ML19332B237

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Testimony in Response to Ucs Contention 8 & Environ Coalition on Nuclear Power Contention 1(e) Re Addl LOCA Analysis.Describes Purpose,Assumptions & Results of Analyses.Prof Qualifications Encl.Related Correspondence
ML19332B237
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 09/15/1980
From: Broughton T, Rosalyn Jones
METROPOLITAN EDISON CO.
To:
Shared Package
ML19332B231 List:
References
ISSUANCES-SP, NUDOCS 8009260322
Download: ML19332B237 (27)


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UNITED STATES OF t.MERICA N NUCLEAR REGULATORY COMMISSION 1s te

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1 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i

In the Matter of )

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4 METROPOLITAN EDISON COMPANY ) Docket No. 50-289

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(Three Mile Island Nuclear )

Station , Unit No. 1) )

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l LICENSEE'S TESTIMONY OF ROBERT C. JONES, JR., AND T. GARY BROUGHTON IN RESPONSE TO UCS CONTENTION NO. 8 AND ECNP CONTENTION NO. 1(e)

(ADDITIONAL LOCA ANALYSIS) l l

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OUTLINE The purposes and objectives of this testimony are to respond to UCS Contention 8, which asserts that adequate small-break loss of coolant accident (LOCA) analyses have not been performed, and to respond to related Board questions. The testimony addresses the small-break LOCA analyses performed prior to the TMI-2 accident and their conformance to 10 CFR Part 50, Section 50.46. The purpose, assumptions and results of small-break analyses subsequent to the TMI-2 accident are described. Operating guidelines and procedures for small-break LOCA mitigation are discussed. It is shown that adequate protection for small-break LOCA's is provided.

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INDEX INTRODUCTION .............................................. 1 RESPONSE TO UCS CONTENTION NO. 8 .......................... 2 TABLE SUMMARIZING PRE-TMI-2 LOCA EVALUATIONS .............. 12 TABLES SUMMARIZING LOCA ANALYSES RESULTS .................. 13 1

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INTRODUCTION This testimony, by Mr. Robert C. Jones, Jr., Supervisory Engineer , ECCS Analysis Unit, Babcock & Wilcox Company, and Mr.

T. Gary Broughton, GPU Control and Safety Analysis Manager, is addressed to the following contention:

UCS CONTENTION NO. 8 10 CFR 50.46 requires analysis of ECCS performance "for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the entire spectrum of postulatec' loss-of-coolant accidents is covered." For the spectrum of LOCA's, specific perameters are not to be exceeded. At TMI, certain of these were exceeded. For ex ample ,

the peak cladding temperature exceeded 2200' fahrenheit (50.46(b)(1)), and more than 1% of the cladding reacted with water or steam to produce hydrogen (50.46(b)(3 )) . The measures proposed by the staff address primarily the very specific case of a stuck-open power operated relief valve. However, any other sm could lead to the same consequences.all Additional LOCA ,

analyses to show that there is adequate protec-  !

tion for the entire spectrum of small break  !

locations have not been performed. Ther e for e ,

there is no basis for finding compliance with 10 i 1

CFR 50.46 and GDC 35. None of the corrective actions to date have fully addressed the demonstrated inadequacy of protection against small LOCA's.

ECNP Contention 1(e) was accepted only to the extent that ECNP was permitted to adopt UCS Contention 8. Consequen tly, ECNP Contention 1(e) is not quoted here. (See, Board Memorandum and Order, September 8, 1980.) UCS withdrew its

sponsorship of UCS Contention 8, which has been adopted as a Board Question (See, Board Memorandum and Order of Prehearing Conference of August 12-13, 1980, dated August 20, 1980).

RESPONSE TO UCS CONTENTION NO. 8 BY WITNESS JONES :

UCS Contention 8 asserts that analyses to demonstrate conformance with 10 CFR Part 50, Section 50.46 (10 CFR 50.46) l for the entire spectrum of small-break loss of coolant accident j (LOCA) locations have not been performed. Additionally, it is stated that none of the corrective measures being implemented for TMI-l assure that adequate protection is provided for small-break LOCA 's. Contrary to the contention, compliance with 10 CFR 50.46 has been demonstrated and adequate protection for small-break LOCA's is provided.

