ML20195E896
| ML20195E896 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/09/1998 |
| From: | Meisner M Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-309-96-09, 50-309-96-10, 50-309-96-11, 50-309-96-16, 50-309-96-9, 50-309-97-01, 50-309-97-1, MJM-98-59, MN-98-71, NUDOCS 9811190146 | |
| Download: ML20195E896 (23) | |
Text
,4 MaineYankee P.O. BOX 408
- WISCASSET, MAINE 04578 a (207) 8824321 s
November 9,1998 MN-98-71 MJM-98-59 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555
Reference:
(a)
License No. DPR-36 (Docket No. 50-309)
( b) Notice of Violation (NRC Inspection Reports Nos. 50-309/96-09;96-10;96-11;96-16;97-01 and NRC Office ofInvestigations Reports Nos. 1-95-050,1-96-025 & 1-96-043)
(c)
Letter to NRC Dated February 28,1997, MN 97-39 Response to NRC Inspection Report No. 50-309/96-16 (d)
Letter to NRC Dated April 3,1998, MN 98-23 Response to Apparent violations stemming from NRC OIR Nos. 1-96-25,1-95-50,1-96-043 (c)
Letter: M. B. Sellman to USNRC; Certification of Permanent Cessation of Power Operation and Pemianent Removal of Fuel From the Reactor; MN 89, dated August 7,1997.
Gentlemen:
This letter provides Maine Yankee Atomic Power Company's response to the Notice of Violations dated October 8,1998, Reference (b).
Maine Yankee is gratified that the NRC concluded that Maine Yankee never intentionally sought to mislead the NRC, thus setting the record straight on the charges of wilfulness. It is unfortunate that the Office ofInvestigation (01) investigation prevented a discussion of the allegations with the NRC until the spring of this year (1998), over two years after the allegations were first received in December of 1995.
Ilj[1 The NOV also brings to a close a number of other violations that Maine Yankee agreed with in early 1997 following an Independent Safety Assessment (ISA) of the plant in mid-1996. The NRC found Maine Yankee was operated safely but needed significant upgrades, prompting a series of improvements and corrective actions at the plant.
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r MaineYankee UNITED STATES NUCLEAR REGULATORY COMMISSION MN-98-71 Attention: Document Control Desk Page Two Maine Yankee has specifically responded to the majority of the ISA derived NOV's previously in Reference (c). Responses to two new violations are included in this letter. In addition, an updated response to the previously identified (apparent) violations is included, taking into account Maine Yankee's decommissioning status.
Maine Yankee had previously responded to a number of OI derived apparent violations in reference (d). Maine Yankee notes that as.a result of that response, and the result of a pre-decisional l
conference held during April,1998, a number of apparent violations have been withdrawn, and a i
number have been more accurately portrayed. A response to the cited violations is included as Attachment B to this letter.
Ve ruly yo Mi ael
. Meisner, President Ma' ankee Atomic Power Company c: Mr. Hubert Miller Mr. Michael T. Masnik Mr. Michael K. Webb Mr. Mark Roberts Mr. Patrick J. Dostic Mr. Uldis Vanags STATE OF MAINE Then personally appeared before me, Michael J. Meisner, who being duly sworn did state that he is the President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf ofMaine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief.
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l ATTACilMENT A RESPONSE TO NOTICE OF VIOLATION (NOTICE 1)
ISA DERIVED i
1.
VIOLATIONS RELATED TO INADEQUATE TESTING A.
Technical Specification (TS) 3.9.B, " Engineered Safeguards Features Actuation System,"
Table 3.9-2 No.1, " Safety injection," requires, in part, a minimum of 3 operable channels for both high containment pressure and low pressurizer pressure per safety injection actuation system (SIAS) subsystem to be operable whenever automatic initiation of Engineered Safeguards Feature (ESP) systems is required to be operable. TS 3.6.C requires, in part, two operable and redundant emergency core cooling system (ECCS) trains including one in each high pressure safety injection (HPSI) pump subsystem, an ESF system, to be operable whenever the reactor is in a power operation condition.
Contrary to the above, during periods of power operation from December 1991 until August 17,1996, there were no operable channels of high containment pressure or low pressurizer pressure in the 'A' subsystem of the SIAS. Specifically, the 'A' HPSI pump would not have automatically started in response to a SIAS signal (high containment i
pressure or low pressurizer pressure) due to a missing wire in the HPSI pump circuit.
(01013)
B.
TS 4.0, " Surveillance Requirements," requires that each surveillance requirement in Section 4 be performed within the specified surveillance interval.
1.
TS 4.1, " Instrumentation and Control," requires, in part, that testing of engineered safeguards system logic channels be performed as specified in Table 4.1-2. TS Table 4.1.2, requires, in part, that Channel 3, SIAS actuation relays; Channel 10, refueling water tank level recirculation actuation signal (RAS) initiation; Channel 20, feedwater trip system; and Channel 21, emergency feedwater (EFW) initiation, be tested at least once every 18 months.
Contrary to the above, prior to August 18,1996, surveillance tests required by TS 4.1, Table 4.1-2, were not performed at least once every 18 months. Specifically:
a.
Channel 3 - HPSI pump start signals for SIAS and undervoltage (UV) conditions were not tested independently; and the dual function swing pump (P-61S) was not tested as a low pressure safety injection (LPSI) and containment spray pump for UV and SIAS actuation; i
b.
