ML19344E846
| ML19344E846 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/08/1980 |
| From: | Smith I Atomic Safety and Licensing Board Panel |
| To: | Environmental Coalition on Nuclear Power, METROPOLITAN EDISON CO., UNION OF CONCERNED SCIENTISTS |
| References | |
| ISSUANCES-SP, NUDOCS 8009110799 | |
| Download: ML19344E846 (40) | |
Text
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Bd 9/8/80 1IN UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION c):
mw "i
t:sNRt
. g g, ATOMIC SAFETY AND LICENSING BOARD N>M CMenoftfske Ivan W. Smith, Chairman Occieueg &
Dr. Walter H. Jordan Eacfr e
j Dr. Linda W. Little
'c
%,o IO In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289 (Restart)
(Three Mile Island Nuclear Station, Unit No. 1)
)
MEMORANDUM AND ORDER (September 8,1980)
The board's memorandum and order of August 15, 1980 stated which issues and contentions would be heard during the first hearing segment, commencing on October 15, 1980.
The attachment to this Memorandum and Order sets forth in full the contentions and the board questions which we will hear during the first hearing segment.
As previously ordered (except for TMIA's revised Contention 5),
written direct testimony and proposed exhibits must be served by Sep temtm_- 15, 1980.
This includes responses to board questions to be heard during the first hearing segment.
Unless otherwise indicated by the question, board questions are directed to the staff and the licensee.
Other parties may respond to board ques-tions which are related to their interests.
The board questions were discussed during the prehearing conference of August 12 and 13, 1980, at the transcript pages referenced with the questions.
Y Ds tsoogy,,g7g THIS DOCUMENT CONTAINS POOR QUAUTY PAGES
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. In some instances the questions have been refined.
The board has considered the staff's motion for clarification of some board questions, dated September 2, 1980,'and Mr. Lewis' motion for correction of August 28, 1980.
Questions 6(b), (f) and (g) have been clarified in response to the staff's questions.
UCS' Contention 12 - Question 1 has been modified, in part, because of Mr. Lewis' filing.
However, other e,xtensions of the question requested by Mr. Lewis are denied.
The board's concern in our amplification of UCS' Contention 12 is directed to environ-mental qualification of equipment, not to possible failure of equipment due to design inadequacies or operator errors, contrary to the thrust of Mr. Lewis' motion.
The board's question on the June 27, 1980 leak at TMI-1 (now board question No. 10) included a q testion about separation of storage facilities at Three Mile Island.
Tr. 2069.
Although this aspect of the question would otherwise be included in the issues to be heard at the opening segment, it will be put off until the i
parties address the entire question.
i TMIA's revision of its Contention 5, dated August 26, 1980, does not comport with the board's order of August 20, nor the
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board's directive and counsel's agreement at the August 12 prehear-ing conference session.
Tr. 2138-53.
Therefore, the board rej ects TMIA's August 26 revision, and has redrafted the contention on its own to-accurately reflect the agreement at the prehearing conference.
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. ECNP Contention 1(e) was accepted only to the extent that ECNP was permitted to adopt UCS Contention 8.
First Special Prehearing Conference Order, LBP-34,10 NRC 828, 844 (December 18, a
1979).
Therefore, ECNP Contention 1(e) is not included in the attachment.
The listed contentions are intended to be in a form reflecting all revisions.
Parties are directed to report to the board any corrections no later than September 15, 1980.
In the near future the board will issue a list containing all of its questions.
The board will also publish a notice of hearing announcing the commencement of the evidentiary hearing beginning at 9:00 a.m. on October 15, 1980 at 34 North Court Street (ground floor), Harrisburg, Pennsylvania.
The notice will require all parties and Commonwealth agencies to be present at the opening session of the hearing unless a participant is specifically excused.
THE ATOMIC SAFETY AND LICENSING BOARD
' ni///Im9W an W.
