ML20054B727
| ML20054B727 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/09/1982 |
| From: | Lobel R, Sheron B, Thadani A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054B726 | List: |
| References | |
| NUDOCS 8204190139 | |
| Download: ML20054B727 (15) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION REFORE THE AT0f11C SAFETY AND LICENSING BOARD In the Matter of
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LOUISIANA POWER & LIGHT COMPANY
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Docket No. 50-38? OL
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(Waterford Steam Electric Station, )
Unit 3)
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AFFIDAVIT OF RICHARD LOBEL, BRIAN SHERON AND ASHOK C. THADANI CONCERNING FEED-AND-BLEED AND EMERGENCY FEEDWATER SYSTEM RELIBILITY Q.'
Please state your nanes and by whon you are employed.
A.1(ai My nane is Richard Lobel.
I am employed by the United States ficclear Regulatory Connission as Section Leader, Auxiliary Systens Branch, Division of Systens Integration, Office o' Nuclear Reactor Regulation. A copy of ny professional qualifications is attached hereto.
(b) My name is Brian Sheron, I am employed by the United States Nuclear Regulatory Commission as Branch Chief, Reactor Systems Branch, Division of Systens Integration, Office of Nuclear Reactor Regulation.
A copy of my professional qualifications is attached hereto.
(c) fly name is Ashok C. Thadani.
I am employed by the Urited States Nuclear Regulatory Comnission as Branch Chief, Reliability and Risk Assessment Branch, Division of Safety Technology, Office of Nuclear Reactor Regulation. A copy of my professional qualifications is attached hereto.
8204190139 820412 PDR ADOCK 05000382
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. Q.2 Please provide the NRC Staff's probabilistic estinate of the reliability of the emeroency feedwater systen (EFWS) for the Waterford tinit 3 facility, and the bases relied upon by the Staff in formulating that estimate.
A.2 The Staff has estimated the frequency of the three events which require consideration of EFWS reliability.1/ First, potentially irre-coverable loss of main feedwater events (both main feedwater pumps incapacitated by other than loss of offsite power) has been estimated to be about 0.1/RY (Reactor Year) (Ref. 1).SI It has been further pcstulated that the electric-driven condensate pumps could be used to provide feedwater if the steam penerators were depressurized to below the condensate purp shut-off head (Ref. 1). Thus, assuning procedures and operator training are in place for restoring nain feedwater under various degraded conditions including the use of condensate pumps alone, it is reasonable to expect, for the reasons set forth in Ref. 1, that the frequency of irrecoverable losses of main feedwater event < could be reduced to the value of 0.03/RY.
Secondly, the frequency of loss of offsite power has been estimated by the Staff to be 0.27/RY based on the operating experience of nuclear power plants. The probability of non-recovery of offsite power in 1-1/2 hours is estinated to be 0.3/D (Denand) based on operatina experience
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In this discussion, consideration is limited to those events which
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relate to the feed and bleed issue.
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Attached hereto are copies of Refs. 1, 2 and 5.
Copies of Refs. 3 and 4 are readily available elsewhere.
. (Ref. 2).
Thus, the frequency of an extended loss of offsite power (and thus a loss of main feedwater) is calculated to be 0.08/RY.
Thirdly, the frequency of very small loss of coolant accidents (LOCAs) requiring energency core cooling system operation is estimated by the Staff to be about 0.01/RY (Ref. 1).
For purposes of this discussion, it is conservatively assuned that main feedwater cannot be restored for l
decay heat removal, although it could be mada available by opening the main stear isolation valves.
A combination of the frecuencies of the three events discussed above results in a composite value of 0.12/RY for sequences where energency feedwater must function or core melt may ensue. These frequencies are generic and nay vary from plant to plant depending on design, operation, and site differences.
The demand failure probability for the EFWS is estinated to be less than 4x10-5/D for the above events, within an uncertainty range of a facter of 10, as described in Section 10.4.9 of the Steff's SER for Waterford Unit 3 (Refs. 3,4).
While all potential common cause failures (such as connon support systems, equipment layout, and external eventsi are not considered in this isolated system reliability analysis, these types of considerations are included in the Staff's review of the EFWS.
A combination of the estimated initiating frequency of 0.12/RY and the demand failure probability of 4x10-5/D for the EFWS results in a core-melt estinate of 5x10-6/RY, which is in the range of core melt l
estimates for similar events calculated for the PWR reference plants in
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WASH-1400.
