ML19347F529

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Testimony of Ma Ross Re Doherty Contention 48,control Rod Drive Return Line.Prof Qualifications Encl.Related Correspondence
ML19347F529
Person / Time
Site: Allens Creek File:Houston Lighting and Power Company icon.png
Issue date: 05/11/1981
From: Ross M
GENERAL ELECTRIC CO.
To:
Shared Package
ML19347F516 List:
References
NUDOCS 8105190536
Download: ML19347F529 (10)
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l ) RELATED. CORRESPONDE522.1-8 c) ry 1 UNITED STATES OF AMERICA q 3cew $'-

NUCLEAR REGULATORY COMMISSION mnE 2 - 2 BEFORE THE ATOMIC SAFETY AND LICENSING BOAD [",,1Y 13198) > -4 ,

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Office of theSec.

In the Matter of )

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Docket No. 50-466 D y 6'

5 HOUSTON LIGHTING & POWER COMPANY)

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6 (Allens Creek Nuclear Generating)

Station, Unit No. 1) )

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8 TESTIMONY OF MONTY A. ROSS ON BEHALF OF HOUSTON LIGHTING & POWER CO. ON DOHERTY CONTENTION 48 -

9 CONTROL ROD DRIVE RETURN LINE 10 Q. Please state your name and place of employment.

11 A. My name is Monty Ross and I am employed as Manager of 12 the Data Acquisition and Operator System Unit at the General 13 Electric Company.

l 14 Q. Please describe your professional qualifications.

15 A. My professional qualifications are set forth in Exhibit 16 MAR-1 to this testimony.

17 Q. What is the purpose of your testimony?

l 18 A. The purpose of my testimony is to address Mr. Doherty's 19 Contention 48, which alleges:

20 ACNGS should be designed with a control rod drive return line (CRDRL) , because this source of high 21 pressure water functions as an additional safe guard against events where there is water loss 22 f rom the reactor vessel yet pressure remains high.

23 Intervenor relies on incidents which occurred at three BWR 24 facilities: Drowns Ferry, Dresden IT, and Oyster Crech.

25 In these three incidents, the Control Rod Drive (ChD) 26 system, in its normal mode of operation, provided a small 27 amount (4100 gpm) of reactor coolant makeup water by pumping 28 through the CRD return line and control rod drivua themselves.

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m j 1 2- The most significant of these incidents was the Browns Ferry 3 I fire in that the demands placed on the CRD system were 4 greater than for the other two incidents. For this reason 5 this testimony will address primarily the Browns Ferry incident.

6l 7 Intervenor also states that CRD return line offers a 8 backup system in the event the HPCS pumps are out of 9 service.

10 o. As a preliminary matter, would you briefly describe the 11 Control Rod Drive System?

12 A. The CRD system functions to control overall reactor 13 power level and provides the principal means of quickly and l

14 safely shutting dow the reactor. The power level of the 15 reactor is maintained by use of moveable control blades, one i

16 cruciform control blade per four fuel bundles, throughout 17 the core. The blades are moved vertically by a hydraulically 18 actuated, locking piston type drive mechanism. The drive 19 mechanisms perform both a positioning and latching function, 20 and a scram (reactor shutdown) function with the latter 21 overriding any other signal. The CRD pump supplies pressurized 22 demineralized water from either the condensate demineralizer 23 via the condensate booster pumps (primary source) or the 24 condensate storage tank (secondary source) to the CRD system 25 to provide hydraulic operating pressure and cooling water 26 for the drive mechanisms. Two CRD pumps are provided, one 27 being a backup for the other, however, the two pumps are 28 capable of operating simultaneously. Under normal operation, l

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1 2 one CRD pump operates to p ovide cooling and drive water 3 to the CRD's. This water is discharged through the CRD's 4 into the reactor vessel. Thus, the CRD system does possess 5 a limited capacity to provide makeup water to the reactor 6 vessel. In safety analyses, no credit is taken for the CRD 7 system's capacity to provide makeup water since it is not a 8 ' part of the Emergency Core Cooling Systen (ECCS).

9 Q. What was the function of the CRD return line in the 10 CRD system?

11 A. The CRD return line (CRDRL) was designed to return, 12 to the reactor vessel, water in excess of system require-13 ments.