Prior to the TMI-2 accident, small-break LOCA evaluations had been performed to verify conformance of TMI-l to 10 CFR 50.46. In order to perform these analysec, the break location which imposes the most severe requirements on the ECCS was identified. As a result of this identification, an analysis was performed of the core flood line break, which results in only one core flood tank and one high pressure injection train available to mitigate the accident under the worst single failure assumption. Also, an analysis of a spectrum of breaks

in the reactor coolant pump discharge piping was performed, as this location results in the loss of a portion of the high pressure injection fluid. These analjaes were performed using the B&W ECCS evaluacion model which has been approved by the NRC as meeting the requirements of Appendix K to 10 CFR Part

50. The actual analyses which were performed are contained in References 1 and 2, and are summarized in Table 1. For the worst case break, the peak cladding temperature was less than 1100*F and no metal-water reaction nor cladding rupture were calculated to occur. Therefore, conformance to 10 CFR 50.46 was demonstrated.

The analysis performed prior to the TMI-2 accident assumed the use of only safety-grade equipment for accident mitigation, and assumed no mitigating operator actions within ten minutes of the initiating event, except as follows:

o Emergency feedwater was assumed to be available, o Operator action to cross-connect the High Pressure Injection System (HPI) was determined to be required in the event of a small break in the reactor coolant pump discharge piping and the postulated failure of the HPI train which discharges into the unbroken coolant loop.

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BY WITNESS BROUGHTON:

With regard to the first of the above items, Licensee's testimony on the Emergency Feedwater System (in response to Licensing Board Question No. 6) will address the reliability of the Emergency Feedwater System (EFW).

BY WITNESS JONES:

In the event of a loss of all feedwater following the EFW upgrade, the feed and bleed mode of emergency cooling is available for LOCA mitigation. See Licensee's testimony on Natural and Forced Circulation (in response to UCS Contentions Nos. 1 and 2).

l BY WITNESS BROUGHTON.

With regard to the second of the above items, modifica-tions to the high pressure injection lines have been made to add cross connections and flow limiting devices to ensure sufficient flow without operator action (See, TMI-l Restart Report, Supplement 1, Part 3, responses to questions 1, 2 and 3).

BY WITNESS JONES:

Subsequent to the TMI-2 accidenc', additional small-break  ;

LOCA analyses were performed. In light of the fact that the severity of the TMI-2 accident was aggravated by operator l

actions, the purpose of these analyses was to provide an improved analytical basis for emergency operating procedures for small-break LOCA's, not to demonstrate compliance with 10 CFR 50.46. A description of the events analyzed, the key assumptions, and the results of the evaluations are provided in Tables 2 through 8. The analyses performed included an extension of the lower end of the break spectrum previously analyzed, an assessment of the effect of failures in the feedwater system, and an assessment of small-break LOCA's with delayed reactor coolant pump trip. From these analyses it was concluded that multiple failures must occur before a LOCA scenario can result in a challenge to 10 CFR 50.46 limits. A summary of the results of the analyses is also provided below:

o In the event of a loss of all feedwater, without a small-break LOCA, operator action within twenty minutes to either establish emergency feedwater or manually actuate high pressure injection assures that the core remains covered, thus assuring adequate core cooling. (See Table 2.)

o In the event of a small-break LOCA with loss of all feedwater, ECCS may not be automatically actuated. For this situ-ation, operator action within twenty minutes to either establish emergency feedwater flow (which will in turn result in automatic ECCS actuation), or to manually actuate high pressure injection, assures that the core remains covered, thus assuring adequate core cooling. (See Table 3.)

o In the event of a loss of main feedwater followed by the pressurizer power operated relief valve (PORV) opening and failing to close, the automatic actuation of high pressure injection is cufficient to assure adequate core cooling. (See Table 4.)

o In the event of the pressurizer PORV opening and failing to close, followed by the loss of all feedwater , the automatic actuation of high pressure ir.jection is sufficient to assure adequate core cooling.

