ML20127K852

Forwards Responses to Questions Transmitted Via NRC on ABWR Shutdown Risk Study (App 19Q)

ML20127K852
Person / Time
Site:	05200001
Issue date:	01/13/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9301260237
Download: ML20127K852 (6)

Text

__

i GE Nuclear Energy 0t :2'u* h 4 (l';l DJ':fd1p

' -i rw, < " vu.

1 31 ;

1 l

January 13,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Suhmittal Supporting Accelerated ABWR Review Schedule

Dear Mr. Posiusny:

l Attached are the responses to seven questions from the NRC on the ABWR shutdown risk study (Appendix 190). These questions were transmitted to GE by a letter from C. Poslusny (NRC) to P. W. Marriott (GE) on December 23,1992.

Sincerely, 1

J. N. Fox Advanced Reactor Programs l-l ec: J. D. Duncan (GE) l N. D. Fletcher (DOE) i Jrno2 720t:: L

,yvf c

T301260237 930113

~

77 2 i

PDR ADOCK 05200001

'fl

[-

A PDR

<4

t. -

01/13/93 1

Responses to NRC Questions on ABWR Shutdown Risk Dated December 23,1992 l

4 l

Question 1) Reliability of power available for reactor water cleanup (RWCU) system to j

measure coolant temperatures. This open item regards the method used to measure water temperature at the residual heat removal (RHR) pump discharge and the RWCU system j

suction lines, in case ofloss of all power.

I z

i Response 1) - Accurate measurement of Reactor Water Cleanup (CUW) temperature i

requires operation of the CUW pump and the temperature sensors. The CUW motor is.

[

normally powered by offsite power.. Upon loss of ofTsite power, the CUW motor can be i

powered by either the emergency diesel generators or the onsite Combustion Turbine Generator (CTG). The temperature sensors are powered by non -lE DC Bus A10 which is backed up by non 1E batteries Question 2) Adequacy of the decay heat removal (DHR) system parameters monitored by the ABWR instrumentation system. This open item regards the potential applicability 1 of the GL 88-17's recommendations for RHR system indications (e.g., pump noise, pump suction pressure, motor current, alarms, etc.)..

i l

Response 2) GL 88-17 currently applies only to PWRs. It addresses the concern for ioss ofinventory during low inventory (i.e., mid-loop) operations. The ABWR like all BWRs does not require a reduction in RPV inventory to complete routine maintenance.

Nevertheless, the ABWR does have adequate instrumentation to monitor the condition of i

DHR sy.;temsc The ABWR design incorporates the following RHR system monitoring functions to ident;fy a loss of DHR:

f

- Pump discharge pressure high.

l_

- Pump motor overload, i'

l

- RHR system MOV power loss or thermal' overload l-

- RHR loop logic power failure.

[

- RHR pump motor trip.

L

}

- RCW outlet temperature high.

i i

- RHR pump low flow.

The ABWR RHR monitoring features listed above meet the intent of GL 88-17 to L

have " continuous monitoring of the DHR system (s)." These ihnctions all alarm in the l

control room and will give the operator adequate warning of a loss of DHR.

l 9

l l.

r l

....~..w.

1.,

01/13/93 i

i Question 3) GE operating procedure guidelines for conducting outage and planning.

Response 3) Section 19Q.10 of the ABWR SSAR contains a discussion of generic procedure guidelines for outage planning. Reference is made to NUMARC 91-06 l

" Guidelines to Enhance Safety During Shutdown" which contains additional guidelines for i

preparing and implementing an outage plan to minimize shutdown risks. Section 19Q.7 of the ABWR SSAR describes one possible maintenance philosophy to reduce shutdown risks. Using this philosophy, maintenance would be completed on a divisional basis (i.e.,

one entire ECCS division along with support systems be taken out of service at the same time and one ECCS division including support systems be administratively controlled to not be in maintenance). The third division would be operating to remove decay heat and its associated EDG may be in maintenance. The administratively controlled ECCS division would also have its fire / flood barriers intact This configuration has been evaluated to ensure that the risks associated with loss of the operating RHR loop are acceptab!y low if a minimum set of systenis (safety and nonsafety) are not in maintenance.