14 Q. Why was the CRDRL eliminated?

15 A. As early as 1974, a GE task force investigating 16 cracking in austenitic stainless steel piping measured 17 unexpectedly high top to bottom thermal gradients in the 18 CRDRL nozzle. This is due to the fact that CRDRL water l 19 (condensate water) is not heated and is typically 150 F 20 or less. Also, flow through the CRDRL nozzle varies 21 intermittently between low flow and high flow surges 22 depending upon CRD movemeat. Subsequent exaninations at 23 Bwa stations revealed extensive nozzle cracking; not l 24 only was cracking uiscovered in the CRDRL nozzle, but l 25 also on the wall of the reactor vessel beneath the nozzle.

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26 This is attributed to low flow conditions where return 27 water f1t is down the reactor vessel wall. The GE study 28 i

i 1 2 of CRD return line nozzle cracking resulted in a series 3 of recommendations, among which were to " cut and cap" 4 the line and nozzle on existing plants, and to eliminate 5 the line completely on new plants.

6 Q. What was the basis for the recommendation to eliminate 7 the CRD return line?

8 A. The recommendation to remove the return line was 9- based on the need to prevent nozzle cracking, and on GE's 10 determination that the line had never been necessary to 11 attain an acceptable CRD system performance. The CRD 12 Hydraulic System must be operated with a constant 13 differential pressure (reference pressure) above reactor 14 pressure. This reference pressure can be obtained by 15 system adjustments during system operation. GE performed 16 an evaluation of the CRD return line problem and submitted 17 test data and analyses to the NRC Staff for review. The l

18 NRC Staff concurred with GE's recommended design change 19 as an acceptahlc long term solution to the problem. This 20 analysis includes the BWR/6-238" design used at Allens 21 Creek (See NUREG-0619).

22 Q. Does the A13 ens Creek design have a CRDhL?

23 A. No, the Allens Creek pressure vessel has a CRDRL 24 nozzle since it was fabricated before the CRDRL nozzle 25 cracking problem was discovered. However, Allens Creek 26 will not have a CRDRL and the nozzle will be capped.

27 Q. How is water returned to the reactor vessel with the 28 CRDRL omitted?

1 2 A. A Hydraulic Control Unit (HCU) combines all operating 3 valves and components needed for the normal positioning or 4 scram of a single control rod. The HCU's are all inter-5 connected to the CRD Hydraulic System header piping which 6 functions to provide the pressures and flows necessary to 7 insert, withdraw and provide cooling water to the CRD's.

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Exhaust water from a moving CRD is disbursed to the 9 non-moving drives via a reverse flow from the CRD Hydraulic 10 System exhaust header through the insert exhaust solenoid 11- valves (Valve tag no. 121) on the HUC of the non-moving CRD.

12 A small exhaust water header back pressure resulting from 13 drive movement lifts the solenoid valve pistons of the 14 adjacent non-moving CRD and permits exhaust flow to be 15 discharged through the non-moving CRD seal mechanisms and j 16 into the reactor. CRD seal mechanisms are designed to 1

17 function with small bypass flow to the reactor vessel.

18 Thus, contrary to Intervenor's assertion, the CRD system 19 without the return line does have the capability of providing 20 an additional source of reactor coolant makeup water.

21 Q. How does this affect the operability of the CRD 22 system?

l 23 A. As previously discussed, CRD system pressure adjustment can be modified as necessary by reactor operators. Analyses 24 i

25 by GE hnve shown that the capacity of the CRD system to 26 provide makeup water in a Browns Ferry I type incident is 27 adequate to prevent the core from uncovering. Initially, i

28 there was concern about the continued long-term operability l

1 2l< of the insert exhaust directional control valve since 3 it would be required to accommodate reverse flow for which 4 it was not designed. In response to this concern, GE tested 5 ten valves which had been removed from an operating reactor on which the CRDRL has been valved out for six months.

6l 7 These valves were then compared with tests performed on five 8 new valves. The results showed that the reverse flow 9- characteristics of all valves were similar and that 10 degradation of the valves to the point of causing system 11 malfunction would not be expected during long term normal 12 operation of the system. Also, a simulated 40-year life 13 cycle test of five directional control valves (valve tag 14 no. 121) observed no valve functional failures. No adverse 15 effects on the test valves were observed as a result of the In 16 system backflow model and scram times did not increase.

17 addition, effects on the test valves were observed as a 18 result of the system backflow model and scram times did not 19 increase. In addition to these tests, some modifications 20 have been incorporated into the CRD Hydraulic System as a l 21 result of the reverse flow through the solenoid valves (Valve 22 tag no. 121). These include equalizing valves and flush 23 ports. These improvements will ensure that the CRD system 24 witaout the CRDRL will perform as well or better than 25 sysuems with return lines.

26 0 Can the CBD system provide sufficient nakeup water to 27 tha reactor vessel to k6.3p the core covered, if necessary?