(See Table 5.)

o For certain very small breaks (between 2

0.005 ft and 0.01 ft2) which cause a loss of coolant inventory at a rate in excess of the capacity of high pressure injection, the steam generators would normally be utilized to remove a portion of the energy added to the primary system fluid by core decay heat. During the transition from natural circulation to the boiler-condenser o

mode of cooling (see Licensee's testimony on Natural and Forced Circulation in response to UCS Contention No. 1), an interruption of the energy removal process from the primary system will occur due to i

void formation in the hot legs, and the primary system pressure will increase.

However, the subsequent establishment of steam condensation by the steam generators as i heat removal mechanism controls the repressurization and assures that the core remains covered, thus assuring adequate core cooling. (See Table 6.)

o If the reactor coolant pumps operate continuously throughout the LOCA transient, or are tripped promptly upon receipt of a low reactor coolant pressure safety signal, adequate core cooling is provided for all break sizes. For certain break sizes (between 0.025 ft 2 and 0.2 ft2), adequate core cooling has not been demonstrated if the reactor coolant pumps remain in operation and are subsequently tripped at certain times in the transient. Therefore, in order to assure adequate core cooling, the reactor coolant pumps should be tripped i

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l promptly following automatic initiation of high pressure injection. (See Table 7.)

o In the event of a very small LOCA with loss of all feedwater, system repressurization may actuate the PORV which can subsequently stick open. For this situation, operator action within twenty minutes to either establish emergency feedwater flow (which will in turn result in automatic ECCS actuation) or to manually actuate high pressure injection assures that the core remains covered, thus assuring adequate core cooling. (See Table 8. )

Similar to the pre-TMI-2 analyses, the analyses performed after the accident assumed the use of only safety-grade equipment for accident mitigation, and assumed no mitigating operator actions within ten minutes of the initiating event, except for the two items previously identified (at page 3 abcve) and the manual action of tripping of the reactor coolant pumps following automatic initiation of high pressure injec-tion. l The system behavior which results in the instruction for pump trip involves an extended loss of inventory due to continuous operation of the reactor coolant pumps. While l

continued pump operation provides forced circulation cooling of the core, it also causes more fluid inventory to be discharged out the break than would otherwise occur for certain break sizes. As a result of this increased loss of inventory, the fluid in the Reactor Coolant System will evolve to a high void fraction. If the pumps are tripped after a high void fraction is reached, the available water in the Reactor Coolant System would not be sufficient to keep the core covered. If the core is significantly uncovered, the cladding temperature would begin to increase and the ECCS may not provide, under these conditions, reflooding of the core at a rate which assures that cladding temperatures are maintained within the criteria of 10 CFR 50.46. Since all analyses have confirmed that the plant can be maintained in a safe condition (as defined by 10 CFR 50.46) during a small-break LOCA without the reactor ecolant pumps operating during the transient, provision for prompt tripping of the pumpt upon indication of a LOCA (receipt of a low reactor coolant pressure safety injection signal) assures that adequate core cooling is provided. While other, non-LOCA events may lead to a low pressure safety signal, tripping of the reactor coolant pumps for these events still allows adequate core cooling to be provided. .

BY WITNESS BROUGHTON:

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The generic analyses performed by B&W are applicable to TMI-1. The low pressure reactor trip setpoint has been raised to 1900 psig and the Engineered Safety Features Actuation l

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System (ESFAS) setpoint has been raised to 1600 psig, the values assumed in the generic analyses. (See TMI-l Restart Report sections 11. 2.11 and 11. 2.12) .

BY WITNESS JONES:

Based upon the analyses described above, B&W has also developed operator guidelines for managing small-break LOCA's.

These guidelines contain two parts: Part I provides the guidelines which define operator actions during a small-break LOCA; Part II provides a description of plant behavior during a small-break LOCA and discusses the effect of the operator actions given in Part I.

BY WITNESS BROUGHTON:

TMI-1 procedures have subsequently been developed to implement these guidelines. The TMI-l Emergency Procedures which implement the B&W Loss of Coolant Accident guidelines place strong emphasis on maintaining reactor coolant system i pressure-temperature relationships to assure that a subcooling condition of at least 50*F exists. Specifically, procedures require that upon automatic initiation of high pressure injection, flow shall not be reduced unless: (1) low pressure injection pumps are in operation and flowing at a rate of not less than one thousand gallons per minute each and the situ-ation has been stable for 20 minutes; or (2) all hot and cold leg temperatures are at least 50*F below the saturation j

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temperature for the existing reactor coolant system pressure and the flow reduction is necessary either to prevent pressurizer level from going off scale high or to avoid excessive reactor vessel pressure /downcomer temperature limits.