Thus, from t. risk standpoint, the potential effect of these sequences at Waterford Unit 3 is in the raroe of core nelt estinates
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, previously celculated without a reliance on any feed and bleed capability.
0.3 Please describe the factors which were considered by the Staff in assessing the adequacy of the Uaterford Unit 3 EFPS.
A.3 In addition to the probabilistic assessments discussed above, the Staff has considered the effect of NRC requirenents which provide (a) that the EFWS be protected from natural phenomena; (b) that materiels be of high quality; (c) that the EFWS be capable of periodic testing and inspection; (d) that EFWS controls meet high (safety-related) standards; (c) that EFWS outage tire during plant operatior be limited by the Technical Specifications; (f) that the EFWS be controllable from both within and cutside the control room; (g) that the EFWS be adequately isolated from non-essential systems whose failure could interfere with the EFUS's perforrance of its safety function; (h) that the EFWS be capable of withstanding any single failure of an actual component; and (i) th6t the EFWS be powered fron diverse power sources.
l The Staff's review of these and other factors has led it to conclude that the Materford Unit 3 EFWS design satisfies the Comission's requirerents. The Staff's analysis and conclusions in this regard are set forth in Section 10.4.9 of the Staff's SER for Waterford Unit 3 (Ref. 3).
0.4 Has the Staff's review of the accident at Three File Island, Unit 2 (TMI-2) resulted in any changes in or additional requirenents for
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auxiliary feedwater systens, such as the Waterford Unit 3 EFWS?
A.4 Following the TMI-2 accident, the Staff enbarked on a na.ior upgrading of the auxiliary feedwater systems (AFWS) for operating l
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, reactors. This upgrading took the form of recorrnendations ained at increasing the reliability of the AFWS by limiting outage times while the reactor was operating, requiring long term endurance tests for the AFWS pumps, requiring safety-related automatic initiation and flow indication, requiring punp protection from low suction pressure, and assuring adequate water sources by requiring that no single vahe failures could cause loss of water supply and that at least one water supply nust be protected from natural pheromena. These recommendations are also being applied to reactors which are the subject of operating license applica-tions, including the Uaterford Unit 3 EFWS. A discussion of how the Waterford Unit 3 EFWS meets thesa post TMI-2 requirements is provided in Section 10.4.9 of the Staff's SER for Waterford Unit 3 (Ref. 3).
0.5 please describe the reasons why the Staff has concluded that the Waterford Unit 3 EFWS has a greater reliability than that which right be expected from a reading of the reliability estimates provided in the Penorandum fron F. Rowsore and J. Murphy to R. Tedesco, dated January 29, 1982, entitled " Feed and Bleed Issue for CE Applicants."
A.5 The referenced nemorandun regarding feed and bleed as it affects the role of the EFWS can be subdivided in three major categories:
(1) events leading to loss of nain feedwater, (2) ability to recover main feedwater, and (3) reliability estimates for the EFWS.
The Staff has conducted an independent assessnent of EFWS reliability (Ref.1), and has concluded that it is not necessary to provide feed and bleed backup to EFWS to protect the health and safety of the public. The Staff's bases for this conclusion are set forth above.
In addition, the Staff believes that the Rowsone memorandun does not
. point to a different conclusion for the Waterford Unit 3 EFWS, for the following reasons.
The probability of complete losses of the EFilS in the Rowsome and Murphy nemorandun was based on past operating experience as reported in an ORNL report (CR-2497). The Staff is aware of ten events in which there was a loss of all emergency feedwater (two more events than are listed in ORNL report CR-2497). An analysis of this past experience as it relates to the Waterford Unit 3 EFWS design results in the following conclusions.
FirFt, post-THI recommendations should greatly lower the probability of occurrence for several of these events.
For example, in several events, the EFilS punos did not start on the automatic initiation signal; safety-related EFWS flow indication must now be provided, and an indicator such as that available at Waterford Unit 3 would alert the operator innediately that the EFWS was inoperable so that he could initiate a timely manual actuation.
In addition, human error resulting in a closure of valves in the pump discharge path, such as occurred during the TMI-2 accident, should be less probable now as a result of the required increased surveillance of the EFWS flow path after system testing or extended shutdown.
Secondly, some of these events were recoverable in less time than the tire calculated for loss of the secondary heat sink (i.e., stean generator dryout time).
