ML16134A539
| ML16134A539 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/04/1995 |
| From: | King L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Kellogg P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML16134A534 | List: |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 9706120373 | |
| Download: ML16134A539 (4) | |
Text
MEMORANDUM FOR:
Paul Kellogg, Chief Operational Programs Section Operations Branch FROM:
Larry P. King, Reactor Engineer
SUBJECT:
MSLB AT OCONEE DATE:
4 April 1995
Background
On February 8, 1980, IEB 80-04 was transmitted to Duke Power Company (DPC) for a response within days. On May 7, 1980 reference 11 transmitted DPC's response to Region II.
This response relied on operation action to prevent overpressurization and stated a PRA was being done for Oconee. References 9 and 10 requested additional information and stated that DPC's response was not sufficient to allow Franklin Research Company (FRC) to complete the evaluation of the potential for exceeding containment design pressure. Since the ICS is non-safety, the ICS cannot be relied upon to function correctly in an accident. Failure of the ICS may cause both MFW and AFW runout flow to the affected steam generator resulting in overpressurization of the containment.
DPC did not indicate the time necessary for the operator to take action. The request included the following:
- 1) an evaluation of the potential for exceeding containment design pressure using MFW and AFW runout flow rates
- 2) provide the time the start of a MSLB that containment design pressure will be exceeded if no operator action is taken to terminate the accident
- 3) provide the magnitude of the peak pressure and the time at which the peak occurs
- 4) provide action required to be preformed by the operator to prevent exceeding containment design pressure and provide justification for the time at which credit is taken for operator action Reference 10 also required a response to ANSI N660, Time Response Criteria for Safety-Related Operator Actions, dated March 1981 if operator action is required to terminate the action and provide justification for the time at which credit is taken for operator action. On April 9, 1982 reference 8 forwarded the request for additional information contained in reference 9. On May 7, 1980 DPC responded in reference 11 and did not address the stated requests but claimed runout.could not occur because of the design of the MFW and AFW systems. The NRC and FRC accepted this incomplete response and issued reference 5, the SER on October 14. 1982.
Discussion 9706120373 970610 PDR ADOCK 05000269 G
During the servicster inspection at Oconee whileviewing the FSAR for accident analysis it was noted that no analysis was done which required the motor operated valves to terminate feedwater on a MSLB. As part of the service water inspection and questions arising as to operability of the RB coolers, Mr. Rapp and I went to Duke offices in Charlotte. We questioned DPC engineering on the MSLB and were told that recent analysis showed the design pressure of the containment would be exceeded. They stated that a letter and a topical report had been sent to NRR concerning this issue. I mentioned the concern to Mr. Gibson and at his request forwarded reference 16 to him with my knowledge of the issue at that time. During the service water inspection.at Surry, I received a call from you to proceed to Charlotte on the following week. I than received a call to my home on Friday that canceled the trip because Mr. Gibson wanted to talk to me first.
I am still concerned that NRR has done an inadequate review and has accepted the licensees word. In the following paragraphs I will attempt to outline what I know to be the recezit facts. "Tw1il seece here however that fInd it tincredulous that -DPC has just'discovered this problem.
On May 27, 1993 DPC transmitted reference 4 stating that they have been reanalyzing Chapter 15 MSLB transient to determine the limits that the accident might impose on plans for extended fuel designs and reload design optimization. Reanalysis of the containment response identified that containment design pressure is exceeded without operator action to isolate MFW. AIt states that it meets the three acceptance criteria in Chapter 15, but ldoes not mention that Chapter 14 under Reactor Protection Criteria states in te) that the reactor building pressure due to mass and energy release within the containment boundary during thiaccident shall not exceed the reactor rbuilding design limits. Ittalso states that the equipment required to emitigate the consequences of the MSLB is qualified and would perform its safety function. There is no analysis to support this and reference 11 shows that even when the design pressure is not exceeded the temperatures exceeds the EQ requirement of 310*F in the short term and barely stays under the 290*F in the long term. Credit is taken for the RB cooling and RB spray to maintain temperature below the EQ requirements in the long term. The service water inspection identified problems with insufficient flow to the RB coolers in an accident scenario which brings into doubt it's ability to perform during an MSLB. A review of the topical report showed no analysis for the situation on page 3 of reference 4 which stated "Without credit for Automation Main Feedwater Control and Main Feedwater Control Valve Sticks Open." *A 9faims at this analysis requires the operator to take action at 120 seconds to imic the containment pressureCo--pproximately 140 psig.
All the other scenarios require action"in' 170 seconds.
Page 4 of the reference states that k ielding will take place et-A4*4vpieg.
