ML20134F328
| ML20134F328 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/31/1997 |
| From: | Shirley Ann Jackson, The Chairman NRC COMMISSION (OCM) |
| To: | Myers H AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20134F331 | List: |
| References | |
| NUDOCS 9702070361 | |
| Download: ML20134F328 (3) | |
Text
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[A UNITED STATES f'
k' NUCLEAR REGULATORY COMMISSION WASHINoToN, D.C 2055tWOO1
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January 31,'1997 CHAmMAN Mr. Henry R. Myers Post Office Box 88 Peaks Island, Maine 04108
Dear Mr. Myers:
I am responding to the letter you sent me on October 25, 1996, in which you question the basis for the U.S. Nuclear Regulatory Commission (NRC) staff's January 3, 1996, " Confirmatory Order Suspending Authority for and Limiting Power Operation and Containment Pressure (Effective Immediately), and Demand for Information" (Order) to the Maine Yankee Atomic Power Station, which allowed operation at 2440 megawatts (thermal) (MWt), considering the plant's nonconformance with Three Mile Island Action Plan Items II.K.3.30 and II.K.3.31.
Your letter states that the NRC letter of October 18, 1996, does not address the fact that the NRC staff appears to have allowed Maine Yankee to operate at 2440 MWt without having followed procedures for allowing the plant to operate when it does not conform with TMI Action Plan Items II.K.3.30 and II.K.3.31.
As explained in the Order, the NRC staff's letters to you of June 18 and August 9, 1996, and in my letter of October 18, 1996, the Order was issued for the purpose of ensuring the safe operation of Maine Yankee pending completien of the staff's evaluation of the Maine Yankee emergency core cooling systems (ECCS) and containment design.
The Order, the NRC staff's letter of Apri; 10, 1996, and my letters of October 18 and December 5, 1996, explain in detail that the staff appropriately determined that operation at a reduced power level and with a reduced limit on containment internal pressure poses no undue risk to the public health and safety rending completion of the staff's evaluation of these Maine Yankee analg 4 Your letter requests documents showing Commission consideration of the Order issued on January 3, 1996, to Maine Yankee. No documents exist responsive to this request because the discussions between the Commission and the NRC staff regarding the order were conducted orally and were not recorded.
Your letter asks when the Director, Office of Nuclear Reactor Regulation (NRR), explained the basis for the Order of January 3,1996, and whether it Y ur letter also asks whether the Director, was before issuance of the Order.
9 NRR, explained the use of his authority provided by Section 50.46(a)(2) of Title 10 of the Code of Federal Reaulations (10 CFR 50.46(a)(2)).
As explained in the enclosure to the staff's letter of May 16, 1996, the staff's letters of June 18, July 9, and August 12, 1996, and my letters of October 18 t* y and December 5, 1996, the Director, NRR, appropriately issued the Order of January 3, 1996, pursuant to his authority under 10 CFR 50.46(a)(2).
That
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Mr. Henry R. Myers authority, although not specifically referenced by the Order, is included within the general authority cited in the Order. Moreover, the Office of the General Counsel (0GC) provided advice and counsel to the NRR staff during the development of the Order and, in OGC's view, the Order is legally sound.
Your letter states that the NRC's letter of October 18, 1996, does not indicate any analysis demonstrating that a 10 percent reduction in the maximum power level under 10 CFR 50.46(a)(2) compensates for the increased risk resulting from the nonconformance with TMI Action Plan Items II.K.3.30 and I
II.K.3.31.
Similarly, you state that there is no analysis to show the effect of reduced safety resulting from the nonconformance with TMI Action Plan Items II.K.3.30 and II.K.3.31 on the 90 percent power level limitation.
You ask whether the 90 percent limitation is a net restriction of operation under 10 CFR 50.46(a)(2) or a net relaxation of regulatory requirements. As explained in the NRC staff's letters of June 18, and August 9,1996, and in my letter of October 18, 1996, the Order did not relax regulatory requirements. The l
January 3,1996, Order clearly restricted operations at Maine Yankee, as your letter acknowledges. M:reover, as explained in my letters of October 18 and December 5, 1996, and in the January 3, 1996, Order, operation at a power i
level of 2440 MWt and with a containment internal pressure of 2 psig poses no undue risk to public health and safety.
Your letter states that the NRC staff says that a basis for the January 3, 1990, Order is that the large-break loss-of-coolant accident (LBLOCA) analysis bounds credible accidents.
You ask what analysis the NRC staff has done to develop a position regarding power levels at which the LBLOCA bounds credible design-basis accidents, thereby making TMI Action Plan items II.K.3.30 and II.K.3.31 superfluous. As explained in the NRC staff's letters of June 18, and August 9, 1996, and in my letter of October 18, 1996, the January 3, 1996, Order did not waive conformance with TMI Action Plan Items II.K.3.30 and II.K.3.31.
As explained in the NRC staff's letters of April 10 and May 16, 1996, and in the January 3,1996, Order, the NRC staff judged that the reduction in power level to 2440 MWt was necessary to account for post-Cycle 4 small-break loss-of-coolant accident (SBLOCA) model uncertainties.
As required by the Order, the licensee has submitted its evaluation that the SBLOCA for Maine Yankee, under the operating conditions for Cycle 15 at 2440 MWt, continues to be less limiting than LBLOCAs.
The licensee analysis confirmed that there is substantial margin to the criteria specified in 10 CFR 50.46, and that the additional effects of less significant parameters or intermediate break sizes between 0.1 fta and 0.5 ft2 would be accommodated.
The NRC documented the results of its audit of the licensee's calculations in NRC Inspection Report 50-309/96-01, dated April 2, 1996 (enclosed). The NRC staff considers operation at 2440 MWt, using the core operating limit parameters based upon analyses performed for operation at 2700 MWt, acceptable.
Furthermore, as explained in my letters of October 18 and December 5, 1996, and in the January 3, 19.5, Order, operation restricted to a maximum power level of 2440 MWt and a containment internal pressure limit of 2 psig poses no undue risk to public health and safety.
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Mr. Henry R. Myers Your letter asks whether Maine Yankee is in substantial compliance with NRC requirements, and whether the level of compliance has diminished to the point to which protection of public safe +,y cannot be assured in the manner required i
by the Atomic Energy Act. As explained in my letters of October 18 and December 5, 1996, and in the January 3, 1996, Order, operation of Maine Yankee at 2440 MWt and with containment internal pressure limited to 2 psig, pending completion of the staff's evaluation of the Maine Yankee ECCS and containment 4
pressure response analyses, poses no undue risk to public health and safety.
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Issues that arise at the Maine Yankee facility, such as a recent cable i
separation issue that was the subject of a December 18, 1996 Confirmatory Action Letter, and an offsite power source issue that was the subject of a l
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January 30, 1997 supplement to the Confirmatory Action Letter, will be f
evaluated for appropriate action and impact on risk to public health and J
safety.
l Thank you for the concerns that you have expressed about the operation of Maine Yankee.
I have assigned Mr. John A. Zwolinski, the Deputy Director of
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3 the Division of Reactor Projects - I/II in NRR, the responsibility of responding to future correspondence from you. However, I will continue to monitor the staff's actions related to Maine Yankee, including your correspondence.
