ML19346A048
| ML19346A048 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/26/1981 |
| From: | Cole R, Lazo R, Luebke E Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| ISSUANCES-OL, NUDOCS 8106040461 | |
| Download: ML19346A048 (39) | |
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UNITED STATES OF AMERICA Q
g:,'ye teeres,y NUCLEAR REGULATORY COMMISSIGN 6
. 5 Smee ATOMIC SAFETY AND LICENSING BOARD k,
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Before Administrative Judges:
$@jg gjM f 7 Jcgy Robert M. Lazo, Esquire, Chainnan c
Erineth A. Luebke, Ph.D.
Richard F. Cole, Ph.D.
In the Matter of:
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DUKE POWER COMPANY Docket Nos. 50-369-OL
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50-370-OL (William B. McGuire Nuclear Station,
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Units 1 and 2 -- reopened
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operatinglicenseproceeding)
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Mr. William L. Porter, Charlotte, North
, j k,fp/,9 Carolina and Messrs. J. Michael McGarry and Malcolm H. Phillios, Jr., Washington, D.C.,,[',
for tne applicant, Duke Power Company.
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k Mr. Shelley Blum, Durham, North Carolina, l1 C.'/,# '7 L~ 41.II.'fgp/SS7 for the intervenor, Carolina Environmental Q.
Study Group.
,.. Q Messrs. Edward G. Ketchen, Stechen H. Lewis,*
- ~L[p Lawrence J. Chandler and Josech F. Scinto, for tne Nuclear Regulatory Comnission staff.
Dr. John M. Barry, Charlotte, North Carolina, for Mecklenourg County, North Carolina.
Mr. David Carelock, Charlotte, North Carolina, for the City of Charlotte, North Carolina.
SUPPLEMENTAL INITIAL DECISION
' Reopened Operating License Proceeding) l 59 N
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i INDEX Page Nos.
I.
INTRODUCTION 1-5 II. MATTERS IN CONTROVERSY 5-9 III.
FINDINGS OF FACT 9-20 IV ADDITIONAL EVIDENCE PRESENTED 20-29 V
10 CFR PART 2 - APPENDIX B 30-31 i
VI. CONCLUSIONS OF LAW 31-32 32-33 VII. ORDER APPENDIX A - STAFF, APPLICANT AND INTERVENOR'S EXHIBITS i
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ttilSSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges Robert M. Lazo, Chairman Dr. Emeth A. Luebke, Metroer Dr. Richard F. Cole, Member In the Matter of Occket Nos. 50-369-OL OUKE POWER COMPANY 50-370-OL Il (William B. McGuire Nuclear i)
Station, Units 1 and 2)
May 26, 1981 SUPPLEMENTAL INITIAL DECISION (Reopened Operating License Proceeding)
I.
INTRODUCTION 1.
An Initial Decision (Operating License Proceeding) was issued on April 18, 1979, Duke Power Cemaany (William B. McGuire Nuclear Station, Units 1 and 2), LBP-79-13, 9 NRC 489 (1979) and is incorporated by refer-I ence in this Supplemental Initial Decision. On the basis of specific l
findings of fact and conclusions of law, the Licensing Board ordered that I
the Director of the Office of Nuclear Reactor Regulation, upon making requisite findings with respect to uncontested matters not embodied in the Initial Decision, was authorized to issue operating lice.ases for the two l
l units. McGuire, suora, 9 NRC at 547-8. However, at that time, the i
Licensing Board stayed the effectiveness of the Initial Decision "until I
further order by the Board following the issuance of a supplement to the Nuclear Regulatory Comission ("NRC") Staff's Safety Evaluation Report l
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. ("SER") addressing the significance of any unresolved safety issues."
The NRC Staff issued the supplement on May 23, 1980 as Supplement No. 3 to NUREG-0422 (May 1980) (Staff Exhibit H). On May 30, 1980, the Duke Power Company (Applicant) moved the Licensing Board to lif t its stay of the Initial Decision.
2.
On June 9, 1980, Intervenor, Carolina Environmental Study Group (CESG), filed a re.gonse opposing Applicant's motion to lift the stay; also a motion requesting the reopening of the McGuire operating license hearing and the admission of new contentions. CESG amended its motion to reopen on August 15, 1980, and advanced four contentions relating to hydrogen generation and control. By Memorandum and Order of November 25, 1980, the Board granted CESG's motion to reopen and admitted CESG's four Contentions and denied the Applicant's motion to lift the stay of the Initial Decision.
