ML20054H664

Forwards Final Evaluation of SEP Topic III-5.B, Pipe Break Outside Containment. Plant Adequately Protected from Dynamic Effects of Pipe Break Outside Containment,Subj to Resolution of Identified Items

ML20054H664
Person / Time
Site:	Oyster Creek
Issue date:	06/22/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Fiedler P JERSEY CENTRAL POWER & LIGHT CO.
References
TASK-03-05.B, TASK-3-5.B, TASK-RR LSO5-82-06-077, LSO5-82-6-77, NUDOCS 8206240311
Download: ML20054H664 (13)

Text

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June 22,1982 3

Docket Ho. 50-219 LS05-82-06-077 Mr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 4

SUBJECT:

SEP TOPIC III-5.8, PIPE BREAK OUTSIDE CONTAINMENT OYSTER CREEK NUCLEAR GENERATING STATION On July 10, 1980, we transmitted to you a draft safety evaluation report on the SEP Topic III-5.B along with a request for additional information.

By letter dated October 6,1980, you provided responses to the above letter. A further request for information from the staff was sent to you on April 16, 1981. By letter dated September 30, 1981, you provided the infonnation responding to our April 16, 1981 letter.

l Based on the information provided in the above references, we have issued g) l the enclosed final topic evaluation. This topic is now complete.

of pipe break outside containment subject to resolution of the following D We conclude that the plant is adequately protected from the dynamic effects l

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in the Integrated Plant Safety Assessment:

pool The staff's review of the licensee's submittal on the fractu 1.

ics analysis of the emergency condenser piping.

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The various options (as listed in the enclosure) to resolve the issue on the effects of pipe breaks beyond the flued head for the main steam and reactor water cleanup piping systems.

This evaluation will be a basic input to Oe integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is l

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completed.

s206240311 820622 PDR ADOCK 05000219 Sincerely, P

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l 6/14/82 6/ /82 k/J//82 Dennis M. Crutchfield, Chief 6/16/82 6/0 /82 dtFu:UL btVU:UL btFb UL Operating Reattors Branch No. 5 SEPB:DL '

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7 Docket No. 50-219 LS05 Mr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

SEP TOPIC III-5.B. PIPE BREAK OUTSIDE CONTAlfEENT OYSTER CREEK NUCLEAR GENERATING STATION Un July 10, 1980, we transmitted to you a draf t safety evaluation report on the SEP Topic 111-5.8 along with a request for additional information.

By letter dated October 6,1980, you provided responses to the above letter. A further request for information from the staff was sent to you on April 16, 1981. By letter <*ated September 30, 1981, you provided the information responding to our April 16, 1981 letter.

p Based on the information provided in the above references, we have issued the enclosed final topic evaluation. This topic is now complete.

We conclude that the plant is adequately protected from the dynamic effects of pipe break outside containment subject to resolution of the following in the Integrated Plant Safety Assessment:

1.

The staff's review of the licensee's submittal on the fracture mechan-ics analysis of the energency condenser piping.

2.

The licensee's option (as listed in the enclosure) to resolve the issue on the effects of pipe breaks beyond the flued head for the main steam and reactor water cleanup piping systems.

This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assessment may be revised in the future if your facility design is changed or if NRC criteria rela-ting to this topic are nodified before the integrated assessment is cocpleted.

i Sincerely, ORB #5 AD:SA:DL SEPB:DL SEPB:

I DCrutchfield Glainas EMdKEnha Dennis H. Crutchfield, Chief CGrimes 6/ /82 6/ /82 6////82 Operatin Reactors Branch No 6/l(p/82 l

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NRC FORM 318 (1480) NRCM 024o OFFICIAL RECORD COPY j

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Oyster Creek Docket No. 50-219 Revised 3/30/82 Mr. P. B. Fiedler t

CC G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trewbridge c/o U. S. NRC

'/1800 M Street, N. W.

Post Office Box 445 Washington, D. C.

20036 Forked River, New Jersey 08731 i

J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102-

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. Ronald C. Haynes, Regional Admi'nfstrator r

Nuclear Regulatory Commission, Region I

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631 Park Avenue King of Prussia, Pennsyl.vania 19406 J..Knubei BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney Gcneral

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State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Forked 31ver, New Jersey-~ 08731 U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor

~~ Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 e

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SYSTEMATIC EVALUATION OF PIPE BREAK r

OUTSIDE CONTAINMENT' TOPIC III-5.B FOR OYSTER CREEK NUCLEAR GENERATING STATION a

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TABLE OF CONTE'iTS I.

INTRODUCTION

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REVIEW CRITERIA

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III. RELATED SAFETY TOPICS IV.

