05000289/LER-1997-001-02, :on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected

:on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected

ML20136B759
Person / Time
Site:	Crane
Issue date:	02/17/1997
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20136B750	List: 05000289/LER-1997-001, Forwards Corrected Copy of LER 97-001-00 Re Condition Found During Review of GL 96-06 05000289/LER-1997-001-02, :on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected
References
LER-97-001-02, LER-97-1-2, NUDOCS 9703110112
Download: ML20136B759 (7)

LER-1997-001, on 970117,potential Overpressurization of Piping Between Closed Reactor Building Containment Isolation Valves Occurred Due to Inadequate Design Code Guidance. Results of Review Will Be Documented.Pages 2 & 3 Corrected

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(b)(2)(ii)
2891997001R02 - NRC Website

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POTENTTAL OVERPRESSURIZATION OF PIPfNG BETWEEN CLOSED REACTOR Bl.f!LDING l

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on January 17,1997, during a review orpiping systems ths penetrate containment as requested j

. by Generic Letter (GL) 9606, " Assurance of Equipment Operminhty and Containment Integrity During Design-Baals Accidera Conditions,"GPU Nuclear determ'wl that most of the eleven (1I) j sfrected piping segments could be stressed beyond the piping design code (B31.1-1967) allowable

~ stressee during a desipbasis acculent. This condition was ibund to be reportable in accordance l

with 10CFR50 73(aX2XiiXB) as a condition in which the plant was found to be outside cfits i

design basis. The analysis n=l=I=d the efibets on mternal Ruid and piping in response to an j

va,ternal ambient,eenperasure ineresse. The results revealed that ahhough the piping did not meet the design requiremasta, the postulated stresses did not exceed ASME Section III, Appendix F i

criteria for riping. Therefbre containment integrity would be mai==ined durlas an accident.

Usias the sh in NUREGCR-5455, ' Development of the NRC's Human Performance j

Investigation Process (HPIP),"the root cause was found to be that the applicable demsn standards were 'tems than edesguste - conAising or incomplete." Operability datarminations were performed i

l and further evaluations are ongoing to determine the need for enod:Scations or procedural revisicas. There were no adverse safhty convoquences or safety implications that resulted from l

this event, and the event did not afBect the health and safety of the pubbc 9703110112 970304 wacresu m per)

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POTENTIAL OVERPRESSURIZATION OF PIPING BETWEEN CLOSED REACTOR BUILDING CONTAINMENT ISOLATION VALVES DUE TO INADEQUATE DESIGN CODE GUIDANCE I.

Background:

The US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 96-06,

" Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions,"which requested licensees to determine 1) if their containment air cooler cooling water systems are susceptible to either water hammer or two-phase flow conditions during postulated accident conditions and 2) if piping systems that perietrate containment are susceptible to thermal expansion of fluid so that pressurization ofpiping could occur. Systems found to be susceptible to the conditions discussed above were to be assessed for operability and appropriate corrective actions taken.

Piping segments susceptible to condition 2 above will have liquid trapped between two closed containment isolation valves following an accident inside containment. The containment environment following an accident condition (MSLB or LB LOCA) will promote heat transfer to the fluid trapped within the piping resulting in an increase in the.

fluid pressure. The pressure increase will place stresses on the piping and challenge containment integrity to survive the design basis accidents. The increased pressure will also challenge the code allowable design stress limits for the affected piping systems.

II.

Plant Operating Conditions before Event:

On January 17,1997, TMI-l was operating at 100% power.

I 1

III.

Status of Structures, Components, or Systems that were Inoperable at the Start of the j

Event and that Contributed to the Event:

1 i

None.

IV.