Sincerely,
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Enclosure:
NRC Inspection Report 50-309/96-01 1
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U.S. NUCLEAR REGULATORY CO M ISSION REGION I REPORT NUM6ER.
50-309/96-01 DOCKET NUM8Er.:
50-309 LICENSEE WUM8ER:
DRP-36 Maine Yankee Atomic Power Company LICENSEE:
329 Bath Road Brunswick, Maine 04011 Maine Yankee Atomic Power Station FACILITY:
INSPECTION DATES:
January 1, to February 10, 1936 J. Yarokun, Senior Resident Insnectar IN~PECTORS:
W. Olsen, Resident Inspector R. fuhrmeister, Senior Reactor Inspector D. Mannat, Resident inspector, Seabrook Station B. Korona, Resident Inspector, Pilgrim Station E. Trottier, Project Manager, P R S. Brewer, Staff Engineer, NRR S. Sun, Staff Enginecr, NRR b
7, Dat'e APPROVED BY:
- y. f. Rogge, Chiefv V Reactor Projects Branch 8 Resident inspection and safety assessment of plant activities including operations, maintenance, engineering, and overall plant Scone:
support.
Supplemental inspection coverage of restart activities.
Verification of plant compliance with flRC confirmatory order, dated January 3, 1996, limiting power operation to 2440 MWt (90.3 reactor power) and containment operating pressure to 2 psig.
Overview:
See executive sumary.
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MECUTIVE SUMARY l
l Ooerations Operators performed well, properly followed procedures and used goodOpe comunication techniques during plant restart. An instance was noted, of plant conditions and system configurations.
however, involving the control elements and power programmer alarus where operators did not anticipate receiving the alarms during control element The outstanding work order on the coil power assemble (CEA) movement.
programer (CPP) alarms represented less than timely correction of an equipment deficiency.
The establishment sf additional licensee management oversight during the reactor startup was considered a strength. Operators were attentive and Daily plant i
effectively monitored plant parameters during reactor startup. Senior station management meetings were properly focused on safety issues.
managarant demonstrated good safety perspectives by thoroughly questioning the 4
staff regarding the emergent work on safety-related systems, such as with the emergency feedwater pump 25-A and other important key issues relating to the plant startup.
Shift operations supervisors displayed good comand and control during the performance of the reactor startup and frequently monitored control room Test activities to preclude any distractions ts the reactor operators.
personnel were knowledgeable of test procedures, and performed tests satisfactorily.
I Effective plant parameter monitoring and a good questioning attitude were evident when the reactor operator performing the startup questioned reactor engineering personnel regarding minimal source range count changes during the initial portion of control rod withdrawal.
Maintenance Maintenance activities were conducted safely. For example, efforts at addressing the problem with the leak from one of the reactor coolant pump During the (RCP) motor bearing oil reservoirs was safe and well controlled.
However, a pump assembly and test run, good management oversight was noted.
weakness was identified when personnel initially added oil to the wrong Inadequate comuatcation was attributed to the cause of the error.
reservoir.
The planning and scheduling section was effectively implementing the on-line Plant management actively participated maintenance risk management program.
The in discussions concerning outage times of safety significant equipment.
overall plant risk factors were calculated and considered on a daily basis.
In general, system outage times were minimized when they occurred, and personnel maintained the proper safety focus to minimize the impact on safe plant operation.
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Enoineerina The licensee used adequate methods in the analysis to supp operation at 2440 MWt. operating limits (COLs), established at 270 Maine Yankee and not relying on RELAP5YA, bound the operating condit modifications, which originally reifed nn RELAP5YA current operating cycle.
Actions taken to ensure reactor power ifsitation to 2440 MWt and contain operating pressure to less than 2 psig to comply with the conffrmator issued by the NRC were very thorough.
Maine Yankee demonstrated good capability to identify and resolve complex problems as shown with the resolution of the problems with the emergency feedwater isolation valves.
Safety Assessment /Ouality Verification Plant operations review committee meetings were conducted well with evi For example, the procedural controls for limiting the reactor power to 2440 MWt and the containment internal operating pres focus on safety.
psig were properly discussed focused on safety.
Plant management provided good support to the implementatio I
Maintenance Risk Management Program.
maintenance was properly discussed at the daily morning meetings so that no unnecessary increase in on-line safety assessment plant condition was presented.
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TABLE OF CONTENTS 11 l
EXECUTIVE SUfEARY I
1.0 OPERATIONS............................
2 Plant Heatup........................
2 l
1.1 Reactor Startup.......................
4 1.2 Low Power Physics Testing..................
5 1.3 Power Ascension Tests....................
5 1.4 Chemistry..........................
5 J
1.5 Reactor Coolant Pump #1 011 Leakage Problem.........
1.6 6
MAINTENANCE............................
6 2.0 Maintenance Observation...................
7 2.1 2.2 Surveillance Observations.........
7 ENGINEERING...........................
7 3.0 Emergency Feadwater Isolation Valves Post Steam Generator Tube Sleeving Reactor Coolant System 3.1 8
3.2 Flow Rate..........................
9 f
PLANT SUPPORT...........................
9 4.0 4.1 Radiological Controls....................
10 1
4.2 Security..........................
10 4.3 Emergency Preparedness.................
10 SAFETY ASSESSMENT / QUALITY VERIFICATION 10 i
a Plant Operations Review Committee (PORC) Meeting 11 l
5.0 5.1 On Line Risk Management...................
5.2 Il NRC CONFIRMATORY ORDER DATED JANUARY 3, 1996............
Description of Justification for Cycle 15 Operation at 2440 6.0 13 l
6.1 MWt.............................
13 1
6.1.1 Effects of Model Changes 13 6.1.2 Effects of Plant Modifications 14 6.1.3 Core Performance Analysis Report (CPAR)........
6.2.1 Operability Determinations for Other RELAP5YA 16 Applications.....................
18 Power limitation to 2440 MWt................
6.3 Containment internal operating pressure limitation to 2 i
18 6.4 psig............................
19 7.0 ADMINISTRATIVE..........................
19 7.1 Persons Contacted......................
19 7.2 Summary of Facility Activities...............
s 19 7.3 Interface with the State of Maine..............
19 7.4 Exit Meeting........................
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EETAILS 1
1.0 OPERATIONS the reactor was made critical for the first time sinceL 11, 1996, On January the plant was =hutdown in January of 1995. initiated in accordance "it Power The operations was attained iater when reactor power was increased above main generator was initially phased onto the grid on January Following power reduction to conduct startup tests and correct equipm 18, 1996. On January 22, a.m.
problems, final phase on was achieved on January 1996, the pla confirmatory order (2440 MWt, (90.03%)) dated January 3, 1996.
NRC inspectors maintained a round-the-clock coverage of 16, 1996. The inspectors During plant restart.
startup activities.om January 10 through January observed Maine Yankee operations department personnel perform a s Tae reactor and the plant in accordance with the procedures listed below.
h tdown procedures prescribed the processes for operating the plant from Hot S to taking the reactor critical (Hot Standby) and going up t appropriate access control, adherence to procedure Power.
protective systems, including emergency power sources.
j verified operability of selected Ergineered Safety Features (ESF) trains assessed the condition of plant equipment, radiological controls, securi The inspectors monitored the status of control room annunciators radiation monitors to ascertain that they were being respond safety.
maintained adequately.
tags (temporary modifications) prior to startup to ensure that good configuration was maintained and that equipment and systems were l
The inspectors evaluated plant housekeeping and cleanliness to ascertain plant condition and ensure no detrimental effect on plant safety condition.
present.
