ML19332B236
| ML19332B236 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/15/1980 |
| From: | Rosalyn Jones, Keaten R METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19332B231 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8009260320 | |
| Download: ML19332B236 (23) | |
Text
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RELkTED CORRBSPONDENCg LIC 9/15/80 O
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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METROPOLITAN EDISON COMPANY
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Docket No. 50-289
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(Restart)
(Three Mile Island Nuclear
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Station, Unit No. 1)
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4 LICENSEE'S TESTIMONY OF N
C ROBERT W.
KEATEN AND ROBERT C.
JONES O
IN RESPONSE TO UCS CONTENTION NOS. I and 2 (NATURAL AND FORCED CIRCULATION)
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OUTLINE The purposes and objectives of this testimony are to respond to UCS Contentions 1 and 2, which assert that natural circulation cooling at TMI is inadequate to remove decay heat, and that reliable forced cooling should be provided by aeans which meet the NRC's regulations for systems important to safety.
The testimony sho es that the TMI-2 accident did not demonstrate that natural circulation is inadequate.
It explains the natural circualtion phenomenon and the veri-fication of its capability for the TMI-l design.
Other reliable means to remove core decay heat following a small-break LOCA are explained, and it is shown that:
(a) operation of reactor coolant pumps are not required to assure adequate core cooling; (b) the residual heat removal system is not required to operate at the design pressure of the primary system; and, (c) that ECCS can be operated in the feed and bleed mode with adequate capacity and shielding for storage and recirculation of the radioactive water.
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1 l
l INDEX i
INTRODUCTION 1
RESPONSE TO UCS CONTENTION NO. 1...........................
2 i
RESPONSE TO UCS CONTENTION No. 2...........................
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INTRODUCTION This testimony, by Mr. Robert W.
Keaten, GPU Manager of Systems Engineering and Mr. Robert C. Jones, Jr., Supervisory Engineer, ECCS Analysis Unit, Babcock & Wilcox Company, is addressed to the following contentions-UCS CONTENTION NO. 1 The accident at Three Mile Island Unit 2 demonstrated that reliance on natural circula-tion to remove decay heat is inadequate.
During the accident, it was necessary to operate at least one reactor coolant pump to provide forced cooling of the fuel.
However, neither the short nor long term measures would provide a reliable method for forced cooling of the reactor in the event of a small loss-of-coolant accident
( " LOC A" ).
This is a threat to health and safety and a violation cf both General Design Criterion
("GDC") 34 and GDC 35 of 10 CFR Part 50, Appendix A.
UCS CONTENTION NO. 2 Using existing equipment at TMI-1, there are only 3 ways of providing forced cooling of the reactor:
- 1) the reactor coolant pumps; 2) the residual heat removal system; and 3) the emergency core cooling system in a " bleed and feed" mode.
None of these methods meets the NRC's regulations applicable to systems impor-tant to safety and is sufficiently reliable to protect public heal th and safety:
a)
The reactor coolant pumps do not have an on-site power supply (GDC 17),
their controls dp not meet IEEE 279 (10 CFR 50.55a(h)) and they a::e not seismically and environmentally qualified /3DC 2 and 4).
b)
The residual heat removal s/ stem is incapable of being utilized at the design pressure of the primary system.
c)
The emergency core cooling system cannot be operated in the bleed and feed mode for the necessary period of time because of inadequate capacity and radiation shielding for the storage of the radioactive water bled from the primary coolant system.
RESPONSE TO UCS CONTENTION NO. 1 BY WITNESS JONES:
UCS Contention 1 expresses the position that the TMI-2 accident demonstrated that during a small-break loss of coolant accident (LOCA) natural circulation is inadequate to remove decay heat and that actions are necessary to provide a reliable method of forced cooling following a small-break LOCA.
Contrary to the contention, the TMI-2 accident did not demonst-rate that natural circulation is inadequate to remove decay Reliable methods of providing emergency core cooling neat.
following a LOCA are currently available at TMI-1.
Current General Design Criterion (GDC) 34 and GDC 35 of Appendix A to 10 CFR Part 50 require that systems be provided to remove core residual heat and to provide emergency core cooling, respectively.
To demonstrate how these requirements are met, this testimony has been structured into two basic parts.
The first part will provide a discussion of the natural circulation phenomena.
As will be demonstrated, if the reactor coolant pumps become inoperative, natural circulation is a j,
reliable means of maintaining core cooling provided that excessive voiding does not occur in the Reactor Coolant System (RCS).