Event Description

A. Overview On January 17,1997 during the review of piping systems that penetrate containment and i

are susceptible to thermal expansion of fluid trapped between the isolation valves, GPU

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LICENSEE EVENT REPORT (LER)

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DOCMET LER NUMBER (8)

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Nuclear determined that eleven (11) piping segments would pressurize beyond the original specified piping design pressure. The' review indicated that the majority of the eleven -

j segments could exceed the piping design code (USAS B31.1.0 - 1967) allowable stresses j

under the postulated conditions. Seven (7) of the eleven (11) segments could exceed material yield stresses. An operability assessment was immediately conducted in -

accordance with the guidance in GL 91-18. Although seven (7) of the analyzed segments i

could exceed material yield stresses, the criteria in ASME Section III, Appendix F, " Rules j

for Evaluation of Service Loadings with Level D Service Limits," was satisfied. Satisfying l

ASME Section III, Appendix F criteria demonstrated that piping integrity and, therefore, j

containment integrity would be maintained. The piping segments were determined to be operable since their containment function would be maintained.

f Although the piping segments were determined to be operable, they were also determined j

to be degraded since the code allowable stresses are exceeded under the postulated conditions. Since piping design code requirements are part of the design basis of the plant, i

GPU Nuclear concluded that the analysis results show the listed piping segments to be outside design basis of the plant, which is reportable under the requirements of 10 CFR j

50.73(a)(2)(ii)(B).

l The eleven (11) piping segments affected are:

1

1) & 2) Once Through Steam Generator (OTSG) Sampling Lines "A" and "B" (AB/ PSP),*

3) Intermediate Closed Cooling Water (ICCW) Return Line (CC/ PSP),

4) Reclaimed Water Supply Line (CA/ PSP),

5) Makeup and Purification Letdown Outlet Line (BQ/ PSP),

6) Pressudzer and Reactor Coolant Sampling Line (AB/ PSP),

7) Reactor Coolant Pump (RCP) Seal Water Return Line (BQ/ PSP),

8) Reactor Coolant Drain Tank (RCDT) Transfer Line (WD/ PSP),

9) Reactor Coolant Pump Cooling Return Line (CC/ PSP), and 10 & 11) Core Flood Tank Sampling Lines "A" and "B" (BP/ PSP).

The seven (7) of eleven (11) piping segments that could exceed material yield stresses are:

1) & 2) Once Through Steam Generator (OTSG) Sampling Lines "A" and "B,"

3)ICCW Return Line,

4) Makeup and Purification Letdown Outlet Line,

5) Pressudzer and Reactor Coolant Sampling Line

6) & 7) Core Flood Tank Sampling Lines "A" and "B.

NRC FORM 305A (4-96)

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TaxT trmore spece a r.guesa. u.amonet copw onwec rorm assg on Of the remaining four (4) piping segments, the stresses may be found to be within the l'

ANSI B31.1.0 - 1967, Appendix A, Table A-1 allowable stress values for the piping

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i material upon further more detailed analysis. However, the piping segments were l

considered to be within the scope of the event report since the final pressure was found to be above the design pressure. It is likely that some of these lines will be found acceptable i

upon completion of a more detailed analysis.

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B. Description ofPressurization Analysis l

Several containment piping segments were determined to be susceptible to the j

pressurization concern identified in GL 96-06 through a screening of all the containment j

piping segments. Piping segments were evaluated as no.t being susceptible based on the following:

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1) The penetration does not contain a liauid.

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2) The penetration has access to a relief valve.

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3) The inboard containment isolation is a check valve with a downstream pressure relief valve or a pressure relieved component.

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Susceptible piping agments could have liquid trapped between two closed valves following an accident inside containment. The accident conditions inside containment promote heat transfer to the fluid trapped within the piping resulting in an increase in the fluid pressure. The pressure increase would place stresses on the piping and challenge the ability of the penetration to survive the design basis accidents. A stress calculation for each of the identified piping segments assessed the impact of this condition on primary containment integrity.

Results of the stress calculation revealed that several of the affected piping systems exceed their yield stresses, but do not exceed ASME Section III, Appendix F - Level D criteria.

Therefore, the piping segments affected by the GL 96-06 concern would not rupture and would maintain their integrity.

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NRC FORM 3084 U.S. NUCLEAR REAULATORY COMMi&SION (4-95) l LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACluTY NAME(1)

DOCKET LER NUMBER (8)

PAGE (3) m -

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THREE MILE ISLdND, UNIT 1 05000289 5

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96 002 00 TEXT (frmore space is regunod. use ockssonal copos of NRC form 3664) (17)

V.