Smi'th, Chairman ATTACHMENT:
Contentions and Board Questions to be heard at evidentiary session beginning October 15, 1980.
Bethesda, Maryland September 8, 1980
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TMI-l Restart ATTACHMENT Contentions and Board Questions to be heard at evidentiary session beginning October 15, 1980.
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Three Mile Island Alert, Inc. (TMIA) l 1
Revised Contention 5 It is contended that Licensee has pursued a e,ourse of conduct that is in violation of 10 CFR 50.57, 10 CFR 50.40, 10 CFR 50.36, 10 CFR 50.71 and 10 CFR 50 Appendix B, thereby demonstrating that Licensee is not " technically.
. qualified to" operate TMI Unit 1 "without endangering the health and safety of the public."
This course of conduct includes:
a.
deferring safety-related maintenance and repair beyond the point established by its own procedures (see e.g.
A.P. 1407) ;
b, disregarding the importance of safety-related ma'intenance in safely operating a nuclear plant in that it:
1.
uses supervisory and other personnel to perform l
safety-related maintenance; I
2.
proposed a drastic cut in the maintenance budget; 3.
(Deleted]
4.. fails to keep accurate maintenance records related to safety itemn; 5.
has inadequate and understaffed QA/QC programs related to mn'ntenance; 6.
extensively uses overtime in performing safety-related maintenance.
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Union of Concerned Scientists 1.
The accident at Three Mile Island Unit 2 demonstrated
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l that reliance on natural circulation to remove decay heat is inadequate.
During the accident, it*was necessary to operate i
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 ("LOCA"). This is a threat to health and safety and a violation of both General Design Criterion ("GDC") 34 and GDC 35 of 1D. CFR Part 50, Appendix A.
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Union of Concerned Scientists 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 regula-tions applicable to systems important to safety and is sufficiently reliable to protect public health and safety:
a)
The reactor coolant pumps do not have an on-site power supply (GDC 17), their controls do not meet IEEE 279 (10 CFR 50.55a(h)) and they are not seismically and environmentally qualified (GDC 2 and 4).
b)
The residual heat removal system 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.
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Union of Concerned Scientists 3.
The staff recognizes that pressurizer heaters and associated controls are necessary to maintain natural circula-tion at hot stand-by conditions.
Therefore, this equipcient should be classified as " components important to safety" and required to meet all applicable safety-grade design criteria, 1
including buc not limited to diversity (GDC 22), seismic and environmental qualification (GDC 2 and 4), aute=atic initiation (GDC 20), separation and independence (GDC 3 and 22), quality assurance (GDC.1 ), adequate, reliable on-site power supplies (GDC 17) and the single failure criterion.
The staff's proposal to connect these heaters to the present on-site emergency power supplies does not provide an equivalent or acceptable level of protection.
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Union of Concetmed Scientists 4.
Rather than classifying the presurrizer heaters as safety-grade, the staff has proposed simply to ' add the pressuri-zer heaters to the on-si?.e emergency power supplies.
It has not been demonstrated that.this will not degrade the capacity, capability and reliability of these power supplies in viola-tion of GDC 17.
Such a demonstration is required to assure protection of public health and safety.
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Union of Concerned Scientists contention 5
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UCS Contention 5 is as follows:
Proper operation of power operated relief valves, (PO RV 's) associated block valves and the instruments and controls of these valves is essential to mitigate the consequences of accidents.
In addition, their failure can cause or aggravate a LOCA.
Therefore, these valves must be classified ap components important to safety and required to meet all safety-grade design criteria.
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Union of Concerned Scientists 6.
Reactor coolant system relief and safety valves form part of the reactor coolant system pressure boundary.
Appro-priate qualification testing has not been done to verify the capability of these valves to function during normal, transient and accident conditions.
In the absence of such testing and verification, compliance with GDC 1, 14, 15 and 30 cannot be found and public health and safety is endangered.
t Board question regarding UCS Contention 6:
The board wants more than just a schedule for testing of reactor coolant system safety and relief valves, as is required pursuant to NUREG-0578.
