedure change is a good initiative; its effectiveness will be reviewed as part of routine inspection after the change has been implemented.
6.3 hianagement Involvement in Assuring Quality As discussed in Detail 5.1, hiaine Yankee requested relief from the TS (Technical Specification)
requirement for additional cddy current testing in SG-1. Maine Yankee based the request on the evidence that failure of the tube in SG-1 resulted from the effects associated with the steam blanket region in this case, all tubes had been inspected in the SG-1 steam blanket region, and hiaine Yankee would gain no significant additional safety benefit from inspection of tubes outside of the steam blanket region hiaine Yankee postulated that inspection of SG-2 and SG-3 was not required since there had been no evidence of leakage in SG-2 or SG-3, and Maine Yankee believed the flaws in the steam blanket regions of these generators would grow so slowly under normal operation as to preclude a through-wall defect occurring prior to the October 1991 outage. Inspection of SG-2 and SG-3 was undertaken after NRC discussion of a design basis accident such as a main steam line break on the assumptions for growth of a tube flaw and consideration that the basis for this type of inservice inspection intends that, when defects are discovered, an additional sample should be inspected in an area of similar environment to assure evidence of adequate safety margin.
Maine Yankee performed thorough inspection in the remaining steam generators. However, the above apparent weakness in Maine Yankee's understanding of the inspection basis and the initial failure to consider the effects of design basis accidents indicated a performance weakness.
6.4 Resolution of Technical Issues from a Safety Stnndpoint During plant operation from July 1990 until December 15,1990, the air ejector RMS (Radiation Monitoring System) indicated slowly increasing activity. For significant periods, the RMS was inoperable as a result of being taken out of operation. Also, the RMS was frequently in alarm, and the alarm masked omer RMS alarms. A great deal of effort during the period went into attempts, initiated by management, to determine the troubic with the air ejector RMS before management concluded that.MS indication was accurately reflecting primary to secondary leakage. It is worthy of note that the operators concluded that the indication represented actual leakage much earlier than did management. Operator belief in the instrument indication proved to be a key factor in achieving a safe, controlled shutdown on December 17, 1990.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ -
_ _ - _ _ _ _ - - -
.-
-
.
.
.
-
-
..
.
O
Other instances of a faulted approach to instrumentation were manifest during this inspection period: the operators assumed an alarm condition of the service water RMS was due to erroneous instrumentation without having a solid basis; and the reactor engineer assumed core differential pressure indications were due to instrument failure.
The above instances indicate that some Maine Yankee personnel may be predisposed to question the accuracy of instrument indications when the indication is not normal. Although this approach does not appear prevalent, it indicates a partial lack of a sufficiently questioning attitude. This consideration warrants immediate management attention.
7.
ADMINISTRATIVE 7.1 Person Contacted During this report period, interviews and discussions were conducted with various licensee personnel, including plant operators, maintenance technicians and the licensee's management staff, 7.2 Summary of Facility Activities The plant operated at full power from the beginning of the inspection period until December 17, when a tube failure in Steam Generator SG-1 forced a shutdown for repair. On January 7,1991, inspection and repairs in all three steam generators were completed, and operators began a plant heatup. Criticality was achieved on January 8; on January 9, at 10:30 p.m., the plant was sbut down from 19% power to replace MS-70, a failed main steam non-return bypass valve.
7.3 Interface with the State of Mnine Periodically the resident inspectors and the onsite representative of the State of Maine discussed fmdings and activities of their corresponding organizations. Issues discussed included the status of state rule-making regarding Maine Yankee, the technical issues related to the tube leak in the Steam Generator SG-1 and the associated Emergency Technical Specification Change. The State Nuclear Safety Inspector also accompanied the Senior Resident inspector on a tour of the expanded radiological control area. A summary of the findings based on that tour are included in Detail 2,2.
7.4 Exit Meeting Meetings were periodically held with senior facility management to discuss the inspection scope and findings. A summary of findings for the report period was also discussed at the conclusion of the inspection.
The inspection involved 252 inspection hours, including 20 backshift and 28 deep backshift hours.