Channel 10 - Manual initiation of RAS was not tested; and the automatic trip of swing pump (P-61S), when used as a LPSI pump, was not tested; I
Channel 20 - The SIAS permissive was not adequately tested in that the main c.
l feedwater pump, condensate pump, and heater drain pump trip systems were not tested with a SIAS coincident with a steam generator low pressure signal; and
d.
Channel 21 - Emergency feed water pump circuit breaker closure was not tested. (01023) 2.
TS 4.5, " Emergency Power System Periodic Testing," A.2, " Diesel Generators,"
requires, in part, testing of the diesel generators (DGs) during each refueling interval that demonstrates their readiness to start automatically and restore power to vital equipment on loss of all normal a-c station service power supplies.
Contrary to the above, during each refueling interval prior to August 18,1996, tests required by TS 4.5.A.2 were not being performed in that emergency bus loading and load shedding, necessary to demonstrate the DGs readiness to start automatically and restore power to vital equipment on loss of all normal a-c station service power supplies, was not adequately tested. Specifically, for the following vital equipment:
Service water (SW) pumps P-29B and P-29C were not verified to remain a.
operating on the bus if they were the only available pumps in the train, b.
Primary component cooling (PCC) pump P-9B was not tested as the preferred pump.
c.
Secondary component cooling (SCC) pump P-10B was not tested as the preferred pump. (01033) 3.
TS 4.6, " Periodic Testing," D.1.a, "Feedwater Trip System, Main Feedwater Pumps," requires that each main feedwater pump, condensate pump, and heater drain pump trip system shall be tested during each refueling interval by tripping the actuation circuitry with a safety injection signal coincident with a steam generator low pressure signal.
Contrary to the above, during each refueling interval prior to August 18,1996, the testing required by TS 4.6.D.l.a was not perfomied to verify tripping of each main feedwater pump, condensate pump and heater drain pump circuit breaker with a safety injection signal coincident with a steam generator low pressure signal.
(01043)
C.
TS 4.7.A, " Inservice Inspection and Testing of Safety Class Components," requires, in part, the establishment of an " Inservice Inspection Program" that meets the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, " Inservice Testing of Pumps and Valves," for safety class 3 pressure retaining components.
10 CFR 50.55a(f), " Inservice testing requiremems," requires, in part, that safety related valves must meet the requirements applicable to components which are classified as ASME Code Class 3 set forth in section XI of the ASME Boiler and Pressure Vessel Code.
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ASME Code,Section XI, IWV-3520, " Check Valve Tests," requires that valves nonnally open during plant operation whose function is to prevent reversed flow, shall be tested in i
a manner that proves that the disk travels to the seat promptly on cessation or reversal of
- flow, Contrary to the above, as of August 18,1996, inservice testing for 15 safety class 3 pressure retaining check valves that were located at the discharge of safety related pumps did not meet the requirements of the ASME Code,Section XI. This inservice testing failed to demonstrate that the standby pump's discharge check valves, which are normally open during operation and whose function is to prevent reversed flow, would properly close on the cessation or reversal of flow which would be necessary to prevent short-cycling of the operating pump. Specifically, the following safety class 3 valves were not adequately tested:
- 1. Charging /HPSI pump discharge check valves CH-10,19 and 26;
- 2. EFW pump discharge check valves EFW-15, and 314;
- 3. LPSI pump discharge check valves LPSI-50 and 51;
- 4. PCC pump discharge check valves PCC-6 and 13;
- 5. SCC pump discharge check valves SCC-7 and 14; and
- 6. SW pump discharge check valves SW-1,4,7 and 10. (01053)
These violations in Section I represent a Severity Level III problem (Supplement I).
51 AINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (c). In Reference (e) Maine Yankee informed the USNRC that the Board of Directors of Maine Yankee had decided to permanently cease operations at the Maine Yankee Plant and that fuel had been permanently removed from the reactor. In accordance with 10CFR50.82(a)(2) the certifications in the letter modified the Maine Yankee license to permanently withdraw Maine Yankee's authority to operate the reactor. As a result of changes made to the plant during decommissioning, none of the equipment cited in this NOV is in the current design / licensing basis, therefore, no specific corrective actions needed. Corrective actions related to programmatic issues identified in Reference c) with respect to adequate testing of equipment have been implemented as appropriate. Maine Yankee is in compliance with current Technical Specification requirements applicable to testing of equipment.
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l II.
VIOLATIONS RELATED TO ENVIRONMENTAL QUALIFICATION 10 CFR 50.49(d) requires, in part, that the licensee shall include in a qualification file the environmental conditions, including temperature, humidity, and submergence, at the location where electrical equipment important to safety covered by 10 CFR 50.49 must perform.
10 CFR 50.49(j) requires that a record of the environmental qualification must be maintained in an auditable form to permit verification that each item of electric equipment important to safety is qualified for its application and meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function.
10 CFR 50.49(f) requires each item of electric equipment important to safety to be environmentally qualified by (1) testing ofidentical or similar equipment under identical or similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable, (2) experience with identical or similar equipment under similar conditions with a supporting analysis, or (3) analysis in combination with partial type-test data that supports the analytical assumptions and conclusions.
10 CFR 50.49(b) defines electric equipment important to safety within the scope of 10 CFR 50.49 as safety-related electric equipment, non-safety-related electric equipment whose failure under postulated accident conditions could prevent safety related equipment from accomplishing the functions identified in 10 CFR 50.49(b)(1), and certain post-accident monitoring equipment.