The majority of the accidents in the small-break LOCA spectrum will, however, lead to voiding of the RCS such that j
natural circulation cannot be maintained throughout the i
accident.
Thus, the second part of this testimony will address how adequate core cooling is provided for a small-break LOCA even though natural circulation may not be available.
The discussion on small-break LOCA's is subdivided into two basic areas:
(1) the ability to remove the core decay heat from the fuel rod to the primary system fluid in order to provide adequate core cooling, and (2) the ability to remove j
the energy added to the primary system fluid to assure that the i
design conditions of the RCS pressure boundary are not exceeded.
The testimony will show that adequate core cooling during a small-break LOCA is provided by the Emergency Core Cooling System (ECCS).
In addition, it will be shown that although natural circulation cannot be maintained for all small-break LOCA's, other means are available to remove the energy added to the primary system fluid by the core decay heat.
Thus, adequate core cooling can be provided without the reactor coolant pumps, and it is not necessary to provide forced cooling of the reactor during a small-break LOCA.
Natural circulation is the normal means of providing core cooling for pressurized water reactors when all reactor coolant pumps are inoperative.
The natural circulation phenomenon is.-
an inharent danign feature which is dependent upon the pressure losses of the RCS and upon the pressure gains available due to coolant temperature and density distributJ9n.
To establish natural circulation cooling following loss of all reactor coolant pumps, the steam generators are utilized to
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remove the energy from the reactor coolant and establish a 1
temperature distribution, and thus a density distribution, to promote a positive pressure drop in the RCS.
The cooling accomplished by the steam generators provides a colder (higher density) fluid to the inlet region of the reactor vessel which i
is then heated in the reactor core and returns to the top of the steam generator as a warmer (less dense) fluid - see Figure 1.
As the reactor coolant is cooled in the steam generator the density change provides the gravity-induced force to maintain a
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continuous flow in the system.
Analysis and testing show that the primary system fluid will remain subcooled and a core temperature difference of between 20*F and 40*F will result.
This temperature difference also occurs in the steam 4
generators, and results in a natural circulation flow rate between 2% and 4% of the normal value with all four reactor coolant pumps on.
This is adequate for core decay heat removal.
The natural circulation capability of the TMI-l design nuclear steam supply system has been verified by several means:... _
o Analyses have been performed to determine that natural circulation is adequate to maintain core cooling when required due to all reactor coolant pumps being inopera-tive.
These analyses were performed utilizing conservative assumptions over a wide range of plant conditions.
Natural circulation testing has been o
conducted at B&W operating plants.
The testing confirmed that natural circulation can be initiated and maintained over a wide range of conditions, and demonstrated that the design analyses conservatively predict the natural circulation capabilities of the plants.
t Unplanned occurrences of natural circula-o tion core cooling have been experienced at B&W operating plants which further demon-strate the adequacy of the B&W system under this condition.
In all of these events, in which the reactor coolant pumps were inoperative, natural circulation maintained the plant in a safe condition.
As described above, natural circulation can provide adequate core cooling should the reactor coolant pumps become inoperative.
However, during a small-break LOCA, voids may form in the RCS and prohibit natural circulation.
To under-stand how adequate core cooling is provided for a small-break LOCA, it is necessary to describe the heat removal processes that occur during such an accident.
This involves both the i
energy removal from the core and the energy removal from the primary system.
The energy within the core must be transferred into the i
primary system coolant to assure adequate core cooling.
This is accomplished initially in the transient by the forced circulation cooling inherently provided by the. coastdown of the reactor coolant pumps which remove the stored energy of the l
Following this, as long as the core remains covered by core.
liquid coolant or a two-phase mixture, the cladding temperature will remain within a few degrees of the fluid temperature of the coolant and adequate core cooling will be maintained.
Should the fuel rods become uncovered to a limited extent and/or for a limited period of time, cooling of the uncovered portion of the core is provided by the steam generated within the portion which is covered by a two-phase mixture.
The ECCS is designed to provide the necessary makeup fluid to the primary system to compensate for the loss of coolant and assure that sufficient fluid is maintained within the reactor vessel for adequate core cooling.
The energy added to the primary system must also be removed to prevent. excessive system pressures from occurring...
For the larger-sized small LOCA's, those greater than approxi-2 mately 0.02 ft the energy discharged through the break is sufficient to prevent a pressure increase, whether or not forced or natural circulation occurs.
That is, secondary heat removal is not required.
