ML20087D264

Forwards Response to Agenda Items 1,5,9 & 16 Discussed at Ge/Nrc Reactor Sys Branch 911120-21 Meetings.Items Include, Stability Performance in Normal Operating Region,Loss of Ac Power & Loss of Feedwater Heating Transient

ML20087D264
Person / Time
Site:	05000605
Issue date:	01/10/1992
From:	Marriott P GENERAL ELECTRIC CO.
To:	Pierson R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
EEN-9206, MFN-010-92, MFN-10-92, NUDOCS 9201160018
Download: ML20087D264 (13)

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GE Nuclear Energy NUrfiM be-P C rd/04-fhp C

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'9

!MFN No. 010 92-

- Docket No-STN 50 605

- EEN 9206 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:

Robert C. Pierson, Director

- Standardization and Non Power Reactor Project Directorate

Subject:

GE Response to Agenda items 1,5,9 and 16 Discussed During the l

GE/NRC Reactor Systems liranch Meeting on November 20 21 -

1991 Enclosed are thirty-four'(34) copies of the GE response to.the subject itemi

- It is intended that GE will amend the SSAR, where appropriate, with'these responses in a future amendment.

Sincerely,

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P.W. MgrEo t, Manager Regulatory and Analysis Services M/C 382, (408) 925-6948 s cc: F. A. Ross '.

(DOE)

N. D. Fletchei

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(

C. Posiusny, Jr.

(NRC).

. R. C. Berglund

' (GE)-

J. F. Quirk 1(GE)'

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RESPONSE 440.187 As discussed in the Response to Question 100,1, the ABWR design assures the stability performance in the normal operating region is more stable than current operating BWRs by incorporating the following design features:

(1)

Smaller inlet orifices, which increase the inlet single-phase pressure drop, and, consequently, improve the core and channel stability.

j (2)

Wider control rod pitch, which increases flow area, and, consequently, reduces the void reactivity coefficient and improves both core and channel stability, and (3)

More steam separators, which reduce the two phase pressure drop, and improve the stability.

In order to reconfirm this conclusion, a stability analysis based on the procedures developed by the BWROG committee on thermal hydraulic stability (Reference 1) was performed for the ABWR. In this analysis, conservative nuclear conditions, taking into consideration of future core design, were assumed. The results at the most limiting conditions in the normal operating region (i.e.t the intercept of 102 % rod line with all operating RIPS at their minimum speeds, assuming only 9 out of 10 RIPS are in operation) are as follows:

Core Decay Ratio 0.72, Channel Decay ratio 0.36.

These results are also shown in Figure 1 together with the criteria.. From Figure 1, it is confirmed that that ABWR is stable in the normal operating region.

It should be noted that the likelihood of operation outside the normal region has also been minimized by the. ABWR design.

There are ten recirculation pumps served by four power supplies. The Recirculation Flow Control System has a' triplicated logic incorporating a minimum speed demand. In addition, each pump has an Adjustable Speed Drive with a fixed minimum speed setpoint.

1

Furthemiore, automatic logics (Figure 2) which prevent plant operation in the region with the least stability margin are also implemented. This design is similar to Option I-A, one of long-tern, solutions considered by the BWROG. In addition, in order to meet the stability design requirements specified in the ALWR Utility Requirements Document, Option Ill, LPRM based Oscillation Power Range Monitor (OPRM), which is also one of long-i temi solutions considered by the BWROG, will be implemented in the ABWR design, when the OPRM design is approved by the NRC.

As for issues relates to ATWS stability, they sre of no concems to the ABWR design, since the ABWR design has logic to automatically initiate the SLCS, including automatic initiation of feedwater run back. Furthennore, the ABWR EPG will incorporate any changes recommended by the BWROG.

In summary, the ABWR stability design is consistent with the licensing methodology proposed by the BWROG committee on thermal hydraulic stability. The ABWR will be stable in the normal operating region.

Reference 1:

NEDO-31960, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," June 1991.

9 p

Figure 1. ABWR Stability 1.0 0.9 08 Unstable Region 0.7

! 0.6 ABWR (Design Basis) x 0 0.5 8

20.4 0

0.3 0.2 0.1 0.0 O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Channel Decay Ratio ABWR is stable in normal operating domain l

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j LPRMs D

Regional APRMs

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Core-wide Time 5

History Display Core Flow Sensor Power

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1. Power 2 30%: To assure power level below 80% rod line at natural circult. tion.