E8 7.. GE performed an analytical comparison of CRD system injection capability for various BWR designs before and

1 after deletion of the CRD return line using the Browns Ferry i 2l' 3 I fire as a model. The calculations utilized a base case 4 set of conditions that existed during the 1975 Browns Ferry fire, which placed the most severe demands on the CRD system 5p I During that incident, a normal water 6 experienced to date.

f level was maintained above the core (by other high pressure 7

8 systems) until 40 minutes after shutdown. Reactor pressure 9 then increased to the set pressure of the lowest set: point 10 safety /reli f valve setting and concurrently all sources of 11 water other than the CRD system were lost. Under those 12 conditions the flow necessary to keep the core from uncovering _

13 was calculated and compared for CRD systems with and without 14 the CRD return line and with one or two pump operation. The 15 calculations were weighted to maximize the amount of water 16 needed to keep the core covered, and to emphasize any apparent 17 change due to elimination of the return line. The results 18 of GE's c>mparison for a 238" BWR/6, such as ACNGS, show 19 that the injection flow rate for the CRD system provides l 20 enough makeup flow to keep the core covered under the conditions 21 of the Browns Ferry I fire.

22 Q. Intervenor also alleges that the CRD system is the only 23 source of high pressure makeup water when the High Pressure 24 Core Spray System (HPCS) is out of service. Is the CRD 25 system the only high pressure water source if the HPCS is 26 inoperable?

27 A. No, the Reactor Core Isolation System (RCIC) is a high 28 pressure injection system which is powered by extracting

1 2 steam from the reactor pressure vessel. This enables the 3 RCIC system to operate under conditions where there is loss 4 of offsite AC power. It is possible that the reactor could 5 operate without the HPCS being operable. However, certain 6 conditions must bc met before this is allowed. First, the 7 RCIC System, the Automatic Depressurization System (ADS),

8" the ',ow Pressure Core Spray (LPCS) System, and the Low 9 Precsure Coolant Injection (LPCI) mode of the Residual Heat 10- Removal System must be operable. Then the HPCS can be 11 inoperable for only fourteen days. Thus, the RCIC system 12 would provide high pressure coolant injection. Should the 13 RCIC system fail to maintain an adequate water inventory, 14 the Automatic Depressurization System would be initiated to 15 reduce reactor pressure so that the Low Pressure Core Spray 16 or Low Pressure Coolant Injcction systems could inject 17 water.

18 Q. What are your conclusions?

19 A. The CRDRL deletion enhances the safety and operacility 20 of ACNGC. The capacity for the CRD System to perform its 21 functions under normal and accident conditions is not 22 impaired. The CRD system retains the capability to provide l 23 an additional source of high pressure reactor coolant makeup.

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1 Exhibit MAR-1 2 EDUCATION AND PROFESSIONAL QUALIFICATIONS 3 Monty A. Ross 4 Mr. Ross is a manager in the Nuclear Steam Supply 5 Systems design organization of the General Electric Nuclear 6 Energy Business Group, in San Jose, California. His employ-7 ment with General Electric began in 1972, as an Engineer 8 in the Design Engineering section, where he worked on the 9 design and analyses of pressure vessel ccmponents, nuclear 10 piping systems, refueling and servicing tools.

i 11 Starting in 1975, Mr. Ross participated in a career 12 developing program of rotating assignments. Major activities l 13 while on this program included the experimental testing of 14 primary containment designs in the evaluation of the thermo-15 dynamic transients which may (hypothetically) occur within l

16 the primary containment as a result of a LOCA and non-LOCA l

l 17 events.

18 In February 1979, he took the position of Lead System 19 Engineer (LSE) for the Rod Control System. As the LSE, 20 he was responsible for the design definition of the Rod 21 Control System. Major tasks in this position included 22 gaining NRC acceptance of the Control Rod Drive System return 23 line removal and directing the evaluation and design changes l 24 resulting from the Browns Ferry 3 partial scram insertion 25 of June 28, 1980. In October 1980, Mr. Ross assumed his 26 present position as a manager in the Nuclear Steam Supply 27 System design organization. The group that he manages is 28 responsible for the design definition of six (6) BUR Standard l

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1 2' Plant sy;; ems including the Rod Control System.

8 Mr. Ross is a 1972 graduate of the University of 4 California at Davis, with a BS Degree in Mechanical Engineering 5 (power generation option) and in Material Science. In 1977, 6 he received an MS Degree in Mechanical Engineering from the 7 University of Santa Clara. Mr. Ross is a registered pro-8 fessional Engineer in the State of California.

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