REVIEW GUIDELINES V.

EVAL'sATION VI.

CONCLUSION

SYSTEMATIC EVALUATION PROGRAM TOPIC III.5.B 0YSTER CREEK NUCLEAR GENERATING STATION TOPIC:

III-5.B, Pipe Break Outside Containment I.

INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.B; " Pipe Break Outside of Containment," is to assure that pipe

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breaks would not cause the loss of required function of " safety-related" systems, structures and components and to assure that the plant caqn be safely shut down in the event of such breaks.

The required function of safety-related systems are those functions re-quired to mitigate the effects of the pipe break arid safely shut down

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the reactor plant.

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II.

REVIEW CRITERIA General Design Criteria 4 (Appendix A to 10 CFR Part 50) requires in part that structures, systems and components important to safety be appropriately protected against dynamic effects,. such as pipe whip and discharging fluids, that may result from equipment failures.

The current criteria for review of pipe breaks outside containment are contained in Standard Review Plan 3.6.1, " Postulated Piping Failures in Fluid Systems Outside of Containment," including its attached Branch Technical Position, Auxiliary System Branch 3-1 (BTP ASB 3-1) and Standard Review Plan 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of. Piping," in-cluding its attached Branch Technical Position, Mechanical Engineering Branch 3-1 (BTP MEB 3-1)

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III.

RELATED SAFETY TOPICS A.

This review complements that of SEP Topic VII-3, " Systems Required for Safe Shutdown."

B.

The environmental effects of pressure, temperature, humidity and flooding due to postulated pipe breaks are evaluated under Unre-i solved Safety Issue A-24, " Qualification of Class 1E Safety-Related j

Equipment."

C.

The effects of potential missiles generated by fluid system ruptures and rotating machinery were also considered and are evaluated under SEP Topic III-4.C, " Internally Generated Missiles."

D.

The original plant design criteria in the areas of seismic input analysis design criteria are evaluated under SEP Topic III-6,

" Seismic Design Considerations."

IV.

REVIEW GUIDELINES The staff used an " effects oriented" approach to determine the acceptability of plant response to pipe breaks, i.e., each structure, system, component, and power supply which must function to mitigate the effects of the pipe y

break and to safely shut down the plant was examined to determine its susceptibility to the effects of the postulated break.

Break effects con-sidered were:

ccmpartment pressurization, pipe whip, jet 4apingment, spray, and flooding.

V.

EVALUATION The staff issued a draf t safety evaluation on SEP Topic III-5.B on July 10, 1980 (Reference 1).-

As part-of this evaluation, the staff included f~our items for which additional information was requested and two positions con-cerning actions to be taken by the licensee. The licensee submittal of October 6,1980 (Reference 2), provided responses to the six open items f rom Reference 1 (Four questions and two positions).

Each of these items, the licensee response and the staff resolution are discussed below.

Staff Question 1 The licensee should provide a comparison of the design of the containment penetration piping outside centainment between the containment and the outermost containment isolation valves for the main steam lines, emergency condenser steam and condensate lines, and reactor water cleanup lines with the provisions of section B.1.b of Branch Technical Position MEB 3-1 (ap-pended to Standard Review Plan 3.6.2) in sufficient detail to identify the degree cf conformance with and deviations from these provisions.

Staff Position 1 Because inadequate-protection exists from the effects of postulated bre'aks in the emergency condenser steam and condensate lines on the 75' elevation of the reactor building, the licensee should submit, by September 1,1980, a schedule for modifications to be installed to provide adequate protection from these postulated breaks.

The modifications must be in accordance with the acceptance criteria of Standard Review Plan 3.6.1 and provide protection for the emergency condenser isolation valves and controls and for cable trays V22, V23, 41, 42, and 43.

Licensee Responses to Staff Question 1 and Staff Position 1 In summary, the licensee concluded that the Oyster Creek containment pene-trations and piping outside containment meet the intent of the provisions of section B.1.b of BTP MEB 3-1 except as noted below:

A.

The main steam, emergency condenser, and reactor water cleanup piping systems for Oyster Creek were designed and constructed in accordance with the Code for Pressure Piping ASA B31.1 (1955). Also, the design l

I stress limits for normal and upset plant conditions plus seismic loads were different from those specified in BTP MEB 3-1.

However, the mar-gin of safety provided by ASA B31-1 is considered to be comparable to that provided by ASME Code Section III. Therefore, the Oyster Creek piping is considered to meet the intent of BTP MEB 3-1 in this regard.

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B.