The inspectors attended the daily plant management meetings and fo i
Senior station management meetings properly focused on safety issues.
demonstrated good safety perspectives by thoroughly questioning the sta regarding the emergent work on safety-related systems, such as with th emargency feedwater pump 25-A and other important key issues rela plant startup.
The inspector reviewed the following procedures:
Procedure No. 1-1, " Plant Heatup," Rev.48 Procedu.e No. 1-2, " Reactor Startup," Rev. 33 Procedure No. 1-3, " Plant Startup," Rev. 45 Procedure No. 11-2, " Low Power Physics Testing," Rev. 18 Procedure No. 11-3, " Power Escalation Tests," Rev 14 The procedures were found to be technically accurate and proper fo evolutions for which they were intended.
appropriate controls necessary for reacter power and containment pressure limitations had been properly incorporated into the procedures.
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1.1 Plant Heatup l
On January 4,1996, the inspector observed station op i
l The on-shift personnel closely l
procedure 1-1, Plant Heatup, revision 48.
monitored and plotted the actual temperature and pressure relationship in accordance with Maine Yankee Technical Data Book (TDB leg wide range resistance temperature detectors (RTD's) and core exit Curve.
l themocouples (CET's). The core region subcooling was maintained such that i
adequate net positive suction head was available for the reactor coolant pumps t
j during the plant heatup.
On January 5, the inspector observed station Instrument l
accordance with station procedure 3-6.2.1.19 Rod Drop Time Test and l
functional Test, revision 12.. The inspector noted that the assigned personnel t
l were knowledgeable of the procedure, and that the required testing i
performed satisfactorily.
discrepancies with supervisory personnel before continuing on w procedure.
The inspector reviewed the order initiated to correct the identifled concern.
completed procedure and verf fled that all the required test data was properly l
i recorded and within the acceptance criteria.
1.2 Reactor Startup
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l The inspector observed operators start up the reactor in_accordance withThe l
station procedure 1-2, Reactor Startup, revision 33.
the procedure, held discussions with both operators and operations management l
representatives, walked down the main control boards and back panels, reviewed i
i operator logs and turnover sheets, and independently verified procedure and plant technical specification compitance.
f the insper. tor observed plant operations personnel perform On January 10, 1996, As was being done with all the nr M quisites for starting up the reactor.
shi ts, prior to the on-coming shift personnel assuming their duties, the a
actiag operations department manager conducted a pre-shift briefing concerning f
i the NRC order that limited Maine Yankee's reactor power to 90% and operating The order requirements were incorporated in containment pressure to 2 psig.
station procedures and all operations personnel were trained on the The procedures that were requirements of the order prior to assuming a shift.
j required to be revised due to this order were discussed in detail during the pre-shift briefings to ensure that all operators were knowledgeable of the l
order requirements prior to assuming their duties.
On January 11, the inspector observed operations personnel conduct the 4
T h reactor j
approach to criticality in accordance with the startup procedul A decision was made by the on-shift personnel to use rod at 10:20 p.m.
withdrawal during the criticality approach instead of boron dilution for r., ore j
j responsive reactivity control. The inspector found this to be well thought of l
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3 The licensed and indicative of a focused and knowledgeable operations crew.
l reactor operator performed the operation in accordance with the procedure in an excellent manner with good caution and attention to detail that ensured a The shift operations supervisor displayed very good j
shfe reactor startup.
l comand and control during the performance of the reactor startup and frequently monitored control room activity to preclude any distractions to the Access to the control room was limited to only those reactor operator.
An on-shift briefing was also presented by the required to be present.
reactor engineering section head in accordance with procedure 0-6-9, Infrequently Performed Procedures. He reiterated Maine Yankee's station management philosophy (caution and conservatism) when performing these type of A discussion of the expected positive temperature coefficient vice tasks.
negative' temperature coefficient, early in the fuel cycle with the resultant The chain difference in reactivity control was emphasized to the operators.
of command was also reiterated and a discussion of the next activity (Low The inspector Power Physics Testing) requirements and expectations ensued.
determired that Maine Yankee properly briefed the on-shift operations andA rector engineering personnel with goed safety perspective in evidence.
Maine Yankee qual: Ly programs inspector also conducted an independent assessment of the plant startup in the control room.
Effective plant paraneter monitoring and a good questioning attitude were evident when the reactor operator performing the startup questioned reactor engineering personnel regarding minimal source range count changes during the Reactor Engineering indicated that early portion of control rod withdrawal.
the condition was normal given the extended reactor shutdown.
However, the inspector noted that unexpected coli power programmer (CPP) alarms came in during initial control element assembly (CEA) withdrawal.
After initial investigation, operators discovered that an outstanding work indicated that the alarm would be received any time a CEA was order (#94-4203)
The inspectors considered this a minor weakness regarding knowledge of moved.
plant conditions since operators did not anticipate this alarm nor did they Additionally the inspector initially understand why the alarm was received.
considered that the outstanding work order represented less than timely Although work was completed on the rod correction of an equipnent deficiency.
position indication system during the shutdown, several indication problems remained for at least five CEAs including a load sequencing failure, no upper electrical limit (UEL) indication, dual indication of UEL and intermediate position, and improper rod bottom indication when the bank of rods was withdrawn.
Overall, the inspectors determined that the reactor startup was performed The inspectors observed that operators properly safely and effectively.
Except for the followed procedures and used good communication techniques.
j instance involving the CPP alarms, operators were knowledgeable of plant i
Roles and responsibilities for the conditions and system configurations.
The establishment of additional reactor startup were clearly established.
licensee management oversight during the reactor startup was considered a Operators were attentive and effectively monitored plant parameters strength.
Inspectors observed strong command and control and during reactor startup.
procedure adherence during approach to criticality activities using procedure i
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l-2, Reactor Startup. The inspectors verified proper calculation of 1/M In addition, reactor operators were values by reactor engineering personnel.
alert to nuclear power instrumentation and independently calculated the 1/M The criticality approach was by boron dilution and the calculated values.
criticality was at a boron concentration of 1,462.5 ppm.. Upon achieving criticality, the inspector observed the following parameters:
RPS Wide Range Log Channels: RPS A, B, C, and D at 5 X 10" each CEA positions:
Shutdown Groups A, B, and C at 184 each Control Groups 1, 2, 3, and 4 at 184 each Control Groups SA and 5B at 128 and 130 respectively RCS Boron Concentration:
1,462 ppm RCS Tave:
530 'f The difference in actual critical condition from the estimated {l462.5 versus I
1462 ppm (.004% delta rho)) was negligible. The inspectors concluded that l
operators and reactor enqineering personnel demonstrated excellent technical capability and conducted startup activities well.
1.3 Low Power Physics Testing The inspectors observed the performance of low power physics testing (LPPT) in accordance with station procedure 11-2, Low Power Physics Testing, Revision 18, starting early in tne mornir.g of January 12.