Is there reasonable assurance that the tests will be successful, e.g.,
that there is good evidence that the valves will indeed perform in an accident environment? Tr. 2374-85.
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Union of Concerned Scientists 7.
NRC regulations require instrumentation to monitor variables as appropriate to ensure adequate safety (GDC 13 and that the instrumentation shall directly measure the desired variable.
IEEE 279, $4.8, as incorporated in 10 CFR 50.55a(h),
states that:
To the extent feasible and practical protection system inputs shall be derived from signals which are direct measures of the desired variables.
TMI-l has no capability to directly measure the water level in the fuel assenblies.
The absence of such instrumentation delayed recognition of a low water level condition in the rnactor for a long pericd of time.
Nothing proposed by the staff would require a direct measure of water level or provide an equivalent level of protection.
The absence of such instrumentation poses a threat I
to public health and safety.
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Union of Concerned Scientists 8.
10 CFR 50.46 requires analysis of ECC$ performance "for a number of postulated loss-of-coolant accidents of different sizes, locations, and r ther properties sufficient to provide assurance that the en: re spectrum'of postulated loss-of-coolant accidents is covered.
For the spectrum of LOCA's, specific parameters are not to be exceeded.
At TMI, certain of these were exceeded.
For example, the peak cladding temperature exceeded 2200* fahrenheit (50.46(b)(1)), and more than 1% of the cIadding 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 smali LOCA could lead to the same consequences.
Additional analyses to show that there is adequate protection for the entire spectrum of small break locations have not been l
performed.
Therefore, there is no basis for finding compliance with 10 CFR 50.46 and GDC 35.
None of the corrective actions I
to date have fully addressed the demonstrated inadequacy of l
protection against small LOCA's.
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Union of Concerned Scientists Board question regarding UCS Contention 8:
m The board directs the staff and the licensee to present experts and the fundamental documents involved in the small break LOCA analysis, and to have very complete testimony on this subject.
The recommendations of NUREG-0565 and NUREG-0623 should be addressed.
It appears from the small break LOCA analysis that there is a large amount of reliance upon operator action and on non-safety-grade equipment.
The board wants that issue explored by testimony, including why such reliance is proper.
Tr. 2374-85.
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Union of Concerned Scientists 9.
The accident at TMI-2 was substantially aggravated by the fact that the plant was operated with a safety system inoperable, to wit:
two auxiliary feedwater system valves were closed which should have been open.
The principal reason why this condition existed was that TMI does not have an adequate system to inform the operator that a safety system has been deliberately dis-abled.
To adequately protect the health and safety of the public, a system meeting tne Regula-tory Position of Reg. Guide 1.47 or providing equiva-lent protection is required.
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Union of Concerned Scientists C
UCS Contention 10 is as follows:
The design of the safety system at TMI is such that the operator can prevent the completion of a safety function which is initiated automatically; to wit:
the operator can (and did) shut off the emergency core cooling system prematurely.
This violates 54.16 of IEEE 279 as incorporatec in 10 CFR 50.55 ( a)( h) which states:
The protection system shall be so designed that, once initiated, a protection system action shall go to conpletion.
The design must be modified so that no operator action can prevent the completion of a safety function once initiated.
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i Union of Concerned Scientists 12.
The accident demonstrated that the severity of the environment in which equipment important to safety must operate was underestimated and that equipment previously deemed to be environmentally qualified failed.
One example was the pressurizer level instruments.
The environmental qualification of safety-related equipment at TMI is deficient in three respects : 1) the parameters of the relevant accident environment have not been identified 2) the length of tine the equipment must operate in the environment has been underestimated and 3) the methods used to qualify the equipment are not adequate to give reasonable assurances that the equipment will remain operable. TMI-l should not be permitted to resume operation until all safety-related equipment has been demonstrated to be qualified to operate as required by GDC 4.