3.
On November 25, 1980, the Board granted Applicant's motion for a low power license to the extent of authorizing fuel loading, initial l
criticality, and zero power testing. We denied Duke's request with
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respect to low power testing at up to 5% of full power. License NPF-9 for fuel loading, initial criticality and zero power testing was issued January 23, 1981.
4.
On November 7,1980, CESG moved to admit two additional conten-tions relating to Class 9 accidents and to emergency planning for such accidents. On February 13, 1981, the Licensing Board denied CESG's motion to add these two additional contentions.
5.
Mecklenburg County on December 23, 1980 and the City of Charlotte on January 13, 1981, requested participation in the reopened t
. hearings as interested government bodies as permitted in 10 CFR 12.715(c).
The Licensing Board approved both requests on January 26, 1981.
6.
Public evidentiary hearings regarding the issues raised by CESG were held in Charlotte, North Carolina on February 24-27, March 3-6, March 10-13, and March 17-19, 1981. The parties presenting evidence at the hearings were Applicant, NRC Staff, and CESG. Mecklenburg County and the City of Charlotte participated. The decisional record in this proceeding consists of the transcripts of the evidentiary hearings, all material received into evidence by the Licensing Board during such hearings, and the decisional record established in the previously issued Initial Decision in the operating license proceeding. At the close of the hearing, the Licensing Board granted the Staff's request to hold the record open for the limited purpose of possible inclusion of the results of the Staff review of a discrete issue regarding polyurethane foam.
(Tr. 5252). A Staff affi-davit was submitted on March 27, 1981 and has been admitted as Staff Exhibit Q. We permitted the parties an opportunity to respond and Applicant filed an affidavit in response which has been admitted as Exhibit 9.
On March 27, Mr. Jesse L. Riley, on behalf of CESG, also filed an affidavit.
It has been identified as CESG's Exhibit 63. However, it is not responsive to the Staff's affidavit and has not been received in evidence. Ar. index of exhibits is attached as Appendix A.
7.
CESG contentions in this proceeding are as follows:
Contention 1: The licensee has not demonstrated that, in the event of a loss-of-coolant accident at McGuire:
1.
substantial quantities of hydrogen (in excess of the design basis of 10 CFR 550.44) will not be generated; and w
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. that, in the event of such generation, the hydrogen will not combust; 2.
and that, in the event of such generation and combustion, the 3.
containment has the ability to withstand pressure below or above the containment design pressure, thereby preventing releases of off-site radiation in excess of Part 100 guideline values.
Contention 2: Neither licensee nor NRC staff has demonstrated that a McGaire ice containment will not breach as the result of the rapid combustion of quantities of hydrogen which a dry containment would withstand.
Neicher licensee nor NRC staff has demonstrated that the Contentien 3:
emergency planning radius of 10 miles is sufficient for protecting the public from the radioactive releases of a low pressure, ice condenser containment ruptured by a hydrogen explosion.
Licensee and NRC planning do not provide for crisis reloca-contention 4:
tion which would be required as a result of containment breach and radfo-active particle release.
CESG Contention 2 attempts to raise as an issue a ccmparison of the 8.
structural capabilities of the McGuire containment and other larger contain-The Licensing Board views Contention 2 as an expression ment structures.
of concern regarding the ability of the McGuire centainment to withstand the effects of a hypothetical hydrogen combustion.
In that such concern is embraced within CESG Contention 1, specific findings regarding Conten-tion 2 are unwarranted.
It also became clear during the course o' the hearing that the issues raised by Contentions 3 and 4 relating to emergency planning are not reached unless CESG is successful regarding Contention 1.
Tr. 2829-34 and 3434-5.
Evidence regarding Contentions 3 and 4 was Tr. 3481-3.
deferred pending this Board's ruling on Contentien 1.
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! In making the findings of fact and conclusions of law which follow, 9.
the Board considered the entire record of the proceeding and all the proposed findings of fact and conclusions of law submitted by the parties.
Each of the proposed findings of fact and conclusions of law which is not incorporated directly or inferentially in this Supplemental Initial Decision is rejected as being unsupported in law or fact or as being unnecessary to the rendering of this Decision. The Board is guided in this operating license proceeding by Appendix A,Section VIII of 10 CFR Part 2, which in subsection (b) provides that the Board will make findings on matters in controversy among the parties.