BTP MEB 3-1 requires that piping between the centainme'nt' penetration and pipe whip restraint located outboard of the outside isolation valve be analyzed for postulated pipe breaks beyond the restraint and meet certain stress limits. There are no pipe whip restraints on the Oyster Creek main steam, emergency condenser, and reactor water clean-up lines outboard of the outside isolation valv'es.

Because of the lack of pipe whip restrainti,'the Oyster Creek piping systems between 'the

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containment penetration and isolation valve outside containment do not meet the stress requirements of BTP MEB 3-1 with ' regard to pipe breaks beyond the outside isolation valve. 'As a result, pipe breaks are as-sumed between the containment penetration and outside isolation valve.

Effects of these postulated breaks are summarized below:

1.

Emergency Condenser The emergency condenser piping is protected with a guard pipe from the containment penetration up to the isolation valves outside con-tainment.

Thus, postulated HELBs in the emergency condenser piping between the containment penetration and outside isolation valve are considered to be the same as pipe breaks inside containment.

Ac-cordingly, such breaks do not have any effect on safe shutdown sys-tems located outside containment.

As part of continuing discussion of this subject, the staff reques-ted additional information concerning the methods for mitigating the effects of high energy line breaks for this piping, the demon-stration of the " leak before break" failure mode, and the clarifi-cation of the proposed augmented inservice inspection program l

(Reference 3).

In its September 30, 1981 submittal (Reference 4),

the licensee agreed to perform a fracture mechanics analysis to demonstrate that the emergency condenser pipings on the 75.' level will leak before a significant break could occur, in addition to l

the previously proposed installation of a local leakage monitoring l

system to detect any leakage cue to small cracks in these lines, l

The licensee also agreed to conduct visual examinations for evi-dence of leakage with the system pressurized privr to or during l

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I each refueling shutdown. The frequency of inservice inspections presently required by the Oyster Creek Technical Specification will ba doubled; that is, the welds in the system on the 75' level will

'nspected on a schedule which will cover all of the welds in each 5-year period, In its May 18, 1982 submittal (Reference 5) the licensee docketed a o

fracture mechanics analysis of the emergensy condenser system and pre sented the results in a meeting on June 2,1982, for Oys'ter Creek Nuclear Generating Station.

2.

Main Steam and Reactor Water Cleanup The effects of pipe breaks for both lines between containment pene -

tration and the isolation valve oGtside conta'inment are acceptable except that an assumed single active failure of the inside contain-ment isolation valve would result in an unisolated break. The licensee concluded that continued operation of the plant is justified on the basis of the extremely low probability of a break in the sec-tion of pipe between the containment rienetration and isolation valve outside containment along with an assumed non-mechanistic failure of the isolation valve inside containment.

t Staff Resolution Since the licensee was unable to demonstrate that portions of the piping systems discussed herein meet certain stress limits as required by the current criteria and the postulated breaks in these portions of piping could result in an unisolated break, the staff has come to the following conclusions:

1.

Emergency Condenser c.

The staff is currently reviewing the fracture mechanics analysis sub-mitted by the licensee to demonstrate " leak before break" for the emergency condenser piping outside of the containment. The staff expects to provide the results of the review as well as a recommenda-tion regarding the need for local leak detection for incorporation in the integrated assessment.

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Main Steam Line and Reactor Cleanup 1

The staff has determined from discussions with the licensee that no stress data were available to demonstrate that these piping systems between the containment penetration and the isolation valve,outside containment meet certain stress limit requirements of BTP MEB 3-1.

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Therefore, the following options are recommended for consideration in l

the Integrated Assessment:

I a.

The licensee should provide information to demonstrate that the maximum stress in the portion of the pipe between the flued head and the outboard isolation valve, as calculated by Eg.(9) in paragraph NR-3652 of ASME Section III code under the loadings re-sulting from a postulated piping f ailttre of fluid system piping beyond the isolation valve does not exceed 1.8 S.

If the maximum h

stress. exceeds 1.8 Sh value, the licensee can either add supports to reduce the maximum stress or proceed with the option (b) below.

b.

The licensee should demonstrate by fracture mechanics analysis that a double ended pipe break could not occur beyond the flued head, or proceed with Option c.

c.

The licensee should demonstrate by risk analysis that the effects of pipe breaks 'f6r'both lines beyond the containment penetration ~

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along with an assumed single active failure of the inside contain-ment isolation valve would not result in an unacceptably high risk.

Staff Question 2 Provide a comparison of the design of the containment penetration piping outside containment for the emergency condenser steam lines and reactor water cleanup lines with the provisions of section B.2.C of Branch Technical Position ASB 3-1 (appended to Standard Review Plan 3.6.1) in sufficient detail to identify the degree of conformance with and deviations from these questions.

