ML18039A398
| ML18039A398 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/24/1998 |
| From: | Beckner W, Sullivan E NRC (Affiliation Not Assigned) |
| To: | Hebdon F NRC (Affiliation Not Assigned) |
| References | |
| TAC-MA0276, TAC-MA276, NUDOCS 9806260141 | |
| Download: ML18039A398 (14) | |
Text
MEMORANDUMTO:
FROM:
SUBJECT:
t Frederick J. Hebdon, Director Project Directorate II-3 Division of Reactor Projects I/II
[original signed by:j William,D. Beckner, Chief Technical Specifications Branch Associate Director for Projects Edmund Sullivan, Acting Chief Materials and Chemical Engineering Branch Division of Engineering TASK INTERFACE AGREEMENT RESPONSE - REACTOR CORE ISOLATIONCOOLING SYSTEM STEAM SUPPLY LINES STEAM TRAP PIPING FLAW Plant Name:
Licensee:
Review Status:
Browns Ferry Unit 3
'ennessee Valley Authority (TVA).
Complete By memorandum dated November 26, 1997, Region II requested assistance in.determining the acceptability of TVA's.actions regarding an unresolved'item for Browns Ferry Unit 3. The unresolved item was identified in Inspection Report 50-25950-260, 50-296/97-1 0, issued on November 24, 1997.
Early on September 11, Browns Ferry management was informed that a steam leak was caused by-a through wall crack in piping associated with a steam trap in the Unit 3 Reactor Core Isolation Cooling (RCIC) system steam supply piping. The inspectors became aware of the issue at the plan-or-the-day meeting.
The flawed piping was subsequently repaired in a reasonable time frame, however, questions, regarding compliance with Technical Specifications (TS) and the NRC Inspection Manual remain unresolved.
Therefore, Region II submitted four (4) questions to the NRR staff which relate to details of the licensee's actions.
The EMCB and TSB staff have answered the four questions submitted by Region ll,.and'ur evaluation is attached.
This completes our efforts for TAC NO. MA0276.
Docket No.: 50-296
Attachment:
As stated M.W. Weston (301) 415-3151'ONTACTS:
A. D. Lee, NRR (301) 415-2735
~le Ce ter PUBLIC EMCB RF/PF A. De Agazio, PM FILENAME:G:>LEE(RCICMEM3.WPD
- See Previous Concurrence To receive a copy of this document, indicate in the box C=Copy w/o attachment/enclosure E=Copy with attachment/enclosure N = No copy OFFICE DE:EMCB E 'ADPR:TSB DE:EMCB g
. DE:EMCB NAME ALee*
MWeston*
, RHermann
'KWichman*
DATE 06/1 9/98 06/19/98 06@3'/98 06/22/98 OFF.ICE
'NAME DATE DE:EMCB EJSullivan 06/Q'98 ADPR:TSB' ti "WDBeckner
'6/
98 980626014i 980624 PDR ADOCK 050002'Pb P
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MEMORANDUMTO:
~Frederick J.,Hebdon, Director Project. Directorate II-3 Division of Reactor Projects,l/II FROM:
SUBJECT:
William D. Beckner, Chief Technical Specifications Branch Associate Director for Projects Edmund Sullivan, Acting Chief Materials and Chemical Engineering Br ch Division of Engineering TASK INTERFACE AGREEMENTS ESPONSE - REACTOR CORE ISOLATIONCOOLING Sf'STEM STEAM SUPPLY LINES STEAM TRAP PIPING FLAW Plant Name:
Licensee:
Review Status:
Browns Ferry Unit 3 Tennessee Valley Auth rity (TVA)
Complete By memorandum dated Nov ber 26, 1997, Q gion II requested assistance in determining the acceptability of TVA's aetio s regarding 6 unresolved item for Browns Ferry Unit 3. The unresolved item was identified in I spectio Report 50-259, 50-260, 50-296/97-10, issued. on November 24, 1997.
Early on Sep mba 11, Browns Ferry management was informed that a steam leak was caused by a through Illcrack in piping associated with a steam trap in the Unit 3 Reactor Core Isolation Coolie ( (IC) system steam supply piping. The inspectors became aware of the issue at the plan-or-tge-day meeting.
The flawed piping was subsequently repaired in a reasopabte time'fqame, however, questions regarding compliance with Technical Specifications (T8) and the NR+lnspection Manual remain unresolved.
Therefore, Region II submitted four (4) questions o the NRR staff which relate to details of the licensee's actions.
The EMCB and TSB sta have answered the four questions submitted by Region II, and our evaluation is atta (ed. This completes our efforts for C NO. MA0276.