The Waterford Unit 3 steam generators have a relatively large water inventory, which provides the operator with a greater period of time to attempt a manual start in the evert that the system does not start automatically.
A hunan error resulting in a closure of valves in the pump discharge path, such as occurred during the TMI-2 accident, should have a high probability of being corrected in the Waterford Unit 3 design before the heat sink is lost.
D e Thirdly,-sone of these events involved a type of failure that could s
not occur in an EFWS of the Waterford Unit 3 design. For example, several of these events resulted from clogged strainers in the EFWS piping; the strainers will be removed from the Waterford Unit 3 EFWS after startup testing.
In addition, one of these events resulted from the interference of a reactor control system with the function of the EFWS. This event was peculiar to reactors designed by Babcock & Wilcox and the problem was fixed following the Crystal River Unit 3 event of February 1980; accordingly, it is not applicable to the Waterford Unit 3 design.
In conclusion, the Staff's reanalysis of these data, taking into account III the post-TMI modifications and corrective actions, (2) the high probability of recovery of some of these events, (3) the limited applicability of sone of these events to the Waterford Unit 3 EFWS design leads the Sta'f to conclude that the Weterford Unit 3 EFWS is subiect to l
a denard failure prcbability of less than 10 4 per demand.
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l 0.6 In view of the Staff's conclusion as to the reliability of the Waterford Unit 3 EFWS, has the Staff concluded that a feed-and-bleed i
i capability is not necessary as a back-up systen to the Waterford Unit 3 EFWS?
A.6 Yes.
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Apart from the issue of feed-and-bleed concerning the auxiliary feedwater system. please describe the Staff's current activities in e
considering the need for a rapid depressurization capability to be installed in Conbustion Engineering (CE) nuclear stean supply systens?
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8-A.7 Recent stean generator tube rupture events have led the Staff to consider the need and advisability of requiring a rapid depressuriza-tion capability for utilization in the event of tube failures in both stean generators.
In addition, the Staff is looking at the benefits of provic'ing this capability to afford greater flexibility for the operator to respond to unforeseen events. The Staff nas requested that infernation be provided by CE as to the need for such a capability in CE Systen 80 systens.
CE has provided an initial response which indi-cated that the Systen 80 design was adequate without this capability, and suggested that any plant design modifications night more appropri-ately be directed to providing a rapid depressurization capability for the secondary systen and providing piping te channel additional ener-gency water sources to the stean generators. The Staff has reviewed CE's response and has reouested that additional information be provided by CE, as well as by the Waterford and San Onofre Applicants. Responses to the Staff's additional requests for infornetion have not yet been received.
The Staff has briefed the ACRS Subconn4ttee on Decay Heat Renoval Recuirements (on flarch 16,1982) and the full ACRS in Executive Session (on April 2, 1982) as to the status of the Staff's evaluation of this matter.
Subsequently, the ACRS issued a letter which indicated that while this evaluation should be conducted expeditiously, its resolution should not now be a condition for operation of Systen 80 plar.ts at full
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power or of plants havino similar features; the need for future hardware or procedural changes should be contingent upon the results of this evaluation (Ref. 5).
Richard Lobel t
/4in Brian Sheron
/1f.[d h%'
AsNok C. Thadani Subscribed and sworn to before me, a Notary Public in the County of fientgomery, State of fiarvland, this 9th day of April,1982.
4,> r _.,
n,,hA Notary Public l
My connission expires y /, /[f2.
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PROFESSIONAL QUALIFICATIONS RICHARD LOBEL AUXILIARY SYSTEMS BRANCH DIVISION OF SYSTEMS INTEGRATION U. S. NUCLEAR REGULATORY COMMISSION I am employed as a Section Leader with the Division of Systems Integration, United States Nuclear Regulatory Commission.
I graduated from California State University at San Jose with a B. S. in Mechanical Engineering in 1966.
I then began work as a Mechanical Engineer
-at Lawrence Livermore Laboratory, Livermore, California. At the same time I began work towards an M. S. degree in Mechanical Engineering at California Since my masters State University at San Jose which I received in 1970.
degree I have taken an additional number of university courses in nuclear and mechanical engineering.
In my current assignment I supervise a group of engineers whose review respon-These sibility includes those systems assigned to Auxiliary Systems Branch.
systems include the Emergency Feedwater System.