10 CFR 50.49(e) specifies the conditions and other location dependent considerations that the electric equipment qualification program must be based upon. These conditions and considerations include, in part, temperature and pressure, humidity, and submergence, as applicable, during and afler the most severe accident environment for which electrical equipment important to safety must remain functional.
A.
Contrary to the above, as of August 2,1996, the qualification files for 30 items of electric equipment important to safety inside the reactor containment did not permit verification that the items were qualified for their applications and met their specified performance requirements when subjected to submergence, a condition predicted to be present when they must perform their safety ftmetions afier a loss of coolant accident (LOCA). The qualification files for these 30 items of electric equipment did not include the correct submergence level at the location where they must meet their specified performance requirements. Specifically, safety-related valve limit switches and associated pigtails, Rosemount transmitters and associated electrical connectors, and certain Rockbestos l
cables were not qualified for post-LOCA submergence in that there were no documents in Maine Yankee's environmental qualification (EQ) file to demonstrate qualification of the items by testing or a combination of testing, experience, or partial type-test data with analysis. (02013)
B.
Contrary to the above, as of March 11,1997, the qualification files for two PCC pump motors and two SCC pump motors, safety related components, did not pemlit verification
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that they were environmentally qualified to remain functional during and following a high energy line break (HELB) in the turbine building, which is the most severe design basis event at their location during or after which they must remain functional. Specifically, j
- there were no documents in the Maine Yankee EQ file to demonstrate that the PCC and SCC pump motors were qualified for high temperature and high humidity resulting from l
a HELB. (02023)
These violations in Section 11 represent a Severity Level III problem (Supplement I).
g MAINE YANKEE RESPONSE Maine Yankee agrees with these violations.
. Reply to this Notice of Violation (NOV) was submitted in Reference (c). In Reference (e) Maine Yankee infomied the USNRC that the Board of Directors of Maine Yankee had decided to permanently cease operations at the Maine Yankee Plant and that fuel had been permanently
.l removed from the reactor. In accordance with 10CFR50.82(a)(2) the certificationc in the letter modified the Maine Yankee license to permanently withdraw Maine Yankee's authority to operate the reactor. As a result of changes made to the plant during decommissioning, none of the equipment cited in this NOV is in the current design / licensing basis, therefore, no corrective actions are required. Also as a result of the certifications in Reference (c), Maine Yankee is no longer subject L
to the requirements of 10CFR50.49 (" Environmental qualification of electric equipment impoitant to' safety for nuclear power plants").
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III.
VIOLATIONS RELATED TO INADEQUATE SAFETY REVIEW A.
10 CFR 50.59, " Changes, tests and experiments," permits the licensee, in part, to make changes in the facility and procedures as described in the safety analysis report without prior Commission approval provided the ch.nge does not involve an unreviewed safety l
question (USQ). A proposed change shall be deemed to involve a USQ, in part, if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created. The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve an USQ.
1.
Contrary to the above, in May 1992, Maine Yankee made a change to procedures i
as described in the FSAR that involved an USQ without prior Commission approval f
'due to an inadequate safety evaluation. Specifically, Maine Yankee established procedure 1-22-2, "AC and DC Vital Bus Operation," which allowed cross l~
connecting redundant 125 Vdc vital buses for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during plant operation.
l This was a change from FSAR Appendix A, Criterion 39, " Emergency Power for ESFs," which provides, in part, that the alternate power systems be provided and designed with adequate independence and redundancy to permit the functioning required of the ESFs and, as a minimum, that the onsite power system shall independently provide required capacity assuming a single failure. With the 125 Vdc buses cross connected, all 125 Vdc power to the ESFs could have been lost due to a single failure. This created the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report and represents an USQ. As of August 30,1996, the safety evaluation performed for this procedure change was inadequate in that it failed to identify this USQ. (03013) 2.
Contrary to the above, Maine Yankee made the following changes to the facility as described in the FSAR without performing a written safety evaluation for these changes to provide the basis for the determination that the changes did not involve a USQ, each of which constitutes an individual violation:
In January 1996, Maine Yankee restricted the maximum SW operating a.
temperatures to 70.2 oF for component cooling wter(CCW) heat exchangers E-4B and E-5A, and 78.5 oF for CCW heat exchangers E-4A and E-5B to support desiga basis post-LOCA condition heat removal capability. This was a change from FSAR Section 9.4.1 which assumed SW inlet temperatures of 80 oF for E-4B and E-5A, and 90 oF for E-4A and E-5B. As of August 30, L
1996, no safety evaluation had been performed for the change in SW operating temperatures. (03023) b.
On February 21, 1997, Maine Yankee changed the layout within the i
protected area by installing and filling a 1000 gallon propane tank contrary to FSAR, Section 1.3," Plant Description Summary." This addition had the potential to damage the circulating water (CW) pumphouse ifit exploded, l-and could negatively affect both trains of the SW system since the SW pumps are located in the CW pumphouse. As ofMarch 5,1997, no safety evaluation L
had been performed for the propane tank. (03033)
c.
On March 11,1997, a drain hose was temporarily installed on a spent fuel pool pump suction pipe which was contrary to the configuration of the spent fuel pool cooling system as shown in plant drawings and the FSAR, Section 9.8, " Fuel Pool Cooling System." As of March 15,1997, no safety evaluation had been performed for this change in the configuration of the spent fuel pool cooling system. (03043) d.