If 50*F subcooling cannot be maintained, the procedure requires the high pressure injection system to be reactivated. In situations where high pressure injection is manually initiated, flow reductions are permitted only if reactor coolant system pressure is above 1600 psig and the 50*F subcooling margin exists and can be maintained, or if the criteria for flow reductions following automatic initiation are satisfied.

BY WITNESS JONES:

In summary, extensive small-break analyses have been performed for the TMI-l facility. These analyses demonstrate l

that small LOCA's can be mitigated within the criteria of 10 l CFR 50.46. Also, additional small-break analyses have been performed 1.2 order to develop improved emergency procedures.

Thus, contrary to the contention adequate protection for small LOCA's has been demonstrated and is provided.

References

1. Topical Report, BAW-10103A, Rev. 3, "ECCS Analysis of B&W's 177-FA Lowered Loop NSS," July 1977.
2. Letter, J. H. Taylor (B&W) to S. A. Varga (NRC), July 18, 1978.

TABLE 1 PRE-TMI-2 LOCA EVALUATIONS Topical Report BAW-10103A, Rev. 3 o Core Flood Tank Line Break o 2 0.5 ft Reactor Coolant Pump Discharge Piping Break o 2 0.04 ft Reactor Coolant Pump Suction Piping Break Letter Report, J. H. Taylor (B&W) to S. A. Varga (NRC), July 18, 1978 o 2 0.15 ft Reactor Coolant Pum'. Discharge Piping Break o 0.10 ft 2 Reactor Coolant Pump Discharge Piping Break o 2 0.085 ft Reactor Coolant Pump Discharge Piping Break o 0.07 ft Reactor Coolant Pump Discharge Piping Break o 2 0.055 ft Reactor Coolant Pump Discharge Piping Break o 2 0.04 ft Reactor Coolant Pump Discharge Piping Break l I TABLE 2 LOSS OF ALL FEEDWATER WITHOUT SMALL-BREAK LOCA Sequence of Events and Assumptions o Loss of main feedwater occurs.

o Direct trip on loss of feedwater fails and reactor trips on high reactor coolant pressure.

o Loss of offsite power occurs coincident with reactor trip.

o Emergency feedwater is not provided to steam generators, o Reactor coolant pressure continues to increase, o Pressurizer PORV does not open.  !

o Pressurizer safety valves open. )

o Core decay heat is 1.0 times the ANS standard value.

o Single failure occurs in the High Pressure Injection System.

Summary of Results o Operator action within twenty minutes to initiate emergency feedwater will lower reactor coolant pressure and terminate loss of reactor coolant inventory, assuring adequate core cooling; or o Operator action within twenty minutes to activate high pressure injection provides sufficient reactor coolant inventory to assure adequate core cooling.

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TABLE 3 SMALL-BREAK LOCA WITH LOSS OF ALL FEEDWATER  !

Sequence of Events and Assumptions o Small-break LOCA occurs.

o Reactor trip occurs on low reactor coolant pressure.

o Loss of offsite power and loss of main feedwater occur coincident with reactor trip.

o Emergency feedwater is not provided to steam generators.

o Core decay heat is 1.2 times the ANS standard value, o Both high pressurw injection trains function.

Summarv of Results o For break sizes greater than 0.01 ft 2 emergency core i

cooling is automatically initiated anc no operator action is required to assure adequate core cooling.

o For break sizes equal to or less than 0.01 ft2 the setpoint for initiation of high pressure injection is not reached. Operator action within twenty minutes to initiate emergency feedwater (which will subse-quently result in high pressure injection) or to actuate high pressure injection will assure adequate core cooling.

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TABLE 4 LOSS OF MAIN FEEDWATER WITH PORV FAILURE Sequence of Events and Assumptions o Loss of main feedwater occurs.

o Direct trip on loss of feedwater fails and reactor coolant pressure increases.

o Pressurizer PORV opens and does not close.

o Reactor trip occurs on high reactor coolant pressure.

o Emergency feedwater is provided to steam generators.

o Core decay heat is 1.2 times the ANS standard value, o Single failure occurs in the High Pressure Injection System.