The criteria for determining qualification should be those set forth in Regulatory Guide 1.89 or equivalent..
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Union of Concerned Scientists Board's questions regarding UCS Contention 12:
1.
The TMI-2 accident demonstrated that some safety-related equipment may have been exposed or was in imminent danger of being exposed to environmental stresses beyond that for which it was qualified.
The board's concern is primarily with such equipment qualification.
In addition, environmental stresses to safety-related equipment will be of concern to the extent that such equipment is not included in existing staff requirements.
2.
Whichitems of Regulatory Guide 1.89 have been grandfathered with respect to TMI-l?
Explain any justification for allowing restart without compliance with the grand-fathered items.
3.
What are the environmental qualification criteria which equipment inside of contain-ment must meet with respect to radiation levels and length of time of exposure?
(Address the Interim Staff position on Environmental Qualification of Electrical Equipment, (NUREG-0588. )
Tr. 2374-85.
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Union of Concerned Scientists 14.
The accident demonstrated that there a're systems and components presently classified as non-safety-related which can have an adverse effect on the integrity of the core because they can directly or indirectly affect temperature, pressure, flow and/or reactivity.
This issue is discussed at length in Section 3.2, " System Design Requirements," of NUREG-0578, the TMI-2 Lessons Learn Task Force Report (Short Term ).
The following quote 'fzom page 18 of the report describes the problem:
There is another perspective on this question provided by the TMI-2 accident.
At TMI-2, operational problems with the conden-sate purification system led to a loss of feedwater and initiated the sequence of events that eventually resulted in damage to the core.
Several nonsafety systems were used at various times in the mitigation of the accident in ways not considered in the safety analysis; for exanple, long-term maintenance of core flow and cooling with the steam generators and the reactor coolant pumps.
The present classifica-tion system does not adequately recognize either of these kinds of effects that nonsafety system can have on the safety of the plant.
Thus, requirements for nonsafety systems may be needed to reduce the frequency of occur-rence of events that initiate or adversly affect transients and accidents, and other requirements may be needed to improve the current capability for use of nonsafety systams during transient or accident situations. In l
its work in this area, the Task Force will l
include a more realistic assessment of the interaction between operators and systems.
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i Union of Concerned Scientists Contention 14 (continued) e The Staff's proposes to study the problem further.
This is not a sufficient an swer.- All systems and components which can either cause or aggravate an accident or can be called upon to sitigate an accident must be identified and classified as components important to safety and required to meet.all safety-grade design criteria.
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,invironmental Coalition on Nuclear Power (ECNP)
Contention 1.
ECNP contends that:
(a) The plant conputer for TUI-1 is old, ob::olete, and inadequate to respond appropriate 1/ in energency situations.
During the accident at the adjacent CII-2, the alarn printer on the sinilar conputer at Unit 2 had a delay tine of over two and one-half hours at one point, and)ran nore than an hour behind events for over seven hours."
This delay cannot be viewed c~s having adequately served the needs of the operators of TUI-2, and there is no reason to believe that a sinilar accident situatior, with as severe or worse consequences, cannot occur at TUI-1 and bc' severely aggravated by slow and anbiguous conputei^ alarn printer readings.
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Environ = ental Coalition on Nuclear Power (ECN?)
Contention 1.
ECN? contends thac:
a
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e (c) The electronic signals sent to the control room in cany cases record the wrong parameters, and =ny, thereby, mislead the the reactor operator.
For instanco, in the case of the Electromatic Relief Valve ("ERV; the tiotropolitan Edisen designation is RC-HV2), the signal cent to tha control room to indicate a closure of this valve indicates only the electrical energi=ing of the solenoid which closes the val re, not the actual physical valve closing itselfl4).
Thic misleading signal aggravated the accident at T'II-2.
There is no reasonable assurance that this same problem, or comparable oneg cannot arise many times over at T!.tI-1.