As of August 30, 1996, no safety evaluation had been performed for approximately 89 equipment and procedure changes that were made to-equipment.and procedures described in the FSAR. These changes were identified by Maine Yankee as a result of an initiative to upgrade the FSAR and are listed in the " Final Safety Analysis Report (Revision 13) Maine Yankee FSAR Update (MFU) Status Report." (03053) j B.
10 CFR 50.71(e) requires the licensee to update the FSAR to assure that the information
]
included in the FSAR contains the latest material developed. Updates must be filed annually or 6 months after each refueling outage. The updates must reflect all changes made in the facility or procedures as described in the FSAR up to a maximum of 6 months prior to the date of filing.
Contrary to the above, as of August 1996, the FS AR was not updated to reflect 27 changes made to the facility as a result of Engineering Design Change Requests and Plant Design Change Requests that were implemented between 1980 and August 1996. These changes were identified by Maine Yankee as a result of an initiative to upgrade the FSAR and are listed in the " Final Safety Analysis Report (Revision 13) Maine Yankee FSAR Update (MFU) Status Report." (03063)
These violations in Section 111 represent a Severity Level Ill problem (Supplement 1).
MAINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (c).
Correction actions taken to date have proven very effective in improving Maine Yankee's 10CFR50.59 process. This includes training of all 50.59 preparers and reviews, and explicit utilization of NEl guidance document 96-07 for conducting 50.59 evaluations, and process improvements to identify when 50.59 evaluations must be performed.
In addition, the FSAR update / review process (which was started prior to the ISA) has continued. At this time, the Maine Yankee SAR has been substantially rewritten to reflect the decommissioning status of the plant in accordance with 50.71(e). Maine Yankee is in compliance with 10CFR50.59 and 10CFR50.71(e).
4 IV.
VIOLATIONS ASSOCIATED WITIIINADEQUATE CORRECTIVE ACTIONS 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.
A.
Contrary to the above, from October 31,1995, until August 16,1996, the inability of train
'A' of the control (CR) breathing air system to maintain a positive pressure in the control room during accident conditions was not corrected. Specifically, during testing of the 'A' train of the CR breathing air system on Octobera31,s1995, in accordance with Surveillance Procedure 3.17.5, pressure in the CR was slightly negative. These test results indicated that the 'A' train of control room ventilation system (CRVS) was not operable, a significant condition adverse to quality. Maine Yankee did not take measures to assure that the cause of this condition was determined and did not take corrective actions to l
preclude repetition. No action was taken to restore operability of the 'A' train of CRVS prior to making the reactor critical on January 11, 1996 contrary to Technical Specifications.(04013)
B.
Contrary to the above, as of August 3,1996, a significant condition adverse to quality identified in 1991 had not been corrected. Specifically, a loss of non safety-related instrument air could cause the air operated dampers (VP-A-56 and VP-A-57) in the containment spray building (CSB) fans' ducts to fail shut, rendering the fans (FN 44A and 448) incapable of perfonning their safety function of providing ventilation to the low pressure safety injection (LPSI) and containment spray pumps and heat exchangers area (i.e., by reming more than 10,000 cfm of air as specified in the Maine Yankee FSAR, l
Section 9.o.2.3) in the CSB. Without adequate ventilation, the LPSI and containment spray pump motors could fail due to overheating. This potential to lose CSB safety-related fans was identified during a ventilation system review by engineering in 1991 and was not corrected until August 3,1996. (04023)
C.
Contrary to the above, between 1994 and 1996, actions to determine the cause and preclude repetition of icing and clogging of the CSB heating, ventilating, and air conditioning (HVAC) unit, HV-7, a significant condition adverse to quality, were inadequate. Specifically, the clogging occurred at least three times during that period, and even though corrective actions were taken, they were not effective in precluding repetition of the adverse condition. The clogging of the HVAC unit caused the CSB ventilation l
system (a support system for the LPSI and containment spray systems) to be inoperable, thereby potentially rendering both trains of LPSI and containment spray systems inoperable. (04033)
D.
Contrary to the above, as of August 30,1996, actions to determine the cause and preclude repetition of Auxiliary Feedwater (AFW) control system failures, a significant condition adverse to quality, were inadequate. Specifically, repetitive problems between 1992 and l
1996 resulted in degraded reliability for the AFW pump to respond to a start /run demand.
l Even though corrective actions were taken, they did not preclude repetition of the control l
system problems. (04043)
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E.
Contrary to the above, as of April 1996, a design deficiency, which was a condition adverse to quality, involving the plant being outside ofits design basis for a turbine hall flood, had not been promptly corrected. Specifically, during the Service Water System Operational Performance Inspection in 1994, Maine Yankee identified that the plant was outside of the design basis for a turbine hall flood in that during a design basis flood in the turbine building, safety-related equipment in the control room, the DG room, and the
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turbine building would be rendered inoperable. (04053)
F.
Contrary to the above, from December 20,1996 until February 21,1997, Maine Yankee did not promptly establish compensatory corrective actions regarding an identified condition adverse to quality that would challenge the operability of the SW system.
Specifically, in a ventilation system assessment report, dated December 20,1996, Maine Yankee identified that a loss of ventilation in the circulating water pumphouse during periods of extreme cold temperatures, could create potentially freezing conditions for SW system components. Frozen water in stagnant lines could restrict flow to the SW pump bearings and gland cooling or create the potential for a line break. Compensatory actions to prevent freezing in the circulating water pump house were not taken until February 21, 1997.(04063)
These violations in Section IV represent a Severity Level III problem (Supplement I).
M AINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (c).