Docket No.: 50-29
Attachment:
s stated CONTA S:
A. D. Lee, NRR (301) 415-2735
~
0 F'le Center PUBLIC EMCB RF/PF A. De Agazio, PM ILENAME:G:>LEE>RCICMEM2.WPD To receive a copy ofthis document, indicate i the box C=Copy w/o attachment/enclosure E=C py with attachment/enclosure N = No co'p 0 A.Ai~
OFFICE DE:EMCB E
ADPR;"TS DE:EMCB NAME ALe DATE 06 98 MWeston 06/
98, RHermann 06/
/98 KWichman 06 8
OFFICE DE:EMCB NAME EJSullivan ADPR:TSB
, WDBeckner DATE 06/
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t UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 3une 24, 1998 MEMORANDUMTO:
FROM:
Frederick J. Hebdon, Director Project Directorate II-3 Division of Reactor Projects t/it William D. Beckner, Chief Technical'Specifications Branch Associate Director for Projects Edmund Sullivan, Acting Chief Materials and Chemical Engineering Branch Division of Engineering
SUBJECT:
TASK INTERFACE AGREEMENT RESPONSE - REACTOR CORE ISOLATIONCOOLING SYSTEM STEAM'SUPPLY LINES STEAM TRAP PIPING FLAW Plant Name:
Licensee:
Review Status:
Browns Ferry Unit 3 Tennessee Valley Authority (TVA)
Complete By memorandum dated November 26, 1997, Region II requested assistance in determining the acceptability of TVA's actions regarding an unresolved item for Browns Ferry Unit 3. The unresolved item was identified in Inspection Report 50-259, 50-260, 50-296/97-10, issued on November 24, 1997.
Early on September 11, Browns Ferry management was informed that a steam leak was caused by a through wall crack in piping associated with,a steam trap in the Unit 3 Reactor Core Isolation Cooling (RCIC) system steam supply piping. The. inspectors became aware of the issue at the plan-or-the-day meeting. The flawed piping was subsequently repaired in a reasonable time frame, however, questions regarding compliance with Technical Specifications (TS) and.the NRC Inspection Manual remain unresolved.
Therefore,,'Region II submitted four (4) questions to the NRR staff which relate to details of the licensee's actions.
The EMCB and TSB staff have answered the four questions submitted by Region II, and our evaluation is attached.
This completes our efforts for TAC NO. MA0276.
Docket No.: 50-296
Attachment:
As stated CONTACTS:.A. D. Lee, NRR (301) 415-2735 M.W. Weston, NRR (301) 415-3151
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1.0 By memorandum dated November 26, 1997, Region II requested assistance in determining the acceptability of TVA's actions regarding an unresolved item for Browns Ferry Unit 3. The unresolved item was identified in Inspection Report 50-259, 50-260, 50-296/97-10, issued on November 24, 1997.
Region II submitted four (4) questions for consideration by the NRR staff Early on September 11, Browns Ferry management was informed that a steam leak was caused by a through wall crack in piping associated with a steam trap in the Unit 3 Reactor Core Isolation Cooling (RCIC) system steam supply piping. The inspectors monitored the licensee's actions after learning of the incident at the plan-of-the-day meeting.
The licensee performed a safety assessment. where they identified the flawed pipe as part ofthe reactor coolant pressure boundary.
The flawed pipe is American Society of Mechanical Engineers (ASME) Code Class 2 piping. Approximately two days after the source of the leak was determined, the licensee replaced the steam trap, isolation valves, and the connected piping.
Although the piping was repaired in a reasonable time, several questions remained regarding compliance with Technical Specifications (TS) and the NRC Inspection Manual. Further details are included below in the answers to questions 1R.
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Io The licensee's position is that the requirements of T.S. 3.6.G.1.b are not applicable for conditions (such as this flaw) found at power.
Is this correct' QgggK: T.S. 3.6.G.1.b. states, "With the structural integrity of any ASME Code Class 2 or 3 equivalent component not conforming to the above requirements (TS 3.6.G.1), restore the structural integrity. of the affected component to within its limits or isolate the affected component from all OPERABLE systems".
This" TS is applicable for conditions found at power, and this conclusion was also reached by NRC Regional Management for Browns Ferry Unit 3.
Specifically, T.S. 3.6.G.1 states that the structural integrity of Class 1, 2, and 3 components ATTACHMENT
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shall be maintained throughout the. life of the plant, which means under every condition (i.e. at power or not at power).
Since the steam leak was non-isolable, even after shutting the available manual isolation valves and the requirements of T.S. 3.6.G.1 could not be met, the requirements of T.S. 3.6.G.1.b. were applicable.
IfT.S. 3.6.G.1.b. is applicable, did the licensee's actions meet all.of the requirements of the T.S.?
~~vL~Z: The licensee did not meet the requirements of T.S. 3.6.G.1.b. which are stated in the answer. to question 1 above.
Several hours after the flawwas found, the licensee shut the manual isolation valves which did not isolate the steam leak. To isolate the leak at this point, the licensee would'have had to shut the RCIC steam line isolation valves and declare RCIC inoperable.
- However, the licensee did not declare RCIC inoperable while pursuing repair options. The T.S. requires that a component be declared inoperable when the structural integrity cannot be, restored within its limits, which was the case for approximately two days.