Licensee Response to Staff Question 2 Based on the comparison of the containment penetration and piping outside scatainment for the emergency condenser steam lines and reactor water cleanup lines with the provisions of section B.2.C of BTP ASB 3-1 the licensee conclu-ded that the Oyster Creek containment penetration and piping meet the intent of the provisions of section B.2.C of BTP ASB 3-1 except as those noted in the above Licensee Response to Staff Question 1 for the emergency condenser and reactor water cleanup piping systems. The licensee's proposed corrective actions are the same as those described in the same response above.

Staff Resolution Same as Resolution for Staff Question 1.

Staff Question 3 The licensee should provide an evaluation of the potential effects of damage to cable tray 13A on the 51' elevation of the reactor building from a postula-ted break in the reactor water cleanup system, considering the effects of pipe whip, jet impingement, and high temperature on the electrical cables.

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Licensee Response to Staff Question 3 The licensee noted that cable tray 13A contains only one circuit related to a safe shutdown system, a control cable for one safety relief valve on the auto depressurization system. The licensee concluded that unavailability of this component would not prevent safe shutdown following a postulated break in the reactor water cleanup system.

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' However, it was also noted that the power cabler to the' reactor' water cleanup isolation valves 1.ocated outside containment are located in cable tray 13A and 14A. Damage to these cables in combination with a single active failure could result in failure of both isolation valves (inside and outside containment) to close on demand.

Continued operation was considered by the licensee to be justified on the basis.of the low probability of' the postulated br,eak in com-bination with a non-mechanistic failure of the inside isolation valve.

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Staff Resolution

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.The interaction with the auto depressurization system control cable is consid-ered acceptable since this component would not be needed to mitigate the break and bring the plant to a safe shutdown condition. The interaction with the cleanup isolation valves is more fully addressed under Staff Question 2 above.

Staff Question 4 The licensee should provide an evaluation of potential flooding in the cable spreading room from a postulated break in the fire water system or turbine building closed cooling water system. The evaluation should determine the depth of flooding, what equipment could be flooded, and the effects of loss of that equipment.

Licensee Response to Staff Question 4 The floor of the cable sprading room contains two 4-inch diameter drains which discharge to the basement floor of the turbine building. The total. capacity of these drains exceeds the flow rate possible assuming a break in the-piping.

Thus the water would be drained from the room before the water depth could flood safety-related equipment.

Staff Resolution l

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The capacity of even one drain far exceeds the flooding rate from the postula-ted through-wall leakage crack in the fire water or turbine building closed cooling water system piping. The water is collected in the turbine building basement sump which alarms on high level.

No safety-related equipment is I

located in the basement. Therefore, the staff considers that this issue is l

i closed and no further licensee action is required, t

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_7 Staff Position 2 To provide adequate protection from the effects of postulated main steam and main feed line breaks in the turbine building mezzanine area', the licensee should move or provide protection for all core spray system control cables in that area.

By September 1, 1980, the licensee should provide a schedule for resolution of this issue.

Licensee Response The licensee stated the additional protection was not required for the core spray system control cables in the turbine building mezzanine area because the NRC draft evaluation did not fully consider -the redundancy provided in the Oyster Creek core spray systems. Each of the two redundant core spray systems contains sufficient power supplies, pumps and valves such that no, single active failure would prevent the system from operating.

Staff Resolution The staff concurs that each core spray system can withstand a single active f ailure and stili fulfill its safety function. Therefore, a main steam /

feedwater pipe break affecting one core spray system in combination with a single failure in the other system would not disable the core spray function.

Therefore, we consider this item to be resolved with no further licensee action required.

VI.

CONCLUSION Based on the above discussion, the staff concludes that the plant is adequately protected from the dynamic effects of pipe break outside containment subject to resolution. of the following in the, Integrated Plant Safety Assessment:

A.

The staff's review of the licensee's submittal on the fracture mechanics analysis of the emergency condenser piping.

B.

.The licensee's option to resolve the issue on the effects of pipe breaks beyond the flued head for the main steam and reactor water cleanup piping systems.

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REFERENCES 4

1.

NRC Letter, D.M. Crutchfield to I.R. Finfrock, Jr. (JCP&L) dated July 10, 1980.

2.

JCP&L Letter, I.R. Finfrock, Jr. to D.M. Crutchfield JNRC) dated October 6, 1980.

3.

NRC Letter, D.M. Crutchfield to 1.R. Finfrock, Jr. (JCP&L) dated April 16, 1981.

4.

JCP&L Letter, J.T. Carroll, Jr.-to D.M. Crutchfield (NRC) dated September 30, 1981.

5.

JCP&L Letter, P.B. Fiedler to D.M. Crutchfield (NRC) dated May 18, 1982.

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