In Reference (e) Maine Yankee informed the USNRC that the Board of Directors of Maine Yankee had decided to permanently cease operations at the Maine Yankee Plant and that fuel had been pennanently removed from the reactor. In accordance with 10CFR50.83(a)(2) the certifications in the letter modified the Maine Yankee license to permanently withdraw Maine Yankee's authority to operate the reactor. As a result of changes made to the plant during decommissioning, none of the equipment cited in this NOV is in the current design / licensing basis, therefore, no equipment specific corrective actions are needed. Maine Yankee has subsequently completely revamped our Corrective Action system to more appropriately address our decommissioning status. This new program has been effective. Maine Yankee is in compliance with 10CFR Part 50, Appendix B, Criterion XVI," Corrective Action".
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V.
SEVERITY LEVEL IV VIOLATIONS TS (TS) 5.8.2.a requires, in part, that written procedures, as recommended in Appendix A of Regulatory Guide 1.33, (Rev. 2), February 1978, shall be established and implemented.
A.
Regulatory Guide 1.33, Appendix A, section 1, " Administrative Procedures," states, in part, that the maintenance of minimum shift complement; log entries; and authorities and responsibilities for safe operation and shutdown should be covered by written procedures.
1.
Contrary to the above, as of August 30,1996, Maine Yankee had not established procedural requirements, such that, in the event of a fire coincident with a medical emergency, the minimum control room staffing required by TS Section 5.2.2frable 5.2-1, would be satisfied. Specifically, only two Senior Reactor Operators (SROs) were required to be on duty. As a result, there would be no SRO in the control room, as required,if the two SROs on duty had to respond to a fire and a medical emergency concurrently. (05014)
This is a Severity Level IV violation (Supplement 1).
2.
Maine Yankee administrative procedure No. 1-200-10, " Conduct of Operations",
section 4.13, " Operability Assessment," specifies that if there is not a reasonable expectation that the equipment is operable, then the equipment shall be declared inoperable. Section 4.13 also specifies that an operability detemiination must assess the ability of the equipment to perform its intended safety action in the accident environment it would be subjected to when it would be called upon to do so and that tests or partial tests should be used for completing operability assessments.
Contrary to the above, on August 17,1996, administrative procedure No. 1-200-10 was not implemented in that the Operations Manager issued a memorandum that stated that TS testing discrepancies did not render the HPSI and containment spray swing pumps inoperable. This was contrary to the requirements of procedure 1-200-10 in that without performance of the testing that verifics that the pumps would perfomi their intended safety action when called upon, there was no reasonable assurance that the pumps were operable. (06014)
This is a Severity Level IV violation (Supplement 1).
B.
Regulatory Guide 1.33, Appendix A, section 9," Procedures for Performing Maintenance,"
states, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures or documented instructions; that preventive maintenance schedules should be developed to specify inspection or replacement of parts that have a specific lifetime; and that general procedures for the control of maintenance should include the method for obtaining permission and clearance for work.
1.
Maine Yankee maintenance procedure 5-9-3, " Maintenance of Emergency and Auxiliary Feedwater Pumps," Rev. 4, section 6.3.11 specifies the inspection of parts to detemiine if they are suitable for reuse. Maintenance procedure 5-9-3, section 6.3.12 and preventive maintenance (PM) card, M-18-3X-J, "P-25A Emergency Feedwater (EFW) Pump and Motor," specify performance of a liquid penetrant or magnetic particle examination of the cast iron diffuser assembly.
Contrary to the above, during the 1995 overhaul of the EFW pump P-25A, maintenance procedure 5-9-3 and PM card M-18-3X-J were not implemented in that no liquid penetrant or magnetic particle examinations were performed prior to reuse of the cast iron diffuser assembly. (07014)
This is a Severity Level IV violation (Supplement 1).
2.
Maine Yankee maintenance procedure 0-16-3, " Work Order Process," Rev.10, Attachment A, section I.A specifies that work performed on safety class equipment must be perfonned in accordance with procedure that provide specific information for the intended actions.
Contrary to the above, as of August 7,1996 Maine Yankee failed to establish procedures that provided specific instructions to reinstall fastener lock wire as intended and, as a result, lock wire was not reinstalled aller maintenance was perfomied on the following safety class equipment: Reactor coolant system loop No. 3 stop valve's motor operated valve actuator mounting fasteners and In-core instrumentation seal housings F-11, V-11, N-17, D-11, and T-16. (08014)
This is a Severity Level IV violation (Supplement 1).
3.
Maintenance procedure 0-16-3, sections 6.5 and 6.6 specify that, if necessary, equipment shall be tagged out prior to commencing work and that maintenance govemed by this procedure shall not commence until the Work Order has received all required reviews and approvals. Work Order No. 94-02278-01 for replacement of a pipe support specified that a white tagging order was required for SW pump P-29C to be out of service.
Contrary to the above, on August 13,1996, procedure 0-16-3 was not implemented in that maintenance personnel removed a seismically qualified pipe support on a seal water line for SW pump P-29C without a white tagging order being issued to tag the pump out of service. Removal of the existing pipe support caused the pump to be inoperable and; therefore, out of service. (09014)
This is a Severity Level IV violation (Supplement 1).
M AINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (d). Maine Yankee is in compliance with Technical Specification 5.5. (Procedures), which is the current reference to the cited Technical Specification (5.8.2.a)
ATTACIIMENT H RESPONSE TO NOTICE OF VIOLATION (NOTICE 2)
OI DERIVED VIOLATIONS I. PRINCIPAL PROBLEM RELATED TO INADEQUATE SMALL-BREAK-LOSS-OF COOLANT ANALYSES A.