It is the obligation of the Suspended Licensee to provide sufficient information en the perfor=ance capability of all pertinent components of the control system to reasonably ensure that electronic signals will record, accurately and in a ti=ely manner, all necessary and correct parameters.
Board Ruling:
Therefore, as a matter of board discretion and to assure an adequate evidentiary record, we retain contentions 1(c) and 1(d).
Licensee should address in contention 1(c) the topic of the aclequacy of Class lE control room instrumenta-tion following a feedwater transient and small break LOCA.
In contention 1(d) the licensee should address the ranges of instrumentation in connection with contention 1(c).
Memorandum and Order, 6/12/80, p.23.
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Environmental Coalition on Nuclear Power (ECNP)
Contention 1.
ECNP contends that:
i (d) The TMI-2 accident showed that many monitoring instru=ents were of insufficient indicating range to properly warn control room operators of ambient conditions.
For example, the " hot-leg" thermocouples went off-scale at 6200? and stayed off-scale for over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for reactor cop about 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> for reactor coolent loop Bt gpt loop A and i.
A higher temperature limit would have provided important information to the reactor operators.
This situation is unchanged at TMI-1.
All monitoring instruments for TMI-1 must be calibrated to provide full and accurate readings of the complete range of possible conditions under both normal and norst-case conditions.
In addition, it is reported that th9 pa went off-scale during the TMI-2 accidentt6.diation =onitors 1
It should'be noted here that this eventuality was predicted in 1974 by the EII-2 Intervenors, but dutifully denied by the NRC Staff and the Applicant durirg the TMI-2 licensing hearings.
Needless to say, the EII-2 Licensing Board accepted the assurances of adequate monitoring offered by the Staff and Applicant.
Yet a similar situation still exists at THI-1.
All radiation monitoring equipment cust ba capable of recording the maximum g
possible releases of radiation in the event of a worst-possible accident (Class 9) in excess of Design Basis Accidents l
Board Ruling:
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Therefore, as a matter of board discretion and to assure an adequate evidentiary record, we retain contentions 1(c) and 1(d).
Licensee should address in contention 1(c) the l
topic of the adequacy of Class 1E control room instrumenta-tion following a feedwater transient and small break LOCA.
In contention 1(d) the licensee should address the ranges of instrumentation in connection with contention 1(c).
Memorandum and Order, 6/12/80, p.23.
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Anti-Nuclear Groun Representing York (ANGRY) d V.
The NRC Order fails to require conditions fo-g MA e
restart following modifications in the design of 'the TM.-l 3
reactor without which there can be no reasonable assurance that l
TMI-l can be operated without endangering the public health and i
safety:
(B)
Installation of instrumentation providing reactor operators direct information as to the level of primary coolant in the reactor core.
(C)
Performance of an analysis of and implementation of modifications in the design and layout of the TMI-l control room as recommended in NUEIG 0560.
(D)
Installation in effluent pathways of systems for the rapid filtration of large volumes of contaminated gases and fluids.
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Steven C. Sholly CW TENTIN 01 It is contended that in order to adequately protect the public health and safety, the containment isolation signals for nfI-1 must include the following:
1.
A safety-grade high radiation signal for the reactor building vent and purge system.
2.
A safety-grade high radiation signal for the reactor building sump discharge piping.
It is further contended that such additions ta the containment isolation signals must be nade pric,r to the Restart Of DII-l in order to adequately
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Steven C. Sholly C
i Contention # 3 It is contended that as a result of LLeensee's Operating Procedures, the emergency core cooling system can be defeated by operator actio'ns during the, course of a transtent and/or accident at Unit 1. such defeat consisting of either throttling back the high-pressure injectLon pumps or tripping these pumps.
