ML20154H642
| ML20154H642 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 10/08/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154H589 | List: |
| References | |
| EA-96-397, EA-97-374, EA-97-559, NUDOCS 9810140278 | |
| Download: ML20154H642 (7) | |
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(NOTICE 2)
Maine Yankee Atomic Power Company Docket 50-309 Maine Yankee Atomic Power Station License No. DRP-36 EA 96-397;97-374; 97-559 l
Based on investigations by the NRC Office of Investigations (01), conducted between j
December 1995 and October 1997, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement t
l Actions, NUREG-1600, the violations are listed below:
1 1.
PRINCIPAL PROBLEM RELATED TO INADEQUATE SMALL-BREAK-LOSS OF-COOLANT ANALYSES (OI Report No. 1-95-050) i
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A.
VIOLATION RELATING TO INABILITY TO ANALYZE ENT RE BREAK SPECTRUM FOR CYCLE 14 10 C.F.R. I 50.46(a)(1) requires, in part, that emergency core cooling system (ECCS) performance must be calculated with an acceptable i.
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 ll.4, requires that to the extent practicable, predictions of the evaluation model, or portions thereof, shall be j
compared with applicable experimental information.
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Contrary to the above, from October 14,1993, through January 25,1995 l
(during Cycle 14 operations), and in the Cycle 14 Core Performance Analysis Report (CPAR) submitted August 25,1993, Maine Yankee Atomic Power i
Company (MYAPCo) used unacceptable models to calculate ECCS l
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 l
spectrum between.35 fta and at least.6 ft for which no acceptable j
evaluation model was capable of calculating cooling performance or reliably ca'culating cooling performance. MYAPCo calculated Small-Break Loss-of-Coolant Accident (SBLOCA) ECCS performance with the code described in l.
"YAEC 13OOP, 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 SSLOCA ECCS performance only up to the.35 ft2 break
' size because the RELAPSYA SBLOCA evaluation model documented in 9810140278 9C1008 L
PDR ADOCK 05000309 G
PDR I
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YAEC-1868 was incapable of calculating ECCS performance for break sizes 2
of and greater than 0.35 ft as a result of the model's terminating after the l
safety injection tank actuation due to numerical convergence errors for the 2
l break size of.35 ft. MYAPCo calculated Large-Break Loss-of-Coolant l
(LBLOCA) ECCS Performance with the LBLOCA analysis described in YAEC-1160," Application of Yankee WREM-Based Generic PWR ECCS Evaluation l-Model to Maine Yankee", dated July 1978 (WREM). Although the WREM LBLOCA evaluation model was subsequently demonstrated in 1996 to be 2
l capable of calculating ECCS performance down to the.6ft 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 been acceptable to calculate ECCS performance for break sizes in the small-break l
region of 0.6 ft and above because the evaluation model was not compared I
to applicable experimental data to demonstrate its reliability in calculating ECCS performance in the small-break region. (01012) i B.
VIOLATION RELATING TO INABILITY TO ANALYZE ENTIRE BREAK SPECTRUM FOR CYCLE 15 L
10 C.F.R. I 50.46(a)(1) requires, in part, that emergency core cooling l
system (ECCS) performance must be calculated with an acceptable evaluation model and must be calculated for a number of postulated loss-of-l 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 ll.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 i
(CPAR) submitted December 1,1995, Maine Yankee Atomic Power Company (MYAPCo) used unacceptable models to calculate ECCS performance and failed to calculate a number of postulated loss-of-coolant l
accidents of different sizes, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents I
were calculated. Specifically, there was a portion of the small-break spectrum between.35 ft and at least.6 fta for which no acceptable evaluation model was capable of calculating cooling performance or reliably calculating cooling performance. MYAPCo calculated Small-Break Loss-of-l 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 i
of which were described as an Appendix K approach to RELAP5YA).