VIOLATION RELATING TO INABILITY TO ANALYZE ENTIRE BREAK SPECTRUM FOR CYCLE 14.
10 C.F.R.
- 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
10 C.F.R. Part 50, Appendix K, Section 11.4, requires that to the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimentalinformation.
Contrary to the above, from October 14,1993, through January 25,1995 (during Cycle 14 operations), and in the Cycle 14 Core Performance Analysis Report (CPAR) submitted August 25,1993, Maine Yankee Atomic Power Company (MYAPCo) used unacceptable models to calculate ECCS performance and failed to calculate a number of postulated loss-of-coolant accidents of different sizes, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents were calculated. Specifically, there was a portion of the small-break spectrum between.35 ft2 and at least.6 R2 for which no acceptable evaluation model was capable of calculating cooling performance or reliably calculating cooling performance. MYAPCo calculated Small-Break Loss-of-Coolant Accident (SBLOCA) ECCS performance with the code described in "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2 3," dated October 1982 (RELAP5YA) and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868,
" Maine Yankee Small Break LOCA Analysis" (both of which were described as an Appendix K approach to RELAP5YA). MYAPCo calculated SBLOCA ECCS performance only up to the.35 R2 break size because the RELAP5YA SBLOCA evaluation model documented in YAEC-1868 was incapable of calculating ECCS performance for break sizes of and greater than 0.35 R2 as a result of the model's terminating aRer the safety injection tank actuation due to numerical convergence errors for the break size of.35 R2. MYAPCo calculated Large-Break Loss-of-Coolant (LBLOCA) ECCS Performance with the LBLOCA analysis described in YAEC-1160,
" Application of Yankee WREM-Based Generic PWR ECCS Evaluation Model to Maine Yankee", dated July 1978 (WREM). Although the WREM LBLOCA evaluation model was subsequently demonstrated in 1996 to be capable of calculating ECCS performance down to the,6R2 break size, the WREM LBLOCA evaluation model was not used to calculate ECCS performance in the small-break region for Cycle 14, and would not have
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been acceptable to calculate ECCS perfonnance for break sizes m the small-break region of 0.6 ft2 and above because the evaluation model was not compared to applicable experimental data to demonstrate its reliability in calculating ECCS performance in the small-break region. (01012)
B.
VIOLATION RELATING TO INABILITY TO ANALY7,E ENTIRE BREAK SPECTRUM FOR CYCLE 15 10 C.F.R.
- 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
10 C.F.R. Part 50, Appendix K, Section 11.4, requires that to the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimental information.
Contrary to the above, in the Cycle 15 Core Performance Analysis Report (CPAR) submitted December 1,1995, Maine Yankee Atomic Power Company (MYAPCo) used unacceptable models to calculate ECCS perfonnance and failed to calculate a number of postulated loss-of-coolant accidents of different sizes, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents were calculated. Specifically, there was a portion of the small-break spectrum between.35 R2 and at least.6 ft2 for which no acceptable evaluation model was capable of calculating cooling performance or reliably calculating cooling performance. MYAPCo calculated Small-Break Loss-of-Coolant Accident (SBLOCA) ECCS perfonnance with the code described in "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2 3," dated October 1982 (RELAP5YA) and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868,
" Maine Yankee Small Break LOCA Analysis" (both of which were described as an Appendix K approach to RELAP5YA). MYAPCo calculated SBLOCA ECCS performance only up to the.35 R2 break size because the RELAP5YA SBLOCA evaluation model documented in YAEC-1868 was incapable of calculating ECCS perfomiance for break sizes of and greater than 0.35 R2 as a result of the model's terminating after the safety injection tank actuation due to numerical convergence errors for the break size of.35 R2. MYAPCo calculated Large-Break Loss-of-Coolant (LBLOCA) ECCS Perfonnance with the LBLOCA analysis described in YAEC-1160,
" Application of Yankee WREM-Based Generic PWR ECCS Evaluation Model to Maine Yankee", dated July 1978 (WREM). Although the WREM LBLOCA evaluation model was subsequently demonstrated in 1996 to be capable of calculating ECCS performance down to the.6R2 break size, the WREM LBLOCA evaluation model was not used to i
i calculate ECCS performance in the small--break region for Cycle 15, and would not have l
been acceptable to calculate ECCS performance for break sizes in the small-break region of 0.6 ft2 and above because the evaluation model was not compared to applicable experimental data to demonstrate its reliability in calculating ECCS performance in the small-break region. (01022)
These violations in Section I represent a Severity Level 11 problem (Supplement 1).
MAINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (d). In Reference (e) Maine Yankee informed the USNRC that the Board of Directors of Maine Yankee had decided to permanently cease operations at the Maine Yankee Plant and that fuel had been permanently removed from the reactor. In accordance with 10CFR50.83(a)(2) the certifications in the letter modified the Maine Yankee license to permanently withdraw Maine Yankee's authority to operate the reactor. As a result of changes made to the plant during decommissioning, none of the equipment cited in this NOV is in the current design / licensing basis, therefore, no specific corrective actions needed. Also as a result of the certifications in Reference (e), Maine Yankee is no longer subject to the requirements of 10CFR50.46 (" Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors").
As discussed in Reference d), Maine Yankee has conducted a very thorough investigation of this issue. We have concluded that there was no actual safety consequence, no equipment was made inoperable and there was no wilfulness involved. In their findings, the NRC has also concurred.