Component Data:

Below is design data on the susceptible piping segments:

Penetration 213/214 302 307

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Description

OTSG Samples ICCW Return Reclaimed Water

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"A" & "B" Supply Pipe Size 3/8" Schedule 40 &

6" Schedule 40 2" Schedule 40 j

3/8" tubing

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Pipe A312 Type 304 A106 Gr. B Carbon A312 Type 304 l

Material Stainless Steel Steel Stainless Steel j

Pipe / Tubing l

Design 1050 psi 175 psi 150 psi i

Pressure i

i Penetration 309 328 329

Description

Letdown Outlet PZR/RX Sample RCP Seal Water Return Pipe Size 2-1/2" Schedule 160 3/8" tubing 4" Schedule 160 1

Pipe A312 Type 304 Stainless Steel A312 Type 304 Material Stainless Steel Tubing / Pipe Stainless Steel Design 2500 psi 2500 psi 250'0 psi Pressure I

Penetration 331 347 348/349

Description

RCDT Transfer RCP Cooling Water CF Tank Sample Return "A" and "B" Pipe Size 3" Schedule 40 8" Schedule 40 1" Schedule 80 Pipe A312 Type 304 A106 Gr. B Carbon A312 Type 304 Material Stainless Steel Steel Stainless Steel Design 150 psi 150 psi 700 psi Pressure NRC FORM 386A (4-95)

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LICENSEE EVENT REPORT (LER)

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DOCKET LER NUMBER (8)

PAGE (3) l P - - - (Z)

YEAR 5tuVENTIAL REV N

THREE MILE ISLdND, UNIT 1 05000289 6

OF 7 96 002 00 TEX f (Ymore space la required, use ackssonal copsos c(NRC Form 366A) (17)

VI.

Identification ofRoot Cause a

The guidance of NUREG/CR-5455, " Development of the NRC's Human Performance Investigation Process (HPIP)" was used to determine a potential root cause. In the HPIP Module for Organizational Factors / Management Systems the near root cause screening

)

question, "Was an event caused by standards, policies, or administrative controls (SPAC) that were confusing, incomplete, unclear, ambiguous, not strict enough or otherwise i

inadequate?", can be answered "Yes." Further screening of the applicable root cause questions resulted in identifying that the probable root cause of exceeding the design code i

requirements was: SPAC less than adeauate - confusing or incomplete.

J s

Automatic or Manually Initiated Safety System Responses:

VII.

l No safety system responses occurred or were required to occur.

l VIII. Assessment of the Safety Consequences and Implications of the Event:

j TMI-l Technical Specification 3.6.1 requires that Containment Integrity be maintained whenever the Reactor Coolant System (RCS) pressure is 300 psig or greater, the RCS temperature is 200 degrees F or greater, and nuclear fuel is in the core.

GPU Nuclear concludes that containment integrity would be maintained since the structural analysis of the susceptible piping segments determined that the ASME Section III, Appendix F - Level D Criteria was not exceeded. Additionally, none of the affected piping segments are required to be reopened to perform a giost-accident safety function.

Therefore, based on the assessment above, there were no adverse safety consequences or safety implications that resulted from this event, and this event did not affect the health and safety of the public.

IX.

Previous Events of a Similar Nature:

There have been no previous LERs at TMI-l related to overpressurization of piping systems that penetrate containment.

4 NRC FORM 306A W96)

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Corrective Actions

A. Corrective Actions Taken:

i Upon determining that the susceptible piping segments could be pressurized and stressed i

beyond the design code allowable limits, the analysis per ASME Section III, Appendix F i

for Level D condition was performed. Again, this analysis showed that although some l

stress levels were beyond yield, the stress level was well below the 0.7 Su allowable per

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Appendix F.

L B. Actions Planned to Prevent Recurrence:

1 l

On completion of further more detailed analyses, procedure changes, hardware 1

modifications, or code rectification will be performed as necessary to assure compliance j

with'the applicable design codes'. The results of this review will be documented in a GPU i

l Nuclear technical data report (TDR) and reported to the NRC in a supplement to our response to GL 96-06, dated February 14,1997, i

i i

• The Energy Industry Identification System (EIIS), System Identification (SI) and Component Function j

Identification (CFI) Codes are included in brackets, "[SI/CFl]." where applicable, as required by 10 j

CFR 50.73 (b)(2)(ii)(F).

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' Material Nonconformance Report (MNCR) No. 970002 was initiated to track the disposition of l

the penetration pressurization issue.

NRC FORM 308A (4 06)

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