II. MATTERS IN CONTROVERSY
- 10. CESG Contention 1 is the same as the contention the Board drafted 1
as Contention 11 in the Three Mile Island Restart proceeding to comply with Commissicn policy.
In the TMI Restart proceeding, the Commission issued two dr.cisions regarding the proper scope of litigation of matters involv-ing hydrogen generation (Metrocclitan Edison Ccmoany Three Mile Island Nuclear Station, Unit No.1, CLI-80-16,11 NRC 674 (1980) and Order of i
September 26,1980(OccketNo.50-289(Restart)). The Commission decisions discussed Selow, are applicable to this proceeding.
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- 11. The issue of hydrogen generation was brought before the Ccmmission in the form of two certified questions:
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1.
Whether the provisions of 10 CFR 50.44 should be waived or exceptions made thereto in this proceeding where a crima facie showing has been made under 10 CFR 2.758 that nydrogen gas generation during the TMI-2 accident was well in excess of the amount required under 10 CFR 50.44 as a design basis for the post-accident combustion gas control system for TMI-1.
2.
Whether post-accident hydrogen gas control should be an issue in this proceeding where post-accident hydrogen gas control was perceived to be a serious problem and was in fact a problem
[Three Mile Island.11 NRC at during]the TMI-2 accident.
674-5.
12.
In its ruling the Commission declined to waive or except the hydrogen generation provisions in 10 CFR 550.44. This regulation limits the amount of hydrogen, generated during the course of a loss-of-coo bot accident to hydrogen associated with a five percent metal-water reaction.
It mum be taken into account in the design of nuclear reactor containment systems. Tia Comission, in its Memorandum and Order in the Three Mile Island case stated:
l The Three Mile Island accident has in fact raised a safety issue regarding hydrogen control measures following a loss-of-coolant accident that should be addressed. The Comis-sion believes that, quite apart from 10 CFR 50.44, hydrogen gas centrol could properly be liticated in [the Three Mile Island Nuclear Station, Unit No.1. proceeding under 10 CFR Part 100. Under Part 100, hydrogen control measures beyond those required by 10 CFR 50.44 would be required if it is determined that there is a crt.dible loss-of-coolaiticcident scenario entailing nydrogen generation, hydrogen ccmoustion, containment breacn or leaking, and offsite radiation doses in excess of Part 100 guiceiine values. Tne design casis assump-tions of 10 CFR 50.44, in particular the assumotien that hydrogen generation following a loss-of-coolant accicent is dependent on ECCS design as opposed to actual ECCS oceration, do not constrain the choice of credible accident sequences used under 10 CFR 100.11(a). Union of Concerned Scientists v. AEC, 499 F.2d 1069, 1090 (D.C. Cir. 1974 Thus we answer the second certified question in the affinnative.
(emphasis added) i
. We answer the first certified questica in the negative. We are of course aware that the Three Mile Island accident resulted in hydrogen being generated far in excess of the hydrogen generation design basis assumptions of 10 CFR 50.44. This was because the operator interfered with actual ECCS operation with the result that the safety system did not operate as designed and as 50.44 assumed it would operate. However, this is a safety issue that is not peculiar to Three Mile Island Unit 1 -- it is an issue that is common to all light water power reactors because operators generally have the physical capability to interfere w :h autcmatic ECCS operation. The proper response to this issue is not waiver of the rule under 10 CFR 2.758 because this case presents no "special circumstances," but rulemaking to either amend or suspend the present rule. The Commission is planning a broad rulemaking croceeding that will address the general question of possioie This safety features to ceal with degraded core conditions.
rulemaking proceeoing will inciuce measures to ceal with hydrogen generation followi'1g a loss-of-coolant accident.
(emphasisadded) the hydrogen control issue can be litigated under 10 CFR Unde'. Part 100 the likelihood of an accident entailing Part 100.
generation of substantial (in excess of 10 CFR 50.44 design basis) quantities of hydrogen, the likelinood and extent of hydrogen combustion, and the ability of the reactor containment to withstand any nydrogen conbustion at pressures below or above containment A critical issue here design pressure would all be at issue.
would be the likelinood of an operator interfering with ECCS operation.
However, after the Three Mile Island accident the Staff has given licensees explicit instructions not to turn off prematurely the As nated above, it was operator interference with ECCS syster..
ECCS operation that was the root cause of the hydrogen generaticn problea at Three Mile Island Unit 2.