Summary of Results.

o Automatic actuation of high pressure injection provides sufficient reactor coolant inventory to assure adequate core cooling.

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l TABLE 5 PORV FAILURES FOLLOWED BY LOSS OF ALL FEECWATER Sequence of Events and Assumptions o Pressurizer PORV fails open and does not close.

l o Reactor trip occurs on low reactor coolant pressure.

o Loss of offsite power and loss of main feedwater occur coincident with reactor trip.

o Emergency feedwater is not provided to steam generators.

o Core decay heat is 1.0 times the ANS standard value.

o Single failure occurs in the High Pressure Injection System.

Summary of Results o Automatic actuation of high pressure iniection provides sufficient reactor coolant inventory to assure adequate core cooling.

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TABLE 6 VERY SMALL LOCA WITH LOSS OF MAIN FEEDWATER Sequence of Events and Assumptions o Very small-break LOCA (0.005 - 0.01 ft2) occurs.

o Reactor trips on low reactor coolant pressure.

o Loss of offsite power and loss of main feedwater occur coincident with reactor trip.

o Emergency feedwater is provided to steam generators.

o Core decay heat is 1.2 times the ANS standard value.

o Single failure occurs in the High Pressure Injection System.

Summary of Results o Natural circulation initially removes core decay heat, then is interrupted as reactor coolant inventory decreases.

o Reactor coolant pressure increases when natural circulation is interrupted, then is stablized by steam condensation in the steam generators.

o Automatic actuation of high pressure injection provides sufficient reacter coolant inventory to assure adequate core cooling.

TABLE 7 SMALL-BREAK LOCA WITH DELAYED REACTOR COOLANT PUMP TRIP Sequence of Events and Assumptions o small-break LOCA between 0.025 f t 2 and 0.2 ft occurs.

o Reactor trips on low reactor coolant pressure.

o Reactor coolant pumps initially continue to operate, then are tripped at a later time during the accident.

o Core decay heat is 1.2 times the ANS standard value.

o Both high pressure injection trains function.

Summary of Results o If the re&ctor coolant pumps continue to operate, adequate core cooling is assured.

o If the reactor coolant pumps trip after a high system void fraction is reached, adequate core cooling has not been demonstrated, o If the reactor coolant pumps are tripped promptly upcn automatic initiation of high pressure injection, adequate core cooling is assured.

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TABLE 8 SMALL-BREAK LOCA WITH LOSS OF ALL FEEDWATLR AND SUBSEQUENT PORV FAILURE Secuence of Events and Assumptions o Very small-break LOCA (0.01 ft2) occurs.

o Reactor trips on low reactor coolant pressure.

o Loss of offsite power and loss of main feedwater occur coincident with reactor trip.

o Emergency feedwater is not provided to steam generators.

o Core decay heat is 1.2 times the ANS standard value.

o Both high pressure injection trains function.

o Reactor Coolant System repressurization results in PORV opening and remaining open.

Summary of Resuyg o Operai i ction within twenty minutes to initiate emergency feedwater (which will subsequently result in high pressure injection) or to actuate high pressure injection provides sufficient reactor coolant inventory to assure adequate core cooling. {

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i T. GARY BROUGHTON 1

Business Address: GPU Service Corporation 100 Interpace Parkway Parsippany, New Jersey 07054 Education: B.A., Mathematics, Dartmouth College, 1966.

Experience: Control and Safety Analysis Manager, GPU Service Corporation, 1978 to present. Responsible for nuclear safety analysis and integrated thermal, hydraulic and control system j analysis of nuclear and fossil plants.

Supervised on-site technical support groups at Three Mile Island, Unit 2 during the post-accident period.

Safety and Licensing Engineer; Safety and Licensing Manager, GPU Service Corporation, 1976 to 1978. Performed and supervised nuclear licensing, environmental licensing and safety analysis for Oyster Creek, Three Mile Island and Forked River plants.

Served as Technical Secretary to Oyster Creek and Three Mile Island General Office Review Boards.

Officer, U.S'. Navy, 1966 to 1976.

Trained at Naval Nuclear Power School, Prototype and Submarine School.