I have previously worked in the Division of Safety Technology where I was responsible for evaluating the need to implement proposed new safety reaufre-ments for nuclear reactors.
My previous assignments have included responsibility for fuel rod thermal performance including reviews of computer programs used by fuel vendors for f
predicting fuel conditions during steady state and transient conditions, fuel densification and analysis of fuel rods during a Loss-of-Coolant-Accident.
, During the period of 1966 to 1973 while I was employed by Lawrence Livermore Laboratory I was responsible for the mechanical design of nuclear physics experiments.
I have been a lecturer on nuclear fuel behavior at two University short courses titled " Nuclear Power, Safety and the Public" and " Nuclear Power Reactor Safety Analysis."
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Richard Lobel Safety Program Evaluation Branch Division of Safety Technology l
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STATEMENT OF PROFESSIONAL QUALIFICATIONS t
BRIAN WALTER SHERON My name is Brian Walter Sheron. I graduated from Duke University in Durham, North Carolina, in 1969 with a Bachelor of Science in Engineering (B.S.E.) majoring in electrical engineering.
I received my Masters Degree (M.S.) in nuclear engineering in 1971 and my Doctor of Philosophy (Ph.D) degree in nuclear engineering in 1975 both from the Catholic University of America in Washington, D.C.
I joined the Atomic Energy Commission in 1973 in the Division of Reactor Development and Technology and worked on the LMFBR. I joined the Nuclear Regulatory Commission in 1976 as an engineer in the Analysis Branch in the Division of Systems Safety.
In 1980 I was assigned to the Reactor Systems Branch, Division of Systems Integration, and was promoted to a Section Leader in the Branch that year.
In February of 1982 I was promoted to Acting Chief of the Reactor Systems Branch.
In this capacity I supervise the activities of approximately 30 engineers in the areas assigned to the Branch.
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' OTELLIC'.AL C.U,ali' LATIG:.
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Branch Chief of the Reliability and P.isk Assessment Branch, Divi. ion cf :sfe:y Technology in the Office of I'uclear Reactor Regulation.
I have-heir this position since August,1980.
In this capacity I supervise, direct, and coordir. ate the personnel and programs of the Reliability and Risk A. ess-ner.t Branch which is responsible for activities pertair.ing to the reliability anc risk assessment of the functional capabilities of nuclear power plant sa'ety systems, equipment, and procedures needed for safe plant shutdown fol'.oding transient and accident conditions.
I received a Bachelor of Science Degree in 1965 from the University of Tenressee and a Master of Science Degree in 1967 from the Catholic University.
Both of these degrees are in Chemical Engineering.
Fror 1967 to 1968, I was employed by Melpar Co. where I performed Research and Development studies on coal utilization and air pollution control.
i Fro-1968 to 1969 I continued my studies towards a Doctor of Philosophy Degree in Chemical Engineering at the Catholic University of America.
I l'> /,. : was c Lent.
in gi m.
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nu: 0 Electrit c,r.
Corporaticr. where I was initially involved ir-developin:1 nathematical models I
for the coolcown of the NERVA reactor. Sur.secuently I wa, involved in code development efforts and accident analysis.
From 1972 to 1974, I was a staff engineer at t e Singer Linulation Products where I wa-. involved in real time simulatic Of nuclear power plant behavior during norral and accident conditions.
I was also responsible for directing Research aref Development activities in modeli.3 techniques.
t In February 1974, I accepted employr.ent with :ne Atonic Energy Commission (now the Nu;1 ear Regulatory Commission) in the Reactor Systems Branch.
I was responsible for reviewing and coordinating the review of the Anticipated Transients Without Scram for Light Water Reactors.
In addition I reviewed the safety systems of both PWR and BWR desi;*:.
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REFERENCES 4
l 1.
Menorandum from S. Hanauer, Director, Division of Safety Technology, to R. Mattson, Director, Division of Systems Integration, dated April 2,1982.
2.
Letter from J. Anderson (ORNL) to P. Varanowsky (NRC), dated Auoust 24, 1981.
3.
Safety Evaluation Report Related to the Operation of Waterford Stear Electric Station, Unit flo. 3 (NUREG-0787, July 1981),
Section 10.4.9.
4 Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Combustion Engineering Designed Plants (NUREG-0635, January 1980).
5.
Letter from Paul Shewnon, ACRS Chairman to (lilliam Dircks, l
Executive Director for Operations, dated April 5, 1982.
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