In addition to directly observing the testing, the inspectors reviewed the procedure and held several discussions with Reactor Engineering personnel and involved control room j
operators.
Inspectors independently verified selected data for accuracy, j
The proper calculation and satisfaction of procedure acceptance criteria.
difference between predicted results and actual results were well within the allowed tolerance for critical buron concentration, Moderator Temperature Coefficient (MTC), and control rod worth.
The testing was also observed by a representative of Yankee Atomic Engineering Power level Company in support of Maine Yankee's Reactor Encineering Group.
was controlled based upon the indications of the wide range nuclear The test director worked closely with the engineers and licensed instruments.
J operators to ensure that the prerequisites were met that the procedural guidance and limitations were followed and that the test was conducted safely
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and satisfactorily.
The inspectors noted that the tests were performed safely in a controlled and Communications surrounding the numerous reactivity changes deliberate manner.
required by the low power physics testing were good, Reactor Engineers The coordinated effectively with licensed operators during the activity.
inspectors verified that low power physics test acceptance criteria were Selected independent CEA worth measurement calculation j
satisfactorily met.
were performed by the inspector. Reactor Engineering personnel were i
knowledgeable of the LPPT process. Reactivity computer t.alibrations were current.
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5 1.4 Power Ascension Tests Reactor power escalation was commenced early in the morning of January 16.
Prior to raising power, the Shift Operating Supervisor briefed the control room operating crew on the expected activities, and on precautions for The licensed operating wi'h a positive moderator temperature coefficient.
operators exhibited good communications, command, and power escalation.
Following completion of low power physics tests, power ascension was begun.
16, 1996, the main generator was initially At about 12% power, on January At about 18% power, problems with a valve (SCC-T-227) phased on to the grid.
in the main generator cooling gas coolces required a power reduction to below 10% so that the generator could be taken off the grid for repairs to the Later on, power escalation was resumed and following other tests valve.
including the turbine overspeed trip test, a final phase on was achieved on January 18, 1996, at 6:55 p.m.
On January 22, 1996, the plant attained 90%
pnwer (2427 t.Jt, 846 MWe). Nuclear Instrumentation (N w; channel C, Delta 1 Boron concentration Power was at 90.1% and Channel D Nuclear Power at 90.2%.
1 All control element assembly (CEA) rod groups were at 184 1 was 1,096 ppm.
steps except for Group SA which was at 180.
1.5 Chemistry The inspector observed r.,per sampling techniques by a chemistry technician taking steam generator blowdown water samples.
The purpose of the water test was to identify the levels of fluorides, sulfates, and chlorides in the steam generators as there are specific limits on the maximum levels that may beThe present prior to changing plant power level.of the ion chromatography equ
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.ie The inspector observed proper preparation of the samples, use o history.
The inspector ion chromatography equipment, and evaluation of the results.
also verified compliance with the once ptr day sampling frequency required by chemistry procedure 3-7-4-2, Secondary System Chemistry Surveillance.
1.6 Reactor Coolant Pump #1 Oil Leakage Probles l
On January 28, 1996, Maine Yankee operations personnel noted that the upper j
bearing temperature for reactor coolant pump (RCP) number (#) I was slowly A decision was made to add increasing and that the oil level was decreasing.However, 3.5 gallons of oil were inadv oil to the bearing reservoir.
After added to RCP #2 upper bearing oil reservoir due to personnel error.
discovery of the error, maintenance personnel subsequently added 4 gallons of With the addition of oil the upper oli to RCP #1 upper bearing oil reservoir.
bearing temperature deceased slightly and stabilized at approximately 136*
Fahrenheit.
31, 1996, the licensee determined that with the increase in Later, on January bearing temperature and decrease in oil level, mo-e oil was needed in the A plan was developed to further upper bearing reservoir for RCP #1.
investigate the problem including making a loop entry to try to determine the
6 reason and source of the oil leakage, the destination of the oil being lost from the bearing reservoir and the reasonableness and safety of further oil addition.
The inspector attended the briefing for the containment and loop #1 entries.
The briefing was very comprehensive and involved upper licensee management.
Personnel and equipment safety were properly considered. The plan was to enter and conduct visual inspection of the sight glass for oil level and the areas for any oil splashes or spill. Then personnel would add oil as required There were very and drain the collection tank for oil leakage measurement.
detallad task, safety, and radiological work safety briefings. The plant He was very shift supervisor (PSS) was designated the working party leader.
aware of task termination and/or plant shutdown criteria as discussed during Work orders (WO) #96-563 and 96-5v3-01, were generated for the the briefing.
Entry into the loop was to be made under radiological work work activities.
permit (RWP) #96-102, which was also discussed during the briefing.
Thr results of the inspection identified that the leak was from one of the motor bearing oil reservoir inspection doors at about 2 drops every 20 seconds.
No oil splashes or spills were observed. Personnel added about 9 gallons to the RCP upper reservoir and drained about 11 gallons from the collection tank.
The inspector concluded that the licensee's efforts at addressing this problem had been safe and well controlled. However, a weakness was identified when personnel added oil to the wrong reservoir initially.
Inadequate communication was attributed to the cause of the error.
2.0 MAINTENANCE Overall, maintenance and surveillance activities continue to be performed The inspectors ascertained that activities were performed safely and in well.
accordance with approved plant procedures.
2.1 Maintenance Observation The inspectors observed and reviewed selected maintenance activities to assurc that the activities were conducted safely; complied with technical specificattons and work order (WO) requirements; that required approvals and releases w..e obtained prior to commencing work; that the work procedures were appropriately detailed and followed; and that equipment was properly tested and returned to service. The inspectors observed portions of the following work activities nd noted no significant discrepancy:
WO 94-04057, SW-24, Replace Bolting i
e WO 96-0563, #1 RCP Oil Addition e
e WO 96-0736, Trcubleshoot and repair of 4KV circuit breaker j
WO 96-0516, Repair of P-38 e
The inspectors also observed portions of work activities on emergency feedwater pump, P-25A under work order 96-00102-02, Disassemble, inspect and reassemble the P-25A spare rotating assemble. There were goed foreign material exclusion (FME) controls in place during the maintenance activities.
Exposed ends of pipes were taped to preclude the entrance of foreign material i
7 Good quality control (QC) and vendor support was noted l
into the system.
during the reassembly. Mechanics properly followed instructions in the work package and assembly procedure. During the post maintenance testing, a steady oil leak was observed frcm the pump motor end bearing area. The bearing The oil level casing was inspected and the oil level was found to be high.
was restored to the proper level and the subsequent functional and inservice tests run were completed satisfactorily. The inspectors noted while tha overall maintenance activity had been conducted well, the excessive oli,ssue represented a minor weakness sinca the proper amount of oil should have been added in the first case, thereby precluding the need for a rework. During the
'1 pump assembly and test run, good management oversight was noted.