It is further contended that under the conditions of a loss-of-feedwater transient / loss of coolant accident at Unit 1, defeat of the emergency core cooling system high-pressure injection system by pump throttling and/or pump trip results in signLficant cladding metal-water reaction, causing the production of amounts of hydrogen gas in excess of the amounts required by NRC reSulations to be considered Ln the design and accident analysis of nuclear power plants.
It is contended further that such production of hydrogen gas A-22 e
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results in the high risk of breach of containment integrity due to the explosive combustion of the hydrogen gas in the containment.
Inasmuch as the emergency core cooling system is an engineered safety feature which is relied upon to protect the pubile health and safety, and because proper operation of the emergency core cooling system is required to provide reasonable assurance that Unit 1 can be operated without endangering the public health and safety, it is contended that the emergency core cooling system operating procedures must be modified in order to ensure compliance wLeh the GDC 35 requirement of negligible clad metal-water reaction following a loss-of-coolant accident (LOCA).
It is further contended that the emergency core cooling system operating procedures'must be appropriately modtfled prior to restart in order to provide for protection of the public health and safety.
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Steven C. Sholly Contention #5 It is contended that Licensee has not provided radiation monitoring instru=en:s in effluent discharge pathways which are capable of remaining on-scale during anticipated operational occurrences, postulated accidents, and Class 9 accidents as specified in Contention #17.
It is further contended tha: the insufficiency in range of these instru=en:s prevents the Licensee fro: =aking sufficiently accura:e 7:edictions of the quantities of radia: ion which are being released frc= D'.I-1, and tha: this places the public health and safety at significant risk because such inferna: ion is recuired by public officials and plan: operators :o provide the basis - for
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decisicas on the need for protective actions.
It is further centended : hat protec:ica of public heal:h and safety requires that the high-range effluen: =enitoring syste= be installed prior to Res:ar: cf 22-1, at:d tha: the high-range effluen =eni:oring syste= be capable of re=zining on-scale under conditions specified in this centention.
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Steven C. Sholly Contention is It is contended that the short-term actions identified in the Cot:nnission's Order and Nc: ice of Hearing dated 9 August 1979 are insufficient to provide the requisite reasonable assurance of operation without endangering public health and safety because they do not include the following items:
Completion of a f ailure mode and effects analysis (Fe A) of a.
the Integrated Control System.
Completion of the installation of instrumentation for b.
the detection of inadequate core cooling.
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Steven C. Sho11v Contention # 13 It is contended that the Unit I computer system does not meet the requirements for instru=entation and control specified in GDC 13, and is inadequate to insure proper operation of the Unit I reactor under all conditions of normal operation, including anticipated operational occurrences and postulated accident conditions.
It i,s further contended that the lack of real-time printout capability during accident conditions and the lack of sufficient redundancy in the computer system place the public health and safety at significant risk during accident conditions,'especially if co'mputer function is lost and no backup unit is available.
It is contended that until the Unit 1 co=puter system is upgraded to meet the standards of GDC 13 and until suitable redundancy is provided within the computer system to assure real-time printout capability at all times, permission for restart must be denied on the basis of risk to public health and safety due to inadequate availability of operational information to Unit 1 operators.
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Steven C. Sholly Contention # 15 It is contended that the design of the Unit 1 Control Room, instrumentation, and controls is such that operators cannot maintaLn system variables and systems within pr<escribed It is operating ranges during feedwater transtents and LOCA's.
further contended that this violates the provisions of GDC 13 regarding instrumentation and controls.
It is contended that in view of the numerous operating difficultles encountered with Unit 2, and the similartties in design and construction between Units 1 and 2, a thorough human factors enginering review of Unit l's Control Room is called for in order to provide assurancethattheoperator-instrumentation [nterfaceis such that the operators can exercise adequate cor.crol over the reactor and prevent off-site consequences from anticipated operational occurrences and postu'iated accidents.
It is further contended that in order to assure maximum protection for the public health and safety, the human factors engineering relvew and any necessary changes recommended as a result of this review must be completed prior to restart.
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2 Lewis Contention D
3.