In our view this instruction which nad not been issued when 50.44 and General Design Criterion 50 were promulgated, compensates for the less conservative analytical framework of part 100, and serves as a basis to sustain tre present hydrogen generation assumptions of 50.44 at least for the interim until the degraded core rulemaking can be completed.
11 NRC at 675-6.
The Board has limited its scope to consideration of credible accidents.
The degraded core rulemaking is viewed as providing a forum for the treat-ment of other accidents.
. 13. The Commission has provided guidance with these rulings. The regulations recognize and allow some measure of hydrogen production and the Commission granted the parties the right to litigate whether exces-sive amounts of hydrogen can be generated.
Permissible limits for hydrogen are stated in Section 50.44. The Commission, in its TMI ruling provided a way to consider this issue under 10 CFR Part 100. The interpretation is that a party must prove a credible accident that will give rise to the production of excessive hydrogen. A party must show a credible condition wherein the core is inadequately cooled for a sufficient period of time. CESG has raised the hydrogen generation issue, and under the Commission's ruling, it is considered to have the burden to establish a credible accident scenario involving hydrogen production resulting in offsite doses in excess of 10 CFR Part 100 limits. The burden is further clarified by Commissioners Gilinsky and Bradford in their dissent to the Commission's September 26, 1980 Order denying reconsideration of CLI-80-16:
Moreover, Chairman Ahearne and Cormissioner Hendrie are, in l
effect, saying that even after experience has amply demonstrated the adequacy of safety regulations covering the internal components of the reactor, the burden is still on a challenger to lay out a specific accident sequence to the Commission which leads to containment failure and public radiation exposere in excess of '. hose permitted by Part 100.
TMI (Restart), supra, Order, dissenting opinion, slip op. at p.2 (Septemoer 25,1980).]
14.
Part 100 is a siting regulation, and it establishes radiation limits at a certain boundary frem the plant surrounding the " exclusion" area.
These radiation exposure limits are 25 rem to the whole body, or 300 rem to the thyroid from iodine exposure.
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15.
In summary, the question of whether there is a credible loss-of-coolant accident involving hydrogen generation, hydrogen combustion and breach or leakage of the containment, with consequent offsite doses in excess of the Part 100 guideline values, is litigable under 10 CFR Part 100, notwi.thstanding the provisions of 10 CFR 550.44. The Licensing Board admitted CESG's Contentions 1 through 4 on the basis of this Commission precedent.
III.
FINDINGS OF FACT
- 16. The Applicant offered evidence regarding Contention 1 concerning the lack of credibility of a TMI-type accident sequence. The evidence related principally to a sequence characterized as S2D, which is a small break LOCA sequence with the break occurring anywhere in the primary coolant system; not just a TMI-2 small break sequence caused by an open relief valve initially actuated because of a loss-of-feedwater event. Cross examination was permitted into other sequences in which the evidence suggested a relevance or similarity to the TMI-type events or to the S20 sequence.
(Tr.3086-89,3374,4065).
- 17. The excessive hydrogen produced during the TMI-2 accident was a direct result of a reaction between the zirconium in the fuel cladding and steam and/or water during a loss-of-coolant accident which. led to an inadequate core cooling situation. The Emergency Core Cooling System (ECCS) is designed to prevent an inadequate core cooling situation which could resule in high temperatures of the core and excessive hydrogen production.
The sequence of events which occurred at TMI-2 was:
(1) a loss of feedwater transient resulting in high reactor coolant system pressure which was relieved by the pressurizer relief valve; (2) failure of.the relief valve to close resulting in a continued loss-of-coolant; (3) premature operator interference with the emergency core cooling system resulting in inadequate cooling of the reactor core and excessive core temperaturcs; and (4) production of hydrogen from the reaction of approximately 45% of the zirconium clad and steam in the presence of the excessively high temperatures.
If the operators at TMI-2 had not prematurely tenninated the ECCS operation there would not have been excessive hydrogen generation.
(App. Panel I: Canady, Reed and Barron, following Tr. 2864, Tr. 2870-3, 3086-89, 3374-5, 4468-70 and Staff Testimony Regarding Hydrogen Control folicwing Tr. 4353).
- 18. Applicant has made improvements at the McGuire facility subsequent to the TMI accident in the areas of personnel, equipment, procedures and training to effectively preclude improper operator termination of the McGuire ECCS. New technical specifications at McGuire require the following person-nel chages:
(1) a Senior Reactor Operator must be present in the control I
room at all times, in addition to a Reactor Operator; and (2) a