2.2 Surveillance Observations The inspector observed and reviewed selected surveillance activities to assure that the activities satisfied technical specification requirements; that personnel adhered to administrative and surveillance procedures; that test l
ir.truments were calibrated; and tFat test results satisfied the acceptance j
appropriate actions criteria and when they did not, that the licensee toolr
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in addition to the tests discussed in section 1.0, the inspectors observed portions of the following tests and noted no discrepancy:
Station Procedure 3 1-2.4, Routine ECCS Testing Station Procedure 3- :.2.1.19, Rod Drop Time Test and Functional Test e
e 3.0 ENGINEERING The engineering department continued to provide good support to the plant and maintained good safety perspective.
3.1 Emergency Feedwater Isolation Valves During the inspection period Maine Yankee continued to experience problems with the emergency feedwater (EFW) steam generator isolation valves (EFW-A i
These valves failed surveillance leak testing in January 338,339 & 340).
After they were removed, their seat rings were found to be damaged.
1996.
The valves had previously been repaired to modify the valve seat ring due to excessive " crush" by re-machining the seat retainer groove and the seat ring to a new %" radius as directed by the valve vendor.
The valves are butterfly valves manufactured by Contromatics corporation. The valves' disks are asymmetrical and have seating surfaces similar to that in ball valves.
The "W isolation valves are required to isolate emergency feedwater flow to a faulted steam generator in an accident situation involving a steam line break.
The valves receive a closure signal when the associated steam generator When the valves receive a signal to pressure is at 400 psig and decreasing.
close, the faulted steam generator would be at about 400 psig and the upstream pressure is conservatively assumed to be at pump shutoff head of approximately This would thereby create a differential pressure of about 1,100 1500 psig.
i psid across the valve.
It appears that being designed for this type of a high differential pressure across the valve, the seats have had an inability to
l 8
withstand the pressure and/or flow forces developed by the normal dynamic I
l conditions of the valve, Maine Yankee engineering and maintenance personnel investigated the problems 1
i with the valve and after consultation with the valve manufacturer developed When the modifications were modifications to resolve the valve problems.
ccapleted on the /alves, they were taken to a test laboratory where subsequent The valves were testing determined that the modifications were successful.
also tested satisfactory after they were installed in the emergency feedwater system.
Test results showed very minimal leakage.
The inspector determined that Maine Yankee demonstrated good capability to identify and resolve a very complex problem. Engineers conducted detailed review of the situation and were very proactive at getting the vendor to test There were good and approve the appropriate modifications on the valves.
engineering and maintenance skills shown.
Post Steam Generator Tube Sleeving Reactor Coolant System flow Rate 3.2 The licensee applied for an amendment to the Technical Specifications on April to allow the use of the Westinghouse Electric Corporation sleeving 14, 1995, The amendment process for repairing the Maine Yankee steam generator tubes.
was issued on May 22, 1995. Between June and December 1995, Maine Yankee installed sleeves on the inlet side of all steain generator tubes (each of three steam generators contains approximately 5,700 tubes, for a total of approximately 17,000 sleeves). The sleeves were fabricated from Alloy 690
(
material, while the parent (existing) steam generator tubes are Alloy 600
)
material. Three sleeve lengths were used:
12, 20, and 30 inches, distributed within each steam generator as follows:
12" 20" 30" total S/G #1 1,329 3,410 706 5,445 S/G #2 1,963 3,141 387 5,491 S/G #3 1,873 3,089 531 5,493 The remaining tubes in each steam generator were not candidates for sleeving, because of the type or location of the existing defect. Those tubes were plugged.
Each sleeve length presents a slightly different hydraulic resistance to RCS flow, with the longest sleeve (30 in.) presenting the most resistance, and the To assist in determining the t.ombined shortest (12 in.) presenting the least.
effect on RCS flow of both sleeves and plugs, a " sleeve-to-plug" ratio is Because each of the three sleeve lengths presents a different hydraulic used.
resistance to RCS flow, each sleeve has a different total that represents the At Maine Yankee, the approximate sleeve-to-plug same resistance as a plug.
50:1 for 12 in. sleeves; 37:1 for 20 in. si e ves, and 28:1 for 30 ratios are:
in. sleeves.
That is, approximately 50 of the 12 in, sleeves present the same hydraulic resistance as one plug, while 37 of the 20 in s'eeves, and 28 of
O As the 30 in. sleeves are equivalent to the flow resistance of one plug.
noted above, each steam generator has a different combination of sleeve When the combination of sleeve lengths and total number of sleeves installed.' number installed in each gene lengths install n each generator, the total can be referred to as a steam
" effective" plugging.
(Effective plugging is actual plugs plus generat.
equivalent plugs.) At the conclusion of the 1995 steam generator outage, Maine Yankee's approximation of effective plugging for each of its steam generators was:
S/G #1 402 effective plugs S/G #2 351 effective plugs S/G #3 350 effective plugs, This was equivalent to an average of 367 plugs per generator, or approximately Maine Yank:e's cycle 15 analysis 6.s% of available tubes effectively plugged.
allows a maximum of 1000 effective plugs per steam generator, with a maximum An NRC specialist asymmetry between steam generators of 500 effective plugs. insp ratios and found it acceptable.
At the conclusion of the sleeving campaign and before reactor startup, the licensee had predicted a decrecse in reactor coolant system (RCS) flow ofThis about 2,800 gallons per minute (gpm) from the cycle 14 operating flow.
decrease in flow was expected because of the predicted pressure drop of about 1.7 psi across the steam generators as a resbit of the sleeves and plugs When flow is calculated based on the enthalpy rise across the installed.
core, a flow decrease of approximately 4,000 gpm results.
Using the newly installed and highly accurate loop flow instruments, the licensee determined that the current Maine Yankee RCS flow rate was approximately 375,000 gallons per minute.
The flow uncertainty associated 10,000 gpm, which resulted with the new instrumentation was calculated to beThe minimum RCS flow required by in a possible minimum flow of 365,000 gpm.
Technical Specification 2.1.1.d is 360,000 gpm, therefore Maine Yankee was restarted within the requirements of the Itcense for RCS flow.
4.0 PLANT SUPPORT activities in the areas of radiological controls, security, and I
Plant support The emergency preparedness were conducted safely during this period.to requirements and f
inspectors monitored work practices, and con ormance procedures.
4.1 Radiological Controls inspectors routinely reviewed radiological controls including Organization and Management, external radiation exposure control anc tantamination control.
The inspectors also monitored standard industry radiological work practices, and conformance to radiologica' control procedures and 10 CFR 20 requirements.
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10 4.2 Security The inspectors verified that security conditions met regulatory requirements, the requirements of the physical security plan, and complied with approved The inspectors observed security staffing; protected and vital area barriers; vehicle searches and personnel identification; access control; procedures.
badging; and to assure that they were in accordance with requirements and that appropriate compensatory measures were used when required.
4.3 hergency Preparedness The inspectors verified that the emergency response facilities were well maintained and kept ready for use in emergency situations.
5.0 SAFETY ASSESSMENT / QUALITY YERIFICATION 5.1 Plant Operations Review Comittec (PORC) Meeting On January 7, the inspector attended a combined Plant Operations Review I
Comittee/ Nuclear Safety Audit Review Comittee meeting at the Maine Yankee The purpose of the meeting was to review and coment on the corporate office.