Positions held include Nuclear Propulsion Plant Watch Supervisor, Instructor at DlG prototype plant and i

Engineering Officer aboard a fast-attack nuclear submarine.

Publications: EPRI CCM-5, RETRAN - A Program for One-Dimensional Transient Thermal-Hy-draulic Analyses of Complex Fluid Flow Systems, Volume 4: Applications, December, 1978, Section 6.1, " Analysis cf Rapid Cooldown Transient - Three Mile Island Unit 2", with N.G.

Trikouros and J. F. Harrison.

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"The Use of RETRAN to Evaluate Alternate Accident Scenarios at  ;

TMI-2", with N. G. Trikouros.

Proceedings of the ANS/ ENS Topical Meeting on Thermal Reactor Safety, l April 1980, CONF-800403.  :

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"A Real-Time Method for Analyzing Nuclear Power Plant Transients", with l P.S. Walsh. ANS Transactions, Volume 34 TANSAD 34 1-899 (1980).

OUTLINE ADDITIONAL LOCA ANALYSIS EXHIBITS

! The exhibits listed as Items 3 through 13 on Licensee's Certificate of Service, September 15, 1980, are submitted in response to the Board Question regarding UCS Contention 8, and provide a documentary history of the small break LOCA analyses performed by the Babcock & Wilcox Company which are applicable to TMI-1. The results of these analyses are presented in Licensee's Testimony of Robert C. Jones, Jr. , and T. Gary Broughton in response to UCS Contention No. 8 and ECNP Con-tention No. 1(e) (Additional LOCA Analysis) .

Report BAW-10103A, Rev. 3 (Item 3 on Licensee's Certifi-t cate of Service) , and the July 18, 1978 supplemental analysis (Item 4) constitute a complete spectrum of small break analyses which show conformance to the requirements 10 CFR 50.46 and Appendix K to 10 CFR Part 50.

Section 6 of the May 7, 1979 report, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" (Item 5), Supplements 1, 2 and 3 (Items 6, 7 and 8) thereto, and B&W Documents 86-1103585 ';0 (System Response to Total Loss of SG Heat Sink,) (Item 9) and 86-1117679-000 (Small Break with failed PORV) (Item 13) con-sist of additional analyses of plant response to various small break scenarios, which were performed in response to specific NRC requests following the TMI-2 accident. The results of these analyses demonstrate that, with appropriate operator action, the emergency core cooling system is capable of

l controlling the consequences of these scenarios. B&W Document ,

1 69-1106001-00, "Small Break Operating Guidelines" (Item 12),

provides guidance for operator action based upon the results of the small break analyses. l The evaluations contained in the B&W report, " Analysis Summary in Support of an Early RC Pump Trip," (Item 10) and its " Supplemental Small Break Analysis," (Item 11) were per-formed pursuant to NRC IE Bulletin 79-05C. These evaluations demonstrate that, under highly voided reaction coolant condi- l tions, a delayed trip of the reactor coolant pumps will result in unacceptable consequences when Appendix K evaluation techniques are used. The analysis further shows that the prompt reactor coolant pump trip upon receipt of a low pressure ESFAS signal (as required by these results) will provide acceptable LOCA consequences.

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i ROBERT C. JONES, JR. l l

Business Address: Babcock & Wilcox Company Nuclear Power Generation Division 1 P.O. Box 1260 '

Lynchburg, Virginia 24505 Education: B.S., Nuclear Engineering, Pennsylvania State University, 1971. l Post Graduate Courses in Physics,  ;

Lynchburg College.  ;

1 Experience: '

June 1971-June 1975: Engineer , ECCS Analysis Unit, B&W. Performed both' large and small break ECCS analyses under both the Interim Acceptance l I

Criteria and the present Acceptance Criteria of 10 CFR 50.46 and Appendix K.

June 1975-Present: Acting Supervisory i Engineer and Supervisory Engineer, ECCS Analysis Unit, B&W. Responsible i for calculation of large and small break ECCS evaluations, evaluations of mass and energy releases to the containment during a LOCA, and performance of best estimate pretest predictions of LOCA experiments as part of the NRC Standard Problem Program. Involved in the preparation of operator guidelines for small-break LOCA's and inadequate core cooling mitigation.

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