While the T.S. does not address the time frame for isolating the affected component, Section 6.15 of NRC Inspection Manual Chapter 9900 clearly states that upon discovery of leakage from a Class 1, 2, or 3 component pressure boundary, the licensee should declare the component inoperable.
The staff interprets "upon discovery" to mean immediately. The staff agrees that there appears to be no safety consequences associated with the licensee's actions.
The Inspection'Manual Chapter is discussed in more detail in the answer to Question 3.
Did the licensee's actions meet the expectations promulgated in Section 6.15 of NRC Inspection Manual Chapter 9900 regarding actions to be taken ifa leak is discovered in a Class 1, 2, or 3 component?
~Em: The licensee's actions did not meet the expectations promulgated in Section 6.15 of Inspection Manual Chapter 9900 regarding actions to be taken if a leak is discovered in a Class 1,2, or 3 component.
Section 6.15 states that "If a leak is discovered in a Class 1, 2, or 3 component in the conduct of inservice inspections, maintenance activities, or during. plant operation, IWA-5250 of Section XI [ofthe ASME Code] requires corrective actions be taken. based on repair or replacement in accordance with Section XI". Section 6.15 also states that "Upon discovery of leakage from a Class 1, 2, or 3 component pressure boundary (i.e. pipe wall, valve body, pump casing, etc.) the licensee should declare the component inoperable."
Instead of shutting the RCIC system steam isolation valves and declaring the system inoperable, the licensee opted to perform an evaluation while also pursuing repair options.
RCIC was.not taken out of service until the repair was performed which was approximately two days after the source of the leakage was identified. Section 6.15 of Manual Chapter 9900 clearly states that the
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component should be declared inoperable upon discovery of leakage from the component pressure boundary.
Although the repair was reasonably prompt, the Manual Chapter suggests that the delay in declaring RCIC inoperable is not acceptable regardless of the corrective actions that were pursued by the licensee.
Licensee management based its decisions, in part, on a 1992 ASME Code interpretation.
Code interpretations are not a part of NRC regulations, or of the ASME Code. The NRC's position on ASME Code interpretations is, discussed in more detail in the answer to Question 4.
While ASME Code interpretations are clearly not part of the Code, licensees utilize the information presented in the interpretations.
It appears that there is conflict between several interpretations and the Inspection Manual guidance.
Is it appropriate for these apparent disparities.to be addressed and ifso, have they been'
~~ig: The licensee applied the information in a 1992 ASME Code Interpretation to conclude that IWA-5250 did not apply because the flaw was not identiTied during inservice inspection.
Specifically, the Code Interpretation indicated that IWA-5250 is not applicable during maintenance activities or plant operations and that an operability determination should be performed as a result of identification of the leak. This Code Interpretation conflicts with the guidance in Section 6.15 of Manual Chapter 9900.
Conflicts between Code Interpretations and NRC requirements are addressed in the Technical Guidance of Part 9900, and in the proposed rule change to 10 CFR 50.55a.
As stated in the Technical Guidance of Part 9900, "ASME Code Interpretations are not incorporated into the Code of Federal Regulations, and therefore, the NRC is not bound by these interpretations."
The guidance goes on to state that "While the NRC acknowledges that the.ASME is the official interpreter of the Code, the Regulations transcend the Code.
Since Code Interpretations are not part of the regulations, licensees should'exercise caution when applying Interpretations to their facilities.
The proposed rule change to 10 CFR 50.55a highlights the fact that since Interpretations are issued after. the provision that it refers to, it can affect the NRC's understanding of the Code Editions and Addenda that are incorporated by reference into the regulations.
The proposed rule change also notes that, in some cases,. Interpretations have been issued which conflict with NRC requirements, and these cases resulted in enforcement actions.
The Technical Guidance of Part 9900,:the proposed rule change to 10 CFR 50.55a, and'nforcement action are the methods that have been used to alert licensees regarding the NRC's policy on Code Interpretations that conflict with NRC guidance.
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The requirements of T.S. 3.6.G.1.b. were applicable for the flaw found in the Browns Ferry Unit 3 RCIC system steam supply piping. The licensee's position is that the T.S. is not applicable for conditions (such as this flaw) found at power. Published NRC guidance leads the NRR staff to conclude that the licensee did not meet the requirements of the T.S. since RCIC should have been declared inoperable when the steam leak could not be isolated.
The licensee's actions did not meet the expectations promulgated in Section 6.15 of Inspection Manual Chapter 9900 regarding actions to be taken ifa leak is discovered in a Class 1, 2, or 3 component.
Although the repair was reasonably prompt (approximately two days after discovery of the source of the leakage), the licensee did not declare RCIC inoperable while pursuing repair options.
Conflicts between ASME Code Interpretations and NRC requirements are addressed in the Technical Guidance of Part 9900 of the NRC Inspection Manual, and in the proposed rule change to 10 CFR 50.55a.
In addition, the proposed rule change also notes that, in some cases, Interpretations have been issued which conflict with NRC requirements, and these cases resulted in enforcement actions.
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