For breaks smaller than approximately 0.02 f t2, during the period of the transient that the primary system (excluding the pressurizer) remains sufficiently free of voiding, natural circulation flow will be established in the system and the steam generator will remove the added energy, if a secondary heat sink is available.
If primary system voids increase to a volume sufficient to fill the 180* inverted U-bends at the top of both of the reactor coolant system hot legs, the natural circulation process would be interrupted.
However, assuming continued feedwater availability, a boiler-condenser process would then occur.
In this process, steam generated by core decay heat rises through the hot leg and is condensed in the steam generator.
The condensed primary coolant then returns to the core by gravity flow through the cold legs to provide further heat removal.
Should feedwater not be delivered to the steam generators for these smaller LOCA's, heat removal from the primary system can be accomplished by the " feed and bleed" mode of cooling.
this mode, which is a form of forced circulation cooling, In the High Pressure Injection System (HPI) is utilized to " feed" water to the RCS and the pressurizer relief and/or safety !
l
valves " bleed" the water from the system.
In this manner, the inventory injected by the HPI is used to assure the core is covered by liquid coolant or a two-phase mixture and is thus adequately cooled.
The water discharged through the pressurizer relief and/or safety valves removes the energy added to the primary system by the core.
Further discussion of the feed and bleed mode of core cooling is provided in the Licensee's testimony in response to UCS Contention 2.
The analyses which have been performed to demonstrate adequate core cooling during a small-break LOCA are discussed in Licensee's testimony in response to UCS Contention 8.
BY WITNESSES KEATEN AND JONES During the TMI-2 accident, there were periods where adequate core cooling was not maintained.
However, this did i
not occur due to any inherent inability of natural circulation or the other modes described above to remove the decay heat, but rather was due to the fact that HPI was prematurely reduced, resulting in inadequate ECCS injection to keep the fuel rods covered by a two-phase mixture.
After adequate injection flow was restored, and subsequent to the core damage, the core was effectively cooled even though natural circulation was not occurring in the primary system.
The operation of a reactor coolant pump, which was initiated at approximately sixteen hours after the start of the accident, was performed to reestablish a uniform temperature distribution in the primary _
cyotem, by removing voids from the 180' bend in the reactor coolant hot legs, and to establish heat removal via the steam generator.
Approximately one month after the accident, the reactor coolant pump was tripped and since that time natural circulation has provided adequate core cooling even with the core blocksge which is believed to exist.
In summary, and contrary to the contention, the TMI-2 accident did not demonstrate that natural circulation is inadequate to remove decay heat.
Also, contrary to the contention, various reliable means exist to remove core decay heat from the primary system following a small-break LOCA.
RESPONSE TO UCS CONTENTION NO. 2 BY WITNESS JONES:
UCS Contention 2 states that the various methods available for providing forced cooling of the reactor are not suffi-ciently reliable to protect the public health and safety.
Underlying this position, as expressed in UCS Contention 1, is the consideration that natural circulation is inadequate ano that reliable methods do not currently exist to assure adequate core cooling following a small-break loss of coolant accident (LOCA).
However, as shown by Licensee's testimony in response to UCS Contention 1, this concern is not valid.
With regard to the specific methods of providing forced circulation identified in the contention, the first is
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operation of the reactor coolant pumps.
It is ascerted that since the reactor coolant pumps do not meet certain regulations
{
applicable to equipment important to safety, as listed in item a) of the contention, public health and safety are not ade-quately protected.
Contrary to this concern, the reactor coolant pumps at pressurized water reactors have never been classified by the NRC (AEC) as "important to safety" within the meaning of the referenced General Design Criterf.a, except to the extent that the pump casings form part of the reactor coolant pressure boundary.
In fact, current regulations require that adequate core cooling be provided assuming a loss of offsite power.
This results in the tripping of the reactor coolant pumps and analyses have been performed to demonstrate that the required core cooling is assured when forced circula-tion by the reactor coolant pumps is not available -- e.g.,
see Licensee's testimony in response to UCS Contentions 1 and 8.
Therefore, since the reactor coolant pumps are not required to assure adequate core cooling, the regulatory items cited are not applicable to the operation of the pumps as contended.
The second method of providing forced cooling discussed in the contention is che use of the residual heat removal system (Low Pressure Injection System -LPI).
It is stated that, since this system is not capable of being utilized at the design of the Reactor Coolant System (RCS), public health and prest 39 safety is endangered.
Contrary to this concern, while the LPI cannot operate at the design pressure of the RCS, there currently exist the capability of providing forced cooling to 10 -
the core, at tha dasign proccuro of the primary system, without reliance on the LPI at pressures exceeding the design limits of the LPI -- see discussion below on feed and bleed cooling.
