ML20041F033

Response to NUREG-0803,Generic Safety Evaluation Rept Re Integrity of BWR Scram Sys Piping.

ML20041F033
Person / Time
Site:	Hatch
Issue date:	02/08/1982
From:	BECHTEL GROUP, INC.
To:
Shared Package
ML20041F029	List: ML20041F028 ML20041F033
References
RTR-NUREG-0803, RTR-NUREG-803 TAC-43733, TAC-43738, NUDOCS 8203160110
Download: ML20041F033 (8)

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J EDWIN 1. HATCH NUCLEAR PLANT UNITS 1 AND 2 RESPONSE TO NUREG 0803

" GENERIC SAFETY EVALUATION REPORT REGARDING INTEGRITY OF BWR SCRAM SYSTEM PIPING" t

PREPARED BY BECilTEL POWER CORPORATION CAITl!ERSBURG MD.

FOR 1

GEORGIA POWER COMPA!W ATLANTA, CA.

4 FEBRUARY 8, 1982

' 8203160110 p M ADOCK O hh3 PDR

TABLE OF CONTENTS

1. INTRODUCTION II. PIPING INTEGRITY III. MITIGATION CAPABILITY IV. ENVIR0hWENTAL QUALIFICATIONS V. CONCLUSIONS

EDWIN 1. HATCH RESPONSE TO NUREG-0803

1. INTRODUCTION NUREG-0803 " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping" presents a discussion of the possibility and consequences of a crack in the scram discharge (SD) system piping after a scram. The conclusions reached in NUREG-0803 are based in large part on the generic General Electric report on this subject, NEDO-24342 " General Electric Evaluation of BWR Scram System".

This response is prepared by Bechtel for Georgia Power.

This response to NUREG-0803 is intended to:

1) Provide the plant specific input requested by NUREG-0803, Section 5.

2) Support our conclusion that a scram discharge (SD) system pipe crac?

need not be considered.

3) Support our conclusion that if a scram discharge (SD) system pipe crack must be considered, it can be effectively mitigated provided certain minor modifications are made to the plant along with some procedural and Technical Specification changes.

II. PIPING INTEGRITY

1) Periodic Inservice Inspection and Surveillance for the SD System Currently, the scram discharge piping is pressure tested periodically as part of Hatch's ISI program. Weld examination of this piping and its supports is not presently required by the program. With the exception of the scram discharge headers all this piping is exempted from ISI of welds by the Code because of size. The headers are not exempted because of size, but are not presently included in Georgia Power's ISI weld examination program because they are rarely pressurized during normal plant operation, and then only to a pressure far below design pressure. This is believed to meet the intent of the Code.

However, due to the potential consequences of a crack in the scram discharge system, we recommend that Georgia Power (GPC) commit to including weld examination of the scram discharge headers in Hatch's ISI program.

2) Seismic Design Verification The Hatch scram discharge system piping is designed to Seismic Category I requirement. As discussed later in this response, reanalysis of this piping based on as-built information is currently being performed as necessary to ensure the adequacy of this piping to withstand seismic and other original design loads. This effort is being accomplished per the requirements of IEB 79-14.

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3) Hydraulic Control Unit (HCU) - Scram Discharge System (SD) Equipment Procedures Review We' recommend that CPC review and update Hatch maintenance procedures us necessary to provide sufficient guidance to ensure that loss of scram discharge system integrity will not occur at times when such integrity is required.

4) As-built Inspection of SD Piping and Supports As-built inspection of all scram discharge system piping (large and small) as well as the insert / withdraw lines has been performed as part of our IEB 79-14 review.

The short term operability requirements of this Bulletin have been met for major portions of this piping, and the remaining portions are at various stages of screening. The evaluations required to ensure full Code compliance are currently being performed. We plan to complete our evaluations by August, 1982. We recommend that GPC commit to making any necessary modifications during the first refueling outage thereafter.

It should be noted that our scram discharge piping is designed for 280 F maximum temperature, not 450 F as implied by General Electric in NEDO-24342. In a recent letter General Electric indicated that occasional temperatures in excess of the original General Electric design basis of 280 F should be considered, but should not be treated as a new design basis. Regardless, we have performed a sensitivity study based on elevated temperatures in the scram discharge piping and have a high degree of assurance that the piping design is adequate for the elevated temperature.

III. MITIGATION CAPABILITY

1) Improvement of Procedures If the SD crack must be considered GPC would be required to develop procedures to enable the operator to adequately detect and mitigate this event. Upon detection of an unisolable primary system leak in the reactor building, the operator would first be directed to reset the scram. This would reclose the scram valves and effectively isolate the SD crack flow. Detection of this leak could come from many different signals. Of the various detection signals discussed in NUREC-0803, the most definitive would be the sump level alarms, the area radiation monitors and the reactor building low differentini pressure alarm. Any of these alarms would alert the operator to an unisolated primary system leak in the reactor building although not necessarily that the leak is from a SD crack. If scram reset is not possible, the operator would be directed to initiate depressurization of the reactor in order to decre ise the leak rate regardless of the source of the primary system lea'.

After the reactor has been depressurized and the reactor building environment has improved enough to allow personnel access, entry will be made into the reactor building to locate and isolate the leak. If a SD crack is discovered, it will involve closing 137 manual scram discharge line isolatic . valves.

I The revised procedures would conform to any NRC approved BWR Owner's Group Emergency Procedure Guidelines written per NUREG-0737. The implementation schedule for these updated procedures would be in accordance with NUREG-0737 Item I.C.1 requirements.

Note that the crack detection instrumentation we are taking credit for is of high quality, but is non-class IE in nest cases. Also, most is not available if offsite power is lost. However, we believe that the existing detection is adequate since: the probability of losing of fsite power coincident with a SD crack is extremely remote; and the detection instruments need only perform their function early in the event before any possible adverse environment could affect their operability.