j MYAPCo calculated SBLOCA ECCS performance only up to the.35 ft2 break size because the RELAP5YA SBLOCA evaluation model documented in
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YAEC-1868 was incapable of calculating ECCS performance for break sizes of and greater than 0.35 ft2 as a result of the model's terminating after the safety injection tank actuation due to numerical convergence errors for the 2
break size of.35 ft. 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.6fta break size, the WREM LBLOCA evaluation model was not used to calculate ECCS performance in the small-break region for Cycle 15, and would not have been acceptable to calculate ECCS performance for break sizes in the small-break region of 0.6 ft 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 ll problem (Supplement 1).
ll OTHER VIOLATIONS RELATED TO INADEQUATE SMALL-BREAK-LOSS-OF-COOLANT ANALYSES (Ol Report No. 1-95-050)
A.
VIOLATION RELATING TO OPERATING CYCLE 13 Technical Specification (TS) 5.14.2, " Core Operating Limits Report," for the Maine 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 13OOP, 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. 5 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 RELAPSYA 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 Cycle 13 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
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analysis, which was not listed in TS 5.14. The SBLOCA analysis listed by TS 5.14 is "YAEC 1300P, RELAP5YA: A Computer Program for Light Water Reactor System Thermal Hydraulic Analysis, Volumes 1,2,3, dated October 1982" (RELAP5YA). This inaccurate information was material to the NRC because it was a representation that RELAP5YA, which had been approved for application to MYAPS pursuant to the Three Mile bland Action Plan, Item il.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) l B.
VIOLATION RELATED TO IMPROPER APPLICATION OF ALB-CHAMBRE CORRELATION FOR CYCLE 14 10 C.F.R. 5 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 at'ove, 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, MYAPCo calculated ECCS performance for SBLOCAs with an unacceptable evaluation model. MYAPCo used the ECCS code described in YAEC-13OOP,"RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, Volumes 1, 2,3," dated October 1982 (RELAP5YA), and the plant-specific 1
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 of large penetration factors and a large cross flow resistance factor in the split downcomer nodalization. (02024) l This is a Severity Level IV violation (Supplement 1)
C.
VIOLATION RELATED TO IMPROPER APPLICATION OF ALB-CHAMBRE CORRELATION FOR CYCLE 15 10 C.F.R. 5 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, 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 Thermal-Hydraulic Analysis, Volumes 1,2,3," dated October 1982 (RELAP5YA), and the plant-specific RELAP5YA SBLOCA evaluation model described in YAEC-1868, " Maine
- -. - - - - -~_- -.
5 Yankee Small Break LOCA Analysis" (YAEC-1868). RELAP5YA as applied was not an acceptable evaluatica model because the nodalization model of YAEC-1868 incorrectly applied the Alb-Chambre correlation, resulting in the unjustified use of large penetration factors and a large cross flow resistance factor in the split downcomer nodalization. (02034)
This is a Severity Level IV violation.
D.
VIOLATION RELATING TO ANALYSIS OF REDUCED STEAM GENERATOR PRESSURE FOR CYCLE 14 10 C.F.R. 5 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. 5 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.
I 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,"RELAPSYA: 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, 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) 111.
VIOLATION ASSOCIATED WITH SAFETY SYSTEM LOGIC TESTING l
(01 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 concerning 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.
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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 SlAS 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 FTl 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. O-16 3.
10 C.F.R. I 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 XVil, " 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-OOand 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-RASA-2(YAF) using a voit-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 makir.g 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 96-02928-00and WO 96-02929-OOprovided 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)
This is a Severity Level IV violation (Supplement 1)
Pursuant to the provisions of 10 CFR 2.201, Maine Yankee Atomic Power Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region I, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a
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" Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.
Under the authority of Section 182 of the Act,42 U.S.C. 2232, this response shall be submitted under oath or affirma!!on.
Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards info:mation so that it can be placed in the PDR without redaction, if personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of !aformation will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Dated et King of Prussia, Pennsylvania this 8th day of Octobe 1998
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