Therefore, there is no need for the residual heat removal system to be capable of operation at the design pressure of the RCS.
The final method of providing forced cooling of the reactor described in the contention is operation of the Emergency Core Cooling System in a feed and bleed mode.
This is presented to be unreliable because there is inadequate capacity and shielding for the storage of radioactive water bled from the primary system.
Contrary to the contention, adequate capacity and shielding is provided for che water bled from the primary system.
Feed and bleed operation, which is described in Licensee's testimony in response to UCS Contention 1, results in the fluid discharged from the RCS being initially received by the pressurizer relief quench tank.
If this mode of cooling continues, the mass of fluid bled from the primary system will exceed the capacity of the quench tank and will be discharged into the containment af ter the quench tank over-pressure protection rupture disk ruptures.
If feed and bleed cooling is continued, the initial source of High Pressure Injection (HPI) cooling water to the RCS, the borated water storage tank (BWST), will be emptied and supply for the HPI will be changed to the containment sump, via the LPI.
Throughout this sequence the containment, by design, provides adequate capacity and shielding for the discharged fluid.
The LPI is operated consistent with the design of the system (i.e.,
pressure, temperature, etc. ) to transfer water from the containment to the HPI.
It should additionally be noted relative to feed and bleed 1
cooling that, while the pressurizer power operated relief valve (PORV) may be utilized as the fluid discharge path from the RCS, feed and bleed can be accomplished with only safety-grade systems and components, i.e., the pressurizer safety valve (s) in conj unction with the BWST, HPI, containment and LPI.
- Also, this mode of operation can be continued as required to assure adequate core cooling until secondary side cooling is available and/or the primary system can be depressurized to allow the LPI to provide core cooling directly.
The only manual actions required for feed and bleed are:
For certain scenarios, manual actuation of o
the HPI.
If utilizied, manual opening of the PORV.
o If a low level in the BWST is reached, o
switchover of the HPI suction to the containment sump via the LPI.
BY WITNESS KEATEN:
Although the necessary RCS cooling water will be stored inside the containment, the use of the feed and bleed cooling mode will result in the transport of some of the coolant through components and piping located outside the containment building.
In response to the requirement of NUREG 0578 to perform a radiation and shielding design review of the spaces around systems that may as a result of an accident contain highly radioactive materials, Licensee has performed a study to identify any locations in which personnel occupancy may be unduly limited or safety equipment unduly degraded by the radiation fields which might exist after an accident.
The study is described in section 2.1.2.3 of the Restart Report.
The results of this studv h.2va identified only one concern in using the feed and ble 2d mode of cooling even if the coolant were very highly radioactive.
The concern is that a portion of the HPI piping is located in proximity to two motor control centers which perform functions important to safety.
Highly radicactive fluid in the HPI pipes would result in radiation levels at these motor control centers sufficiently high that the integrity of some of the materials found in the motor control centers cannot be demonstrated.
As a result, Licensee is commited to installing new shield walls between the HPI piping and the motor control centers which will reduce the radiation levels at the motor control centers to levels at which material integrity can be assured.
BY WITNESSES KEATEN AND JONES:
In summary, contrary to the basis of the above conten.c. ion, reliable methods currently exist to provide adequate core cooling.
With regard to the items specifically identified:
a) 13 -
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operation of the reactor coolant pumps is not required to assure adequate core cooling, therefore the design criteria listed are not applicable; b) the residual heat removal system is not required to operate at the design pressure of the i
primary system, therefore the system does not need to be designed for such conditions; and c) the emergency core cooling system can be operated in the feed and bleed mode with adequate capacity and radiation shielding for storage and recirculation of the radioactive water...
FIGURE 1 NATURAL CIRCULATICN CCCLING HOT LEG A k STEAM GENERATGR I f O
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CCLD TCOLD CCLD LEG
ROSERT W.
KEATEN Business Address:
GPU Service Corporation i
100 Interpace Parkway i
Parsippany, New Jersey 07054 Education:
B.S., Physics, Yale University, 1957.
Post-Graduate and Professional Courses in Mathematics, Engineering and Business, UCLA, 1960-1972.
Experience:
Manager, Systems Engineering Depart-ment, GPU Service Corporation, April 1978 to present.
Responsible for the development and application of specialized analytical skills in such areas as nuclear core reloads and fuel management; plant dynamic and safety analysis; system generating plant process computers; control and safety systems analysis, and analysis of plant operating performance for nuclear and fossil plants.