2) Limitation of Coolant Iodine Concentration to Standard Technical Specification Values NUREG-0803 recommends that the reactor coolant activity be limited to the standard technical specification (STS) limit of 0.2 u Ci/gm.

This is the current Hatch Unit 2 activity limit. If a SD crack must be considered, GPC would be required to lower the Hatch Unit I activity limit to this same value in order to assure that offsite dose limits would not be exceeded.

IV. ENVIRONMENTAL QUALIFICATIONS

1) Environmental Qualification of Prompt Depressurization Function Refer to Section IV.6 of this evaluation.

2) Verification of Equipment Designed for Water Impingement NUREG-0803 discusses water impingement of essential equipment in corner rooms due to SD crack flow overflowing down the stairwells.

We believe that the floor drains on the main operating floor on which the scram discharge piping is located are more than adequate to handle the maximum expected crack flow rate. However, if a SD crack must be considered, we recommend that GPC install curbing around the stairwell entrances to the RCIC and two (2) RHR corner rooms below.

This would prevent overflow via stairwells into these rooms and would direct any overflow to the CRD corner room which contains no equipment required to mitigate this event.

We conclude based on the above discussion that essential equipment in the corner rooms need not be designed for water impingement.

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3) Verification of Equipment Qualified for Wetdown by Water at 212 F.

NUREG-0803 discusses wetdown of essential equipment in the corner rooms due to leakage through equipment hatches. The equipment ,

hatches for Hatch utilize a step plug arrangement and are located at the high points of the operating floor. Leakage through them is not expected. However, if a SD crack must be considered, we recommend that GPC provide gaskets for these hatches as a precautionary measure to ensure that leakage will not occur.

Weconcludebasedontheabovediscussionthatessentialequfpment in the corner rooms need not he qualified for wetdown by 212 F water.

4) Verification of Feedwater and Condensate System Operation Independent of the Reactor Building Environment Prior to reactor depressurization the operator may choose to use the feedwater and condensate pumps to maintain vessel inventory if they are available. After depressurization he may use just the condensate pumps. Note that the feedwater/ condensate system is non-safety grade and will not be available if offsite power is lost. Also,the feedwater pumps are turtine driven and therefore cannot operate if the main steam isolatian valves are closed. The condensate booster pumps are motor drive.. There are no feedwater/

condensate system components located in the reactor building which could prevent operation of this system.

We conclude that while the feedwater/ condensate system would most likely be available following a SD crack, we would not take credit for its operation from a safety standpoint.

5) Evaluation of Availability of HPCI/RCIC Turbines Due to the long duration of the postulated unisolable SD crack flow it is possible that the leak detection system for the HPCI and RCIC systems may sense a leak in one or both of these systems.

This would cause isolation of the HPCI and/or RCIC systems while they are still required for high pressure makeup to the vessel.

Isolation would not occur for at least fifteen (15) minutes after the SD crack for HPCI and longer for RCIC. At this point or shortly thereafter,we believe the operator would be able to determine that the source of leakage is not from the HPCI or RCIC systens. .He could then bypass the signal calling for isolation of these systems to enstre their continued operation. The operator also has the option to manually initiate ADS to quickly depressurize the vessel so that the low pressure systems can take over. This will occur automatically if HPCI or RCIC are not restarted.

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6) Verification of Essential Components Qualified for Service at 2120 and 100% Humidity.

If a SD crack must be considered, we recommend that GPC ensure that the equipment required to mitigate this event is adequately qualified for the worst expected accident environment. This qualification effort would be performed in acco rdance with the NRC's ongoing equipment qualification (EQ) program.

As discussed earlier, most of the detection instruments we have taken redit for are non-class lE. These would not be included in the EQ program since they need only perform their function early in the event before any possible adve rse environment could affect their operation.

V CONCLUSIONS Based on the discussion presented in this E. I. Hatch plant specific response to NUREG-0803, we have reached the following conclusions:

1) A scram discharge system pipe crack need not be considered for the plant. This conclusion is supported by the fact that the scram discharge system consists of high quality ASME Section II, Class 2

- Unit 2 (or ANSI B31.7, Class 2 - Unit 1) Seismic Catego ry I piping. As built inspection of this piping has been performed per I&E Bulletin 79-14. Reanalysis of this piping is cu rrently being performed to ensure full code compliance. No major deficiencies are expected, and if found we have recommended that GPC ensure that they are co rrec ted in a timely manner. The scram discharge system is rarely pressurized during normal plant operation and then only slightly. Furthe rmo re , we recommend that GPC commit to including the scram discharge header in our ISI weld inspection program.

Also, we have recommended that GPC review and update as necessary, their plant maintenance procedures to ensure that tha scram discharge system integrity will not be jeopardized at times when this integrity is required.

The BWR Owner's Group is in the process of reviewing the GE probability assessment presented in NED0-24342, and plans to revise this overly conservative assessment to make it more repre sentative of the existing plants. The revised probabilistic risk assessment (PRA) will demonstrate that the probability of a SD crack occurring is sufficiently low that it need not be considered.

2) If a sc ram discharge system pipe break must be considered, this event can be effectively mitigated provided certain minor modifications are made to the plant along with some procedural and Technical Specification changes. The modifications and changes that would be required are discussed in the body of our evaluation. In addition, it will have to be assured that the mitigetion equipment is qualified for the expected environment.

This ef fort -wold be covered under the ongoing EQ program. Also, the above conclusion requires that we take credit for short-term operation of certain high quality, but non-class 1E, crack detection instrumentation which may not be available if offsite power is lost.

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