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2. Flow s 36%: To assure flow rate _is higher than that of eight RIPS operations i

with minimum pump speed Figure 2. Stability Controls and Protection Logic' l

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(5)

Inss of AC Power The RCIC :,yst :m is designed to perfonn its function without AC power for at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Supporting systems such as DC power and the water supply will.

support the RCIC system during this time period. Without AC power, RCIC room cooling will not be available. However, room temperature will not rea i the equipment maximum emironmental temperature within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. (also see Subsection 19E.2.1.2.2 for additional information)

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ABWR UWWAB y

i Standard Plant REV C

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y depressurization systems perform adequate core _ (1)laloss of coolant (LOCA)eventt-L coolin g to preve nt - c.xcessive - f u el clad

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temperature during LOCA event.

Detailed (2) vessel isolated and maintained at hot l

discussion of RCIC meeting this GDC is described standbyt i

,L in Subsection 3.1.2.

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- (3) vessel isolated and accompanied by loss of Compliance with GDC 36. ' The RCIC system is coolant flow from the reactor feedwater designed such that in service inspection of the systemt j

l system and its components is carried out in accordance with the intent of ASME Section XI.

(4) complete plant shutdown with 1oss of nornlal l

The RCIC design specification requires layout and Icedwater before the reactor is depressu -

l arrangement of the containment penetrations, ired to a level where the shutdown coolin, process piping, valves, and other critical system can be placed in operationt or j

equipment outside the reactor vessel, to the

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maximum practical extent, permit access by

5) losr5FAC power for 30 minutes.

personnel and/or. appropriate equipment for testing and inspection of system integrity.

A tance criteria 11.3 of SRP Section 5.4.6 states t the RCIC system must p its Compliance with GDC 37. The RCIC system is function w. out the availabilit any AC l

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power.1 Review rocedure 11,1 W ther requires designed such that system and its components can that there be su(ffl4en)t be periodically tested to verify operability, atTery capat>ility' for.-

Systems operability is demonstrated by

.two hours of operayti. While RCIC.is designed:

preoperational and periodic _ testings in for 30 minutta pi operati during loss of ac

)-

accordance with RG 1.68 Preoperational test power, the tery capacity s Id allow over will ensure proper functioning of controls, four ho of operation, which wo tt instrumentation, pumps and valves. Periodic Qequi f

ment.

testings confirm systems ava, lab _ility and.

i operability through out the life 'of the plant.~

Deirg loss of AC power, RCIC when started at--

During normal plant operation, a full flow pump water level 2 is capable of preventing water test is being performed periodically to assure level from dropping below the level which ADS systems design flow and head requirements are mitigates (Level 1). This accounts for= decay attained. Ali RCIC systems components are -

heat boils ff and primary syr. tem leakages, o

capable of individual functional testings during l

plant operation, ' This includes sensors, Following a reactor' scram, steam generation instrumentation, control logics, pump, valves,- will continue at a reduced rate due to the core.

and more. Should the need for RCIC operation ;. fission product, decay heat. At this-:ime-the occur while the system is being tested, the RCIC-turbine bypass system will divert the steam to system and its components will automat, ally re aligned to provide cooling water into the reactor. The above test requirements satisfy GDC 37.

5.4.6.1 Design Basis l

The reactor core isolation cooling -(RCIC) system is a safety system which consists of a turbine, pump, piping, valves, accessories, and E

instrumentation designed to assure that suffi.

l cient reactor water inventory is maintained in

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the reactor vessel to permit adequate core cool-ing to take place. This prevents reactor fuel overheating during the following conditions:

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(9) Loss of FW Heating Transient a

For ABWR design, the following design requirement is specified for the-

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FW heating system design ::

"No single operator error or equipment-failure shall cause loss'of more:

than 55 9C (100 oF) feedwater heating."