Served as a
Deputy Director of Technical Support at Three Mile Island during the post-accident period.
Program Manager, Light Metal Fast Breeder Reactor Technology, Atomics International Division of Rockwell International, 1974 to 1978.
Managed research and development programs performed for U.S.
Cepartment of Energy, including programs in reactor physics, safety and component development.
Manager of Systems Engineering, Light Metal Fast Breeder Reactor Program, Atcmics International Division of Rockwell International, 1968 to 1974.
Responsible for performance of safety 1
analyses, development of cafety criteria and development of instru-mentation, control and safety systems design.
N
American Representative to t.'te CECD Halden Reactor Project in No.uay, 1965-1968.
Participated in research on nuclear fuel performance, appli-cation of digital ecmputers to nuclear reactors, and on development and application of in-core instru-mentation.
Supervisor of Engineering, Sodium Reactor Experiment, Atomics International, Division of Rockwell International, 1962-1963.
Responsibilities included analysis and measurement of the nuclear heat transfer and hydraulic parameters of the reactor core and process systems; specification and installation of nuclear and process instrumentation; design and installation of new control systems.
Senior Physicist, Sodium Reactor Experiment, Atomics International, Division of Rockwell International, 1959-1962.
Performed measurements and analyses of the nuclear and thermal parameters of the reactor.
Experimental Physics Group, DuPont Savannah River Plant, 1957-1959.
Performed measurements and calcula-tions of the nuclear parameters of the reactor lattices.
Honors and Professional Affiliations:
Member of the Nuclear Power Plant Standards Steering Committee of the American Nuclear Society.
Member and past Chairman of the LMFBR Design Criteria (ANS-54) Standards Committee of the American Nuclear Society.
Registered Professional Engineer (Nuclear Engineering), California. =.
Publications:
" Analysis of TMI-2 Sequence of Events Operator Response," prer ntad to a special session of the American Nuclear Society Conference, San Francisco, November 1979; and to Edison Electric Institute Conference, Cleveland, October 1979.
"The Role of Instrumentation in the TMI-2 Accid ent," prescnted at the American Nuclear Society Conference, June 1980.
Safety and Environmental Aspects of Liquid Metal Fast Breeder Reactors" 35th Annual American Power Conference, Chicag o, Ill., May 197 3.
" Safety Aspects of the Design of Heat Transfer Systems in LMFER's" International Conference on Engineering of Fast Reactors for Safe
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and Reliable Operation, Karlsruhe, Germany, October 1972.
" Safety Criteria and Design for an FBR l
Demonstration Plant," ASME Nuclear Engineering Conference at Palo Alto, Calif., March 1971.
" Evaluation of Thermocouples for Detecting Fuel Assembly Blockage in LMFBR's," American Nuclear Society Annual Meeting, Los Angeles,
California, June 1970.
"A Mathematical Model Describing the Static and Dynamic Instability of the SRE Core II," Reactor Kinetics and Control, AEC Symposium Series 2.
( Also published as NAA-SR-8431. )
" Reactivity Calculations and Measurements at the SRE," ANS Topical Meeting:
Nuclear Performance of Power-Reactor Cores, September 1963.
" Measurement of Dynamic Temperature Coefficients by Forced Oscillations in Coolant Fl o w, " Trans-American Nuclear Society 5, No.
1, June 1962. l
1 i
" Analysis of Power Pamp Measurements with an Analog Computer," Trans-American Nuclear Society 5, No. 1, June 1962.
" Reflected Reactor Kinetics,"
NAA-SR-7263.
)
Many other reports covering analytical and experimental work.
I l,
ROBERT C. JONES, JR.
Business Address:
Babcock & Wilcox Company Nuclear Power Generation Division P.O.
Box 1260 Lynchburg, Virginia 24505 Education:
B.S., Nuclear Engineering, Pennsylvania State University, 1971.
Post Graduate Courses in Physics, Lynchburg College.
Experience:
June 1971-June 1975: Engineer, ECCS Analysis Unit, B&W.
Performed both large and small break ECCS analyses under both the Interim Acceptance Criteria and the present Acceptance Criteria of 10 CFR 50.46 and Appendix K.
June 1975-Present:
Acting Supervisory Engineer and Supervisory Engineer, ECCS Analysis Unit, B&W.
Responsible for calculation of large and small break ~ECCS evaluations, evaluations of mass and energy releases to the t
containment during a LOCA, and performance of best estimate pretest i
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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