Maine Yankee response to the NRC Order and Demand for Information received concerning the allegation regarding the development and use of RELAP5b LOCA The Maine Yankee Vice President of Engineering and Licensing provided code.
background information regarding the coordinated response to the allegations in the area of Yankee Atomic Electric Company's development and use of Also the actions of the Allegation Response Team and RELAP5YA computer code.
the two Independent Review teams were reviewed.
The overall conclusion of these efforts was that the current licensing basis was appropriate, however areas for improvement were identified.
The company president attended the meeting and advised the comittees that their function was to review the information to be submitted to the NRC in response to the demand for information and verify that it was complete and accurate, and he was present only as an interested observer.
l Members of the PORC questioned what changes needed to be made to th 4
Containment Weight of Air Monitoring Program to account for the new upper limit on containment pressure. The program provides on-line containment leakage monitoring, and may require modification to reflect the revised PORC members requested that the program be reviewed and any pressure limit.
necessary revisions be made and provided for their review.
The PORC met again on January 8, 9, and 10 to review the completed analyses On January 8 and 9, 1996, the inspectors attended and procedure revisions.
the meetings during which the procedural controls for limiting the reactor 1
power to 2440 MWt and the containment internal operating pressure to 2 psig The inspectors noted that the discussions were were discussed and approved.
The very thorough and safety focused and appeared technically sound.
inspectors independently reviewed some of the affected plant procedures and not M no issues not being addressed by the PORC.
4 i
11 5.2 On Line Risk Management i
Upon returning to power operations, Maine Yankee comenc
'l 10, 1996, directs that planned on-line maintenance be goes into effect July The Maine evaluated and scheduled to minimize the impact on plant safety.
Yankee program document outlines definitions, defines responsibilities and i
Assessment is accomplished by a i
describes the actual assessment process.
computer-based model, designed from inputs from the Maine Yankee Probabilistic i All Maintenance rule identified risk significant Risk Assessment (PRA).
systems are considered as well as significant external events when making a The PRA risk achievement worth of each system, sub-l risk determination.
system, train or individual component is used as a weighing factor when calculating the overall relative margin of safety for key safety functions and i
The assessment model is used to develop schedules that f
the overall plant.
minimize risk associated with on-line maintenance and also to assess the i
impact of equipment failures or other unscheduled activities on plant safety.
The planning and scheduling section is responsible for maintaining the On-Line Risk Management Program.
i The inspector determined that the planning and scheduling Section was i
ef fectively implementing the on-line maintenance as described in the program.
Plant Management actively participated in discussions concerning outage times i
l of safety significant equipment. The overall plant risk factors were i
calculated and considered on a daily basis,
)
6.0 NRC CONFIRMATORY ORDER DATED JANUARY 3, 1996.
On December 4, 1995, the NRC received an allegation against Yankee Atomic H
Electric Company (YAEC), acting as agent for the Licensee (Maine Yankee Atomic i
Power Company, or MYAPCo).
In brief, it was alleged that YAEC knowingly performed inadequate analyses to support two license amendments to increase It was further the rated thermal power at which Maine Yankee may operate.
alleged that Maine Yankee was cognizant of these inadequate analyses, yet misreprasented them to the NRC in seeking the license amendments, which were granted.
As a result of these allegations, the Lu enducted a technical rmew and evaluation of the circumstances and recorde surrounding the cppi m ions to increase MY's maximum rated thermal power. This review and evaluation was conducted at YAEC Headquarters in Bolton, Massachusetts, on December 11-14, 1995, by a five-member NRC team.
The NRC team was accompanied by two representatives of the State of Maine.
On December 18, 1995, the NRC held a public, transcribed meeting in its This meeting was held to afford Maine headquarters in Rockville, Maryland.
Yankee and YAEC the opportunity to further describe the use of computer code RELAPSYA at Maine Yankee, and to present information to the NRC to help the NRC determine if its regulatory requirements related to small-break loss-of-The NRC determined coolant accident (SBLOCA) were satisfied for Maine Yankee.
that RELAPSYA, which was proposed for use in the current operating cycle (cycle 15) SBLOCA analyses to demonstrate, in part, compliance with emergency j
12 core cooling requirements specified in 10 CFR Section 50.46, had not been applied in a manner conforming to the requirements of 10 CfP Part 50, Appendix K, "ECCS Evaluation Model," nor had RELAP5YA been applied in accordance with the conditions specified in the staff's related safety evaluation dated January 30, 1989.
Maine Yankee notified the NRC staff by telephone that On December 19, 1995, On there were other uses of RELAPSYA in appilcations other than SBLOCA.
December 22, 1995, Maine Yankee submitted a letter documenting their comitment to limit reactor power to 2440 MWt and containment pressure to 2.0 Further, Maine Yankee comitted to document the justification for use psig.
of operating cycle 15 limits, using methods approved for Maine Yankee and Finally, Maine Yankee committed to conduct a without reliance on RELAP5YA.
review to identify other applications of RELAPSYA to be used in cycle 15 and verify that operability, as defined in its Technical Specifications, of affected systems and components is maintained.
On January 3, 1996, the NRC issued a confirmatory order and demand for information to Maine Yankee pertaining to the requirements for restart and return to operation at 2440 MWt, and the requirements for return to operation at the currently-licensed maximum power level of 2700 MWt.
On January 10, 1996, Maine Yankee submitted the information required in Section IX of the NRC's January 3, 1996, order and demand for information.
The submittal satisfied the four " restart" requirements of the order.
the Maine Yankee reactor was made critical, and on On January 11, 1996, January 16, 1996, the Maine Yankee generator was connected to the Maine electrical distribution grid.
an NRC inspection was conducted at YAEC Headquarters On January 24-26, 1996, in Bolton, Massachussets.
The purpose of this inspection was to review and verify the detailed files and computer analyses supporting the licensee's submittal cf January 10, 1996.
Specifically, the evaluations performed to justify the folkwing:
(1) use of Cycle 15 operating limits without reliance on RELAP5YA, (2) the operability determinations associated with all other applications where RELAP5YA is relied on for Cycle 15 operation, (3) the measures taken to limit reactor operation to a maximum thermal power of 2444 MWt (approximately 90% of 2700 MWt), and (4) the measures taken to limit containment internal operating pressure to a maximum of 2 psig.
In addition, the team reviewed the reactor coolant system (RCS) flow calculations and measurements to determine the effect of the steam generator repairs performed during the 1995 refueling outage. The inspector's findings in each of the areas are discussed below.
The findings with regard to RCS flow are documented in Section 3.0 of this report.
l 13 Description of Justification for Cycle 15 Operation at 2440 MWt 6.1 The team inspected and evaluated supporting analyses the licensee use' to estimate the effects of model changes and plant modifications on peak cladding i
temperature (PCT), ar.d the cycle 15 core performance analysis report.
1 6.1.1 Effects of Model Changes The ifcensee based SBLOCA analyses for cycle 15 operation at 244'O MWt on its The cycle 4 analyses were performed assuming a power level cycle 4 analyses.
of 2630 MWL and used Combustion Engineering (CE) analysis methods that were I
The CE methods do not involve the use of RELAP5YA.
approved by the NRC.
However, since cycle 4, CE has modified its methods and Maine Yankee has Therefore, the staff reviewed the licensee's implemented plant modifications.
i assessment of the impact of these changes on PCT calculations.