Maine Yr.nkee has further concluded that there were no potential safety consequences. The technical basis for determining that the limiting SBLOCA break had been identified and that LBLOCA continued to be the overall limiting event has been consistently corroborated both historically and contemporaneously as discussed in Reference d). As a result, Maine Yankee requests the Staff to reconsider issuing this violation as a level III violation.
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- 11. OTIIER VIOLATIONS RELATED TO INADEQUATE SM ALL-BREAK-LOSS-OF-COOLANT ANALYSES (OI Report No.1-95-50) l A.
VIOLATION RELATING TO OPERATING CYCLE 13 Technical Specification (TS) 5.14.2, " Core Operating Limits Report," for the Maine j
Yankee Atomic Power Station (MYAPS) requires, in part, that analytical methods used to determine operating limits shall be limited to those previously reviewed and approved by NRC, as listed by TS 3.10. TS.3.10 specifies a Small-Break Loss-of-Coolant (SBLOCA) analysis, "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2,3, dated October 1982" (RELAP5YA). TS.3.10. does not specify any SBLOCA analysis produced by Combustion Engineering Corporation (CE),
10 C.F.R. " 50.9(a) requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects.
Contrary to the above, between April 19, 1992 and July 7,1993 (during Cycle 13 operations), Maine Yankee Atomic Power Company did not determine operating limits for Cycle 13 operations using the RELAP5YA SBLOCA analysis required by TS 5.14.2.
In fact, a Combustion Engineering (CE) SBLOCA code was used to prepare the reload analysis, as stated in the Core Performance Analysis Report for Cycle 13 at Section 5.5.5.3. In addition, on April 7,1992, Maine Yankee Atomic Power Company (MYAPCo) provided to the Commission MYAPCo's Cycleal3 Core Operating Limits Report (COLR), which contained inaccurate information material to the NRC. The COLR stated that MYAPCo used analytical methods listed in TS 5.14 to determine operating limits. In fact, MYAPCo used a CE SBLOCA analysis, which was not listed in TSd5.14. The SBLOCA analysis listed by TS 5.14 is "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Themial-Hydraulic Analysis, Volumes 1,2,3, dated October 1982" (RELAP5YA). This inaccurate infom1ation was material to the NRC because it was a representation that RELAPSYA, which had been approved for application to MYAPS pursuant to the Three Mile Island Action Plan, Item II.K.3.30 (NUREG 0737),
had been used to establish core operating limits for Cycle 13 operations. (02014)
This is a Severity Level IV violation (Supplement 1)
MAINE YANKEE RESPONSE Maine Yankee does not agree with this violation.
As previously discussed in Reference d), Maine Yankee did set Core Operating Limits using analytical methods previously reviewed and approved by the NRC as reflected in Technical Specification TS 5.14.2.
Via oral communication following the April 1998 pre-decisional enforcement conference, Maine Yankee provided several examples of other PWR licensees which did not list their SBLOCA analytical methods when LBLOCA was limiting.
We understand, from the discussion at the April 1998 meeting, that it is the NRC's expectation that any analytical methods NOT used in setting Core Operating Limits, but which do provide a basis l
for using approved methods shall be referenced in the applicable Technical Specification section.
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l Our disagreement notwithstanding, should the NRC continue to maintain this position, we respectfully request that this guidance be made formally available to the industry.
4 As a result of the certifications in Reference (e), Maine Yankee is no longer subject to the requirements of 10CFR50.46 (" Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors").
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B.
VIOLATIONS RELATED TO IMPROPER APPLICATION OF ALB-CIIAMBRE CORRELATION FOR CYCLE 14 10 C.F.R.
- 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model.
Contrary to the above, from October 14,1993, through January 25,1995 (during Cycle 14 operations), and in the Cycle 14 Core Perfomiance Analysis Report (CPAR) submitted August 25, 1993, MYAPCo calculated ECCS performance for SBLOCAs with an unacceptable evaluation model. MYAPCo used the ECCS code described in YAEC-1300P, "RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1,2,3," dated October 1982 (RELAP5YA), and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868,
" Maine Yankee Small Break LOCA Analysis" (YAEC-1868). RELAP5YA as applied was not an acceptable evaluation model because the nodalization model of YAEC-1868 incorrectly applied the Alb-Chambre correlation, resulting in the unjustified use oflarge penetration factors and a large cross flow resistance factor in the split downcomer nodalization. (02034)
This is a Severity Level IV violation.
C.
VIOLATIONS RELATED TO IMPROPER APPLICATION OF ALB-CIIAMBRE CORRELATION FOR CYCLE 15 10 C.F.R. " 50.46(a)(1) requnes, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model.
Contrary to the above, in the Cycle 15 Core Performance Analysis Report (CPAR) submitted December 1,1995, MYAPCo calculated ECCS performance for SBLOCAs with an unacceptable evaluation model. MYAPCo used the ECCS code described in YAEC-1300P, "RELAP5YA: A Computer Program for Light Water Reactor System Thennal-Hydraulic Analysis, Volumes 1,2,3," dated October 1982 (RELAP5YA), and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868, 4
" Maine Yankee Small Break LOCA Analysis" (YAEC-1868). RELAPSYA as applied was not an acceptable evaluation model because the nodalization model of YAEC-1868
' incorrectly applied the Alb-Chambre correlation, resulting in the unjustified use oflarge penetration factors and a large cross flow resistance factor in the split downcomer nodalization. (02034)
This is a Severity Level IV violation.