4 An example of such a minimum set for Mode 5 (reactor cavity unflooded)is: one RHR train, one CRD pump, and the AC Independent Water Addition System (i.e., diesel driven fire water pump). Tables 19Q.7-2 through 4. of the SSAR list other minimum sets for various shutdown Modes.

l The shutdown risk analysis (19Q.7) has shown that systems required to be operable by technical specifications in addition to normally operating systems (e.g., CRD, CUW) are sufficient to ensure adaquete shutdown risk margins. Also, the above configrration ensues adequate internal flooding and fire protection. Attachment lor l

describes the ABWR flooding analysis.

l l

Question 4) Shutdown technical specifications for the ABWR.

Response 4) The ABWR technical specifications are contained in Chapter 16 of the SS AR.. As discussed in section 19Q.7, the results of the shutdown risk study confirmed the adequacy of the existing ABWR technical specitications. No additional. technical specification requirements were identified for shutdown conditions. Availability of current technical specification systems in conjunction with normally running systems (e.g., CRD, CUW, and fire water) are suflicient to maintain adequate shutuown risk margins.

Question 5) Potential impact on inflatable seals due to temperature increase from loss of DHR cooling during reactor internal pumps replacemerit or maintenance.

Response 5) The inflatable seals on the reactor internal pump (RIP) shaft are permanently installed and are thus designed to handle normal plant operating temperatures. Therefore, any temperature increase due to the postulated loss of DHR -

during shutdown will have no impact on the seals performance.

2 J

g-

i. -

01/13/93 4

Question 6) Discussion of direct reactor pressure vessel boiling to the containment as an acceptable alternate decay heat removal method.

}

Response 6) Boiling as a viable decay heat removal method is discussed in section l

19Q.7.5. An analysis was completed to show that offsite doses due to boiling in the RPV l

during Mode 5 (i.e., the RPV head removed) were much lower than regulatory limits. All components relied upon for core cooling are either qualified for a harsh (i.e., steam) environment (e.g., RHR), will not be in a steam environment (e.g., condensate), or due to their hardy construction are expected to remain available for a significant period of time in a low pressure steam environment (e.g., CRD). In addition, operator action inside containment following loss of DHR is only required for actuation of the AC Independent Water Addition System. The two manual valves that require opening can be opened well in advance of the steam environment occuring (i.e., once DHR is lost, several hours are j

typically required before boiling begins, during this period before steaming occurs the -

l operator could open the valves).

Question 7) Impact on safety-related equipment from heavy loads and fuel handling in the containment during the refueling process.

l Response 7) The heavy loads evaluation for the ABWR is contair.ed in SSAR Section 9.1.5. This section discusses heavy i)ad lifts near safety related equipment in the containment. The conclusions from the heavy loads analysis as discussed in section 9.1.5.2 are that no heavy load drops in containment violate the heavy load safety criteria ofcausing:

!)

a release of radioactivity l

2) a criticality accident 3) the inability to cool fuel within the reactor ves.sel or within the spent fuel i

pool, t

or 4) prevent a safe shutdown of the reactor Thus, heavy load drops in containment are not a safety concern.

Two confirmatory items regarding the CUW system were contained in the 12/23/92 letter.

These items were potential design changes to the CUW system 'o:

l 1)

Only isolate and bypass the filter demineralizers instead of the entire

. CUW system on a CUW ion exchanger inlet high temperature signal.

3

~

y i

i 01/13/93

f -

3 1

i 2)

RPV bottom line drain valve in the line connecting to the CUW system to be changed to remote / manual nitrogen-operated instead of the current manual i

design.

j Both of the above design changes are currently being implemented arid the details of these

~

changes will bc incorporated in the next SSAR revision in February 1993.

1 i

+

u 4

i

?-

I i

?

5 i

I l

i 4

4 4

4-

,,.. ~ _.. _ _ - -...,. ~, _. _. _

,,.. _ -... ~., _

..