The reference steam and power conversion system shown in Figures 10.1-1 to 10.1-3. meets thie requirement. In fact, the FW temperature drop based on the-reference heat balance shown in Figure 10.1-2 is as follows:

- isolation of one low pessure heater

< 15 0F

- isolation of one high pressure heater _

< 28 0F

-- isolation of one low pressure heater string

< 530F-

- isolation of one high pressure heater string 1

< 53 0F1 4

Therefore, the use of 100 oF temperature drop in the transient analysis is conservative.-

A drop of 150 0F occurred at a domestic BWR was'a unique condition for that particular plant design. That unique condition will not occur in the: ABWR design.

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4.13.9 The automatic flow control range shall be from 701 to 100% rated power (1001 rod line).

4.13.10 The minimum rip speed shall be greater than or equal to 450 RPM.

4.14 Core Flow Measurement Reauirements 4.14.1 Core flow measurement shall be provided to deliver inputs for scram trip as shown in Figures 1.

d J.

4.14.2

_ae required measurement accuracy shall be within the requirements specified in Section 2.1.2.c.

4.14.3 The design basis maximum sensor response time shall be less than or equal to 0.25 second.

(Analysis condition for E/PA - 3.0 second) 4.15 Ett p ate. Reouirements 4.15.1 Trip of noin fasdwater pumps shall be initiated upon the condition of high vessel water level (Level 8).

This function may be designed as a non safety related trip.

However, the design of this trip function shall be highly reliable.

4.15.2 Th6 trip signal shall be the same signal to be supplied for the high vessel water level turbine trip (see Section 4.10).

4.15.

Kg/cm)g(1065psig)shallbelessthanorequalto130percentofrated.The ma The changeof{lowbelowthepressurespecifiedaboveshallbelessthan2,8%

flow /Kg/cm (0.2% flow / psi). E/P analysis may take credit of the maximum flow limit (1101) imposed by the feedwater control system.

4.15.4 Following a trip of one main feedvater pump, the minimum feedvater available to the vessel shall be greater than or equal to 75% of rated.

4.15.5 A six-heater feedvater heating system shall be designed to provide at least 215.5*C (420'F) tydwater at the rated condition.

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M 4.15.6 No single opentor error or equipment failure shall cause loss of more _\\

than 55'c (100'F) feedvater heating.

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m ws 4.15.7 The la (standard deviation) uncertainty for the feedwater flow measurement system shall be less chan or equal to 1.76% of rated feedwater flow.

4.16 Auxiliary Varer Makeue Recuirements 4.16.1 The Reactor Core Isolation Cooling (RCIC) system shall be initiated upon the condition of low vessel water level (Level 2).

NEO 204 (REV SIM)

3E@mu To ITWV\\ Ik.

Issue 16 Capability of RCIC/RHR Systems to mitigate ATWS (Containment Response for ATWS with Failure of Reactivity Control)

About October 17, the NRC fonvarded the following question:

"During the GE presentation to the staff on the ABWR PRA on August 6, 1991, GE referred to an INEL analysis which showed that RCIC was capable of preventing core damage. INEL performed the analysis of a high pressure ATWS with very low makeup flow to support GE's PRA assessment of the ABWR during degraded conditions. (Ref. DOE /ID 10211, October 1988.) The conclusion of the analysis was that based upon.

a constant vessel superheat of 175K, the equivalent of 3.45 heat exchangers are necessary to keep the peak containment pressure below the design pressure. Confirm that the three heat exchangers as presently designed having (sic.) sufficient heat removal capacity to mitigate ATWS."

ference report does indeed note that 3.45 heat exchangers would be required to maintain the peak containment pressure below the design pressure. However, this is not the correct limit to use in determining the success criteria for the PRA. An accident in which both the rod insertion and boron injection fails is well beyond the design basis. A more appropriate limit on the containment performance is the factored loads or service level C criterion. For the ABWR senice level C corresponds to a pressure above 80 psig (Ref. 2, section 19E.2,3.2). Tab!c 3 of reference 1 indicates that the peak containment pressure t' sing 3 heat exchangers is 72 psig.

Furthermore, even if the more restrictive criteria were to apply, the time until the containment design pressure is reached is approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This allows ample time for the operators to provide boron injectmn using alternate means, which would reduce the power generation rate wc!! below that which could be removed using three trains of RHR. Therefore, no change is required in the ATWR success criteria.

References:

1. K C. Wagner, " Analysis of a High Pressure ATWS with Veiy Low Make-up Flow", DOE /ID-10211, October 1988.

2. ABWR Standard Safety Analysis Report.

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