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MAINE YANKEE RESPONSE
. Maine Yankee agrees with this violation.-
As discussed in Reference d), this issue resulted from a non-consequential error in the rpplication of the Alb-Chambre correlation.
As result of the certifications in Reference (e), Maine Yankee is no longer subject to the requirements of 10CFR50.46.-
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VIOLATION RELATING TO ANALYSIS OF REDUCED STEAM GENERATOR PRESSURE FOR CYCLE 14
)
10 C.F.R.
- 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable evaluation model.10 C.F.R.
- 50.46(a)(1)(ii) provides that an ECCS evaluation model may be developed in conformance with the required and acceptable features of Appendix K ECCS Evaluation Models.
Contrary to the above, in a January 1993 analysis of a decrease in steam generator pressure, performed pursuant to the requirements of 10 C.F.R.
- 50.59, MYAPCo used an unacceptable evaluation model to calculate SBLOCA ECCS performance. MYAPCo used a Best Estimate (BE) plant-specific evaluation model (described in an August 1,1990, report produced by Yankee Atomic Electric Company) to implement the SBLOCA code described in YAEC 1300P, "RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1, 2, 3,"
dated October 1982 (RELAP5YA). In January 1989, the NRC transmitted its Safety Evaluation Report approving RELAP5YA for application to Maine Yankee Atomic Power Station as an Appendix K model, not as a BE model. Furthermore, contrary to 10 C.F.R. Part 50, I
Appendix K, the BE evaluation model calculated decay heat with the 1979 ANS Standard rather than the 1971 ANS Standard plus 20 percent, and calculated the two-phase critical flow with the RELAP5YA mechanistic model rather than the Moody critical flow model.
(02044)
This is a Severity Level IV violation.(Supplement 1)
M AINE YANKEE RESPONSE Maine Yankee agrees with these violations.
Reply to this Notice of Violation (NOV) was submitted in Reference (d).
Correction actions taken to date have proven very effective in improving Maine Yankee's 10CFR50.59 process. This includes training of all 50.59 preparers and reviews, and explicit utilization of NEl guidance document 96-07 for conducting 50.59 evaluations, and process improvements to identify when 50.59 evaluations must be performed.
As a result of the certifications in Reference (e), Maine Yankee is no longer subject to the requirements of 10CFR50.46
(" Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors").
Ill.
VIOLATION ASSOCIATED WITII SAFETY SYSTEM LOGIC TESTING (OI REPORT NO. 1-96-043)
Technical Specification 5.8.2 states, in part, that written procedures be established, implemented, and maintained to control, among other things, activities conceming testing of safety related equipment.
Item 12 of Attachment C to Procedure No. 0-16-3, " Work Order Process," defines a Functional Test Instruction (FTI) as instructions that define the evolutions or operations necessary to prove functionality or operability of a component, system, or structure.
Precaution 3.1 of Work Order 96-02928-00, Attachment A, " Functional Test for P-14A/S on A Train SIAS and Bus 5 Undervoltage," and Work Order 96-02929-00, Attachment A,
" Functional Test for P-14 B/S on B Train SI AS and Bus 6 Undervoltage," states that if any step cannot be completed as specified in the FTI, then the Field Engineer must be contacted and any deviation from this FTI must be authorized in accordance with Procedure 0-16-3.
Deviations to FTIs are permitted through the use of Minor Technical Changes (.MTC) as described in Item 13 of Attachment C to Procedure No. 0-16-3.
10 C.F.R.
- 50.9(a) provides in part that information required by the Commission's regulations to be maintained by the licensee to be complete and accurate in all material respects.
10 C.F.R. Part 50, Appendix B, Criterion XVII, " Quality Assurance Records," requires, in part, that records of tests affecting quality be maintained.
Contrary to the above:
(1)
On August 22,1996, Step 5.3.3 of WO 96-02928-00 and WO 96-02929-00 could not be performed as written, and the licensee failed to resolve the discrepancy by making a Minor Technical Change. Specifically, Step 5.3.3 provided that at Main Control Board (MCB), Section C, open circuit continuity be verified at 86-RAS A-2(YAF) using a volt-ohm meter (VOM) across the 5-5C contacts. The field test engineers could not verify the open contacts with a VOM because of resistance in the circuit caused by a bulb and resistor _ wired into the circuit. Instead of making a MTC to permit visual verification, the field engineers verified open circuit continuity visually and signed Step 5.3.3 as satisfactorily completed.
(2)
On August 22, 1996, the licensee created test records that were materially inaccurate. Step 5.3.3 of WO 96-02928-00 and WO 96-02929-00 provided that at MCB, Section C, open circuit continuity be verified at 86-RASA-2(YAF) using a volt-ohm meter (VOM) across the 5-5C contacts. The field test engineers could not verify the open contacts with a VOM because of resistance in the circuit caused by a bulb and resistor wired into the circuit. Instead, the field test engineers verified open circuit continuity visually and signed Step 5.3.3 as satisfactorily completed.
These inaccuracies were material because the tests concerned functionality or operability of safety-related components. (03014)
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This is a Severity Level IV violation (Supplement 1)
MAINE YANKEE RESPONSE
- Maine Yankee agrees with this violation.
Reply to this Notice of Violation (NOV) was submitted in Reference (d). Maine Yankee is in compliance with Technical Specification 5.5. (Procedures), which is the current reference to the cited Technical Specification (5.8.2.a).
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