Filters: There are New filters on the auxiliary building of ZIId2.
Ther_e are no similar structures ~ on the auiliary building 'of "J" Ii1'.
Further,preheaters cust be placed on the filters of the auxiliary building because they got wet during the accident on 3/28/79 in THI!2, To mitigste a similar accident in TMIy1 preheaterson the filters in the auxiliary building of TMI!1cre necessary.'
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Board questions:
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1.
(Tr. 2390-92)
Prior to the opening of the evidentiary hearing, the staff should inform the board as to when the staff will take a position on the applicability to this proceeding of NUREG-0694, "TMI-Related Require-ments for New Operating Licenses".
The following items in NUREG-0694 and/or NUREG-0660, " Action Plans for Implementing Recommendations of the President's Commission and Other Studies of TMI-2 Accident", are of particular interest to the board:
I.D.1 -- Control Room Design (f ollowing a human a.
factors analysis).
b.
II.E.1.1 -- Auxiliary Feedwater System { reliability avaluation using event-tree logic).
II.B.8 -- Rulemaking proceeding on degraded core c.
accidents, d.
[will be considered in a later hearing segment]
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3.
(Tr. 2392)
The results of the Interim Reliability Evaluation Plan (IREP), as applied to Crystal River, was scheduled for completion in July 1980.
(The board wants to receive a copy of this report.)
When will the IREP be applied to TMI-l?
a.
b.
Does the IREP address the adequacy of the pro-posed actions for B&W plants?
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5.
(Tr. 2393-94) When does the staff plan to report on its review of NUREG-0660 as applied to TMI-l?
(The board and the parties should be kept informed as quickly as the staff has identified any additional action plans that should be required for implementation, l
either before any proposed restart or for the long-term.)
6.
(Tr. 2394-96 )
Emergency Feedwater Reliability Is a loss of emergency feedwater following a main a.
feedwater transient an accident which must be protected against with safety-grade equipment?
Would such an accident be caused or aggravaged by a loss of non-nucIear instrumentation, such as occurred at Oconee?
I b.
In what respect is the emergency feedwater system vulnerable to non-safety-grade system failures and to operator errors?
What has been the experience in other power plants c.
with failures of safety-grade emergency feedwater systems, if they have such systems in other power plants?
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What operator action is required to operate in a feed-and-bleed mode following a loss of emergency feedwater?
If the emergency feedwater system were to fail, what e.
assurance do we have that the system can be cooled by the feed-and-bleed mode?
This is of particular
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l concern if the PORV's and safety valves have not been tested.under two-phase mixtures,
f.
Can the system be taken to cold shutdown with the feed-and-bleed cooling only?
Are both high pressure injection (HPI) pumps required to-dissipate the decay heat in the feed-and-bleed mode?
The board would like an evaluation of the reliability of the feed-and-bleed system.
Has there been any experience using that system?
g.
If there is a loss of steam in the secondary system which results in failure of the turbine-driven feedwater pumps, will both motor-driven pumps be required to supply the requisite amount of feedwater?
Does this meet the usual single-failure criteria since it appears -that a redundant system requires multiple components to operate?
h.
Can the turbine driven pumps and valves be operated on Direct Current, or are they dependent upon the Alternating Current safety buses?
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________,-_,7-_,7,.we.
_.y
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y.o 7,.,
i.
Will the reliability of the emergency feedwater system be greatly improved upon conversion to safety-grade, and is it the licensee's and staff's position that the improvement is enough such that c
the feed-and-bleed back-up is not required?
J.
Will the short-term actions proposed improve the reliability of the emergency feedwater system to the point where restart can be permitted?
k.
Question 6 should be addressed with reference to Florida Power & Light Co. (S t. Lucie, Unit'2),
ALAB-603, (July 30, 1980) ;
i_.e_. whether loss of emergency feedwater is a design basis event not-withstanding whether design criteria are met.
7.