The changes in the CE methods for SBLOCA analyses between cycle 4 and cycle 15 (1) a revised 56 heat transfer model to account r.. the effects of condensation in the primary side when the SG tubes Jrain, and (2) addition of are:
a level swell model to account for variations in drift velocity as a function i
of pressure and the effects of the power shape on the bubble production rate, 6.1.2 Effects of Plant Modifications Modifications (including changes for cycle 15 operation) to the Maine Yankee plant resulted in changes in (1) moderator temperature coefficients, (2) power shape and peaking factors, (3) flow resistance effects due to SG plugging, (4)
SG heat transfer coefficient, (5) cold leg temperature, (6) fuel rod heating effects, and (7) power level.
The licensee has assessed the effects of plant modifications and CE method Using the cycle 4 calculated PCT changes on the results of SBLOCA analyses.
as the baseline, changes in CE methods and plant modifications were These individually assessed to determine each incremental change in PCT.
changes were added to or subtracted from the PCT value for cycle 4 to The results of this incremental calculate a PCT estimate for cycle 15.
calculation show that the estimated maximum PCT for an SBLOCA is less tha To determine the conservatism of the assessment, the licensee 1760 'f.
performed a second analysis.
The second analysis used the NRC-approved CE method of 1977--with a Maine Yankee plant model representing cycle 15 operating conditions--to perform a direct calculation of PCT at 2440 MWt for a spectrum of break sizes.
The results of this second analysis (direct calculation) show that the calculated maximum PCT for an SBLOCA is les The results of both analyses confirm that there is substantial 1620 'f.
margin between the more conservative maximum SBLOCA PCT of 1760 'f and the limit of 2200 af specified in 10 CFR 50.*6, and between the same more conservative maximum SBLOCA PCT of 1760 *f and the LBLOCA maxim Therefore, the analyses provide the basis
- F at 2700 MWt for Maine Yankee.
for conclutiing that the PCT for an SBLOCA at 2440 MWt is bounded by the PCT for a LBLOCA at 2700 MWt and the LOCA-related operating limits for Cycle 15 are restricted by the LBLOCA limits.
l I
i 14 6.1.3 Core Performance Analysis Report (CPAR)
The team inspected the supporting analyses the licensee used to develop the core performance analysis report (revision 2) for Cycle 15 operation at 2440 The supporting analyses formed the bases for operation to 4000 Mwd /mt MWt.
cycle exposure, and the assumptions used for the loss of feedwater event analysis.
In the original cycle 15 CPAR, the licensee used NRC-approved methods toThus, none perform design basis L8LOCA analyses and non-LOCA transients.
the cycle 15 core operating limits (COLs) for operation at 2700 MWt were limited by SBLOCA analyses and therefore the COLs were not defined by RELAP5YA The licensee also performed core physics analyses for core depleoon up to 4000 mwd /mt cycle exposure at 2440 MWt and confirmed that the analys*s.
defects, and coefficients and kinetics parameters) is bound These core the determination of the original cycle 15 COLs at 2700 MWt.
physics analyser provide the basis for concluding that the original Cycle 15 COLs remain valid for operation at 2440 MWt, up to 4000 mwd /Mt cycle exposure.
The licensee will reevaluate selected physics parameters should it operate beyond 4000 mwd /mt cycle exposure at 2440 MWt.
The team found that cycle 15 CPAR credits the automatic steam generator (SG) blowdown isolation on low SG level for mitigating the consequences of a loss-This is a design-basis event (DBE) as analyzed in Chapter of-feedwater event.
The 14, Safety Analysis, of the plant's final safety analysis report (FSAR).
SG blowdown isolation system was recently installed and is designed to safety-related requirements, including redundant Class IE power supplie*.,
separate trains, actuation from the reactor protection system, and redundant The licensee valves procured and maintained to Quality Program standards.
relies on administrative procedures to control operability of the SG blowdown isolation system, but proposed to add the blowdown isolation system valves to the Technical Specifications when the issue was raised during the inspection.
The current procedures allow 3 days of operation if the system is inoperable, and 14 days if partially disabled (that is, single train flow). The procedure that controls this function is a Class A procedure, which Maine Yankee uses for safety related equipment.
The staff considers the equipment adequate for credit in the FSAR Chapter 14 However, the blowdown system isolation valves are not safety analysis.
Subsequent to the currently in the plant technical specifications (TS).
inspection, the staff reviewed the Maine Yankee TS and noted that valves credited for similar functions in the FSAR Chapter 14 Safety Analysis were included in the plant TS. The staff has therefore determined that the blowdown system isolation valve closure function should be added to the TS and concurs with the licensee's decision to add these valves to the plant The licensee currently is preparing a request to Technical Specifications.
amend its operating license to accomplish this.
e 15 Description of Assessment of RELAP5YA Applications Supporting Maine 6.2 Yankee Cycle 15 Operation On February 2, 1996, Maine Yankee submitted its schedule for producing the remaining information required by Section IX of the order.
Specifically, Maine Yankee submitted its schedule for providing an SBLOCA analysis that is specific to Maine Yankee for operation at power levels to 2700 MWt, and an integrated analysis of the containment.
(The SBLOCA analysis is to be submitted no later than M=y ), 1996, and the integrate <f analysis of containment is to be submitted no later than October 1, 1996.)
6.2.1 Operability Determinations for Other RELAPSYA Applications As noted above, Maine Yankee committed to conduct a review to identify other appitcations of RELAP5Ya to be used in cycle 15 and verify that operability,
.. cal Specifications, of affected systems and components as defined in its Te' is maintained.
The Itcensee initially identified 16 applications of RELnP5YA to be used in supporting cycle 15 operation. The licensee found that six applications are These applications are:
steam generator blowdown not applicable to cycle 15.
tank analysis, simulator benchmarking, turbine building environmental qualification, emergency drill support (modeling an SBLOCA), reactor pressure vessel temperature for low temperature overpressurization considerations, and the use of ZIRLO fuel cladding. The use of RELAP5YA in five of these six cases was either in support of a cycle or application previous to cycle 15, or to confirm or verify another calculation.
In the case of ZIRLO fuel cladding, the Commission issued Amendment No. 155 to Maine Yankee's facility Operating 29, 1996, in response to Maine Yankee's application dated Lice se on februaryThe licensee's cycle 15 core contains no ZIRLO clad fuel, Augtst 30, 1995.
but future cores may, consistent with Amendment No. 155 and the terms and The inspectors reviewed the conditions of its Operating License.
documentation of two more RELAP5YA applications that the licensee determined These applications were the simulator were not appitcable to cycle 15.
benchmark analysis and the turbine building environmental qualification analysis.
The inspectors found that the licensee used RETRAN-02/M0005 and RELAPS/M003 The events analyzed for the simulator benchmark for simulator benchmarking.
are consistent with that specified in ANSI /ANS-3.5, " Nuclear Power Plant RETRAN-02/ MOD 02 is Simulators for Use in Operator Training and Examination."
an NRC-approved code to analyze a steam line break (a non-LOCA event) in RETRAN-02/M0005 is an updated version and retains licensing applications.
RELAP5/M003 was mathematical schemes and physical models of RETRAN-02/ MOD 02.