1 n refe
.2.dated August 19, 1993 DPC committed to ak design thange and,sited tYe justification for the delayed times in implementing-hese -cnges--that -t the -equipment.is-Equalifled and e2) the accident is bounded in the USAR based on 'the'bff -site dose Gonsequences. DPC also used PRA justification for the delay in implementing the design changes.
Jdisagree widthA UNOW A
61 e e 2 for the followingreasons.
No analysis was included for the case "Without credit for Automation Main Feedwater Control and Main Feedwater Control Valve Sticks Open" was provided in referenced 12 No proof is shown that the instrumentation will meet the higher temperature as a result of the 140 psig containment pressure.
Saturation temperature for 140 psig is 360*F.
Review of other PRAs indicate that a continuing supply of feedwater to the affected steam generator or failure to isolate the non-affected steam generator will lead to decreasing RCS temperature but the maintenance of high pressure as a result of high pressure injection flow. The net effect will be the potential for pressurized thermal siock. The potential for steam generator tube rupture in the affected steam generator also exists. This is effectively a small break LOCA as the steam line break is in containment. If feedwater flow is terminated to the affected steam generator then steam flow into the containment will be terminated. However, if feedwater is NOT isolated,.there will be a continuous heat transfer from the RCS to the secondary side and into containment. In this case, it is assumed that containment heat removal will be required as the accident is equivalent to a LOCA in that majoroyc of decay heat is being transferred to the containment building.
LER 94-10-01 showed that both EFW headers exceeded the continuous operation flow limit of 1098 gpm as stated in the EFW design document. The limit is imposed to protect the steam generator cubes from the effects of flow vibration. This condition would occur in a MSLB as the EFW pumps attempted to maintain steam generator level with no back pressure. The net effect could result in tube failure. The LER stated that an evaluation of EFW response indicated that design flow could have been exceeded due to extremely low steam generator pressure.
I do not believe this problem was discovered until 1993. Crystal River and Davis Besse have secured feedwater automatically for years.
In conclusion, I learned form the SRI at Surry that the block valves do not automatically close. Failure to automatically secure feedwater would seem to ba a problem at other plants as well.
References
- 1)
L. Weins, NRC to J. W. Hampton, V. P. Oconee Site Subject - Containment Pressurization Due to MSLB Inside Containment Supplemental to IEB 80 Oconee Units 1, 2, and 3 - dated October 6, 1993.
- 2)
J. W. Hampton, V. P. Oconee Site to NRC Subject - Supplemental Response to IEB 80-04 dated August 19, 1993
- 3)
J. W. Hampton, V. P. Oconee Site to NRC Subject - LER: Design Deficiency Results In a Condition Outside the Design Basis for Main Steam Line Break dated July 1, 1993
- 4)
J. W. Hampton, V. P. Oconee Site to NRC Subject - Reanalysis of Main Steam Line Break Inside Containment
- 5)
J. Stolz, Chief Operating Reactors Branch #4 USNRC to H. B. Tucker, Vice President Nuclear Production Department.
Subject - Safety Evaluation Report for IEB 80-04 dated October,14, 1982.
- 6)
Technical Evaluation Report for Oconee Units 1, 2, and 3 dated September 28, 1982.
- 7)
W. 0. Parker Jr., Duke Power Co. to H. R. Denton, USNRC Subject - Additional Response to IEE 80-04 dated July 23, 1982
- 8)
J. Stolz, Chief Operating Reactors Branch #4 USNRC to W. Parker Subject - Request for additional information dated April 9, 1982
- 9)
S. Purdey, Project Manager Franklin Research Institute to S. Bajwa, USNRC Subject - Request for additional information dated March 15, 1982
- 10)
S. Carfagno Franklin Research Institute to S. Bajwa, USNRC Subject - Request for additional information dated January 6, 1982
- 11)
W. 0. Parker Jr., Duke Power Co. to J. P. O'Reilly Region II USNRC Subject - Response to IEB 80-04 dated May 10, 1980
- 12)
Oconee Topical Report DPC-NE-3003-P Mass and Energy Release and Containment Response Methodology transmitted by letter M. S. Tuckman to NRC dated August 11, 1993
- 13)
LER 10 Report No. 04/18/94
- 14)
FSAR Oconee Rev.11 Chapter 14.2 Standby Safeguards Analysis
- 15)
FSAR (13 DEC.1992) Chapter 15.13 Steam Line Break Analysis
- 16)
L. P. King, Reactor Engineer to A. Gibson,.Director Division of Reactor Safety Region II dated February 22, 1994 Subject - Excessive Pressure which exceeds design at Oconee