(Tr. 2396)
Following the investigation of the Crystal River incident, the staff issued NUREG-0667, "Itansient i
Response of Babcock & Wilcox-Designed Reactors".
Which of the recommendations in Table 2.1 of that report does the staff believe should be implemented for TMI-l prior to start-up, which should be included in the long-term actions, and which, if any, are not needed for TMI-1 and why not?
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8.
(Tr. 23 97)
Even though no contentions survive on the issues raised by short-term Item 4 of the Commission's August 9, 1979 order, the board wants testimony presented on the issue raised by this item.
Short-term Item 4 requires :
The licensee shall demonstrate that de-contamination and/or restoration operations at TMI-2 will not affect safe operations at TMI-1.
The licensee shall provide separation and/or isolation of TMI 1/2 radioactive liquid transfer lines, fuel handling areas, ventila-tion systems, and sampling lines.
Effluent monitoring instruments shall have the capability of discriminating between effluents resulting from Unit 1 or Unit 2 operations.
[ emphasis' l
added]
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1 COURTESY NOTIFICATION This is intended solely as a courtesy and convenience to provide extra time to those notified.
Official service will
~
be separate from the courtesy notification and will be made by the Office of the Secretary of the Commission.
I hereby certify that I have today mailed copies of the board's MEMORANDUM AND ORDER, dated this date, to the persons designated on the attached Courtesy Notification List.
SW dA:Jo e
Doris M. Moran Clerk to the Atomic Safety and Licensing Board Bethesda, Maryland September 8, 1980 l
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COURTESY NOTIFICATION LIST George F. Trowbridge, Esq.
Dr. Chauncey Kepford Shaw, Pittman, Potts & Trowbridge Envircr= meal Coalition on Nuclear Power 1800 M Street, N. W.
433 Orlando Avenue Washington, D. C.
20036 State College, Pennsylvania 16801 Counsel for NRC Staff Mr. John E. Minnich, Chaim=1 Office of Executive Iagal Director Dauchin County Board of ramrissioners a
U. S. Nuclear Regulatory Comission Dauphin County Courthouse Washington, D. C.
20555 Front and Market Streets Harrisburg, Pennsylvania 17101 Ms. Marjorie M. Aamodt R. D. #5 Mr. Marvin I. Imwis Coatesv111e, Pennsylvania 19320 6504 Bradford Terrace Philadelphia, Pennsylvania 19149 Daniel M. Fell, Esq.
ANGRY Jordan D. Cunningham, Esq.
32 South Beaver Street Fox, Farr & Cunningham York, Pennsylvania 17401 2320 North Second Street Harrisburg, Pennsylvania 17110 609 Montpelier Street William S. Jordan, III, Esq.
Baltie re, Maryland 21218 Hare n & Weiss 1725 I Street, N. W.,
Suite 506 Karin W. Carter, Esq.
Washington, D. C.
20006 Assistant Attorney General 505 b ~t4ve House John A. Imvin, Esq.
P. O. Box 2357 Assistant Counsel Harrisburg, Pennsylvania 17120 Pennsylvania Public Utility Comission P. O. Box 3265 Walter W. Cohet, Esq.
Harrisburg, Pennsylvania 17120 Consuner Advocate Department of Justice Mr. Steven C. Sholly 1425 Strawberry Square 304 South Market Street Harrisburg, Pennsylvania 17127 Mechanicsburg, Pennsylvania 17055 Ellyn R. Weiss, Esq.
Theodore A. Adler, Esq.
Hanacn & Weiss Widoff Reager Selkcwitz & Adler, P.C.
1725 I Street, N. W. Suite 506 P. O. Box 1547 Washington, D. C.
20006 Harrisburg, Pennsylvania 17105 l
Robert E. Kelly, Jr., Esq.
Duane, Morris & Heckscher P. O. Box 1003 203 Pine Street, Suite 401 Harrisburg, Pennsylvania 17108 l
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