RELAP5/M003 results have been developed for the analysis of LOCA events.
compared with test data, including the LOFT L2-5 test for prediction equilibrium condensation behaviors.
small-break LOCAs (0.01 and 0.15 f t' breaks in the hot leg) were run using However, these two events are RELAPSYA for additional simulator validation.The staff therefore conclud beyond the recommendations of ANSI /ANS-3.5.
16 that the licensee's simulator benchmarks have been performed to industry standards, without relying upon RELAP5YA, and the codes applied are appropriate for the events analyzed.
For the turbine building environmental qualification calculations, inspectors found that the licensee and its contractor used hand-calculated methods to The ru ulting mass determine mass and energy releases resulting from LOCAs.
and energy releases were used as input to RELAP4/"')D5 to calculate the The staff. judges this to be an temperature profile in the turbine building.
appropriate methodology for turbine building environmental qualification c.alcul ations.
The licensee performed formal, documented operability determinations for the All 10 remaining applications of RELAP5YA that support operating cycle 15.
10 operability determinations were performed in accordance with station procedure 1-200-2, Operability Determination, " Ensuring the functional Each of these operability Capability of a System or Component," Revision 0.
determinations was properly documented on Attachment A (sheets 1 and 2) to the procedure, with supporting documentation attached.
The inspectors performed a detailed review of four of these operability determinations as a sample of the adequacy of the 10 remaining operability containment determinations.
These operability determinations were:
environmental qualification, reactor coolant pump trip study, spurious opening of a power-operated reifef valve (in support of the 10 CFR 50 Appendix R fire protection program), and the containment.
(Note that the operability determination performed for the reactor containment building i, not strictly a The licensee performed this operability determination because RELAP5YA issue.
of the December 4, 1995, allegation that the 55 psig containment design pressure could be exceeded during a postulated loss-of-coolant accident if the The containment was at its maximum allowed initial pressure of 3.0 psig.)
licensee's use of RELAP5YA is limited to non-design-basis events for calculation of mass and energy releases to address issues such as these.
These analyses are less complex than licensing basis LOCA calculations and do They are not used to meet the 10 CFR 50.46 not involve PCT calculations.
The staff's previous assessment of RELAP5YA is sufficient to Furthermore, the results of requirements.
support the limited applications of RELAPSYA.
mass and energy calculations inspected do not show unstable and divergent mathematical solutions, such as the oscillations observed in the PCT calculations identified in the NRC's order of January 3,1996.
The inspector found these operability determinations properly documented, supported by analysis, references, and attachments, and satisfactorily Each was signed by the originator and two concurring reviewers.
completed.
Based on the results of our inspection and assessments, the team finds that the licensee has used adequate methods in the analysis to support Cycle 15 The results of the analysis have shown that (1) core operation at 2440 MWt. operating Itmits (COLs), established at 2700 MWt Maine Yankee and not relying on RELAP5YA, bound the operating conditions at 2440 MWt, up to 4000 mwd /Mt cycle exposure, and (2) the procedures and plant modifications, which originally relied on RELAP5YA, remain valid for the
17 Therefore, the team concludes that Conuitions 1 and t
current operating cycle.
2 imposed in the January 3, 1996 NRC Order are adequately addressed for operation at 2440 MWt, up to 4000 mwd /Mt cycle exposure.
6.3 Power limitation to 2440 MWt The inspectors reviewed the actions taken by the licensee to comply with The actions involved procedure limiting power operation to 2440 MWt (90.3%).
changes and personnel briefings but did not involve any physical hardware Procedure 1-4, " Operations at Power," and procedure 1-4-2, " Power Level Control" were revised via temporary procedure change (TPC) to eliminate changes.
In addition, precautions were added to not exceed 2440 full power operation.Copies of the order limiting reactor power were MWL reactor power.
distributed to all holders of controlled copies of technical specifications with a cover letter directing that the order be placed imediately following The Reactor Engineering instructions and the facility operating license.
Daily Plant Status report were revised to show a limit of 2440 MWt for reactor power.
The Assistant Manager of Operations met with each operating crew to explain On January 8, 1996, the order and the measures taken to limit reactor power.
the inspector attended a session during which operators were issues.
All licensed operators were required to be briefed was very comprehensive.
A summary of the actions prior to standing a watch with the reactor critical.
Earlier on the taken was placed in the Operating Crew Document Review Book.
evening of January 14, it was noted that one of the licensed operators had not been briefed on the confirmatory order and actions taken to limit reactorTh power and containment pressure.
that shift until he had been provided the required briefing.
The inspectors found Maine Yankee's actions to ensure power limitations to The inspectors independent review of procedures 2440 MWt to be thorough.
The inspectors also reviewed the technical revealed no discrepancies.
specifications and portions of the final safety analysis report to determine if any other changes or physical modifications would be required to ensure None were identified.
The full compliance with the power limitation.
inspectors particularly verified that no setpoint change to any of the reactor protection system signals was required.
6.4 Containment internal operating pi sure limitation to 2 psig f
The inspectors also reviewed the licensee's actions taken to limit the containment operating pressure to 2 psig or less. The changes involved J
procedure revisions, computer alarm setpoint modifications, instrument re-
)
calibration and personnel briefings.
Procedures 1-2, " Reactor Startup," 3-1-1, " Instrument Surveillance," and 1 2, " Containment Leak Monitoring" were revised via temporary procedure change (TPC) to reflect a containment pressu.e limit of 2 psig rather than 3 psig.
Abnormal operating procedures 2-37R, "PanAlarm Response," 2-10, " Loss of Pressure Control /RCS Leak," and 2-30, " Loss of Containment Integrity" were
l 18 I
2 psig.
revised via TPC to change the 3 psig limit on containment pressure to d
Main control board annunciator " Containment Pressure Hi" setpoint from 3 psig to 2 psig. Computer alarm points relating to containment Operating crew briefings were conducted as noted were reset to lower values. Copies of the order were sent to all hold A sumary of the the facility technical specifications as noted above. actions above.
d above.
An interpretation was added to the Technical Specification !cterprotatio Manuel to reflect the fact that the limit on containment pressure during i
routine operation is 2 psig, not the 3 psig listed in the technical specifications.
The inspectors were satisfied that Maine Yankee had taken appropr less to ensure that the containment cperating oressure would be limited to than 2 psig.
7.0 ADil!NISTRATIVE 7.1 Perscns Contacted du d intery ews and d During this report period, inspector
,{n ena with various licensee personnel, J
technicians and the licensee management.
Sumary of Facility Activities 7.2 the inspectors briefed the Maine State legislatures' Joint Comittee on Utilities and Energy on the scope and status of the 25, 1996, On January investigation of the issues involving the emergency core cooling sy
}
the containment.
During the inspection period the inspectors conducted backshift ins 2, 3, 4, 5, 8, 9, 17, 25 and 26, and February 5, 6 and 8, and deep backshift inspection on January 7,17, 21 and 26.1996, the in January activities.
Interface with the State of Maine 7.3 Periodically, the resident inspectors and the onsite representative of t State of Maine discussed findings and activities of their corresponding organizations.
7.4 Exit Meeting Inspectors periodically held meetings with senior facility managemen At the conclusion of the discuss the inspection scope and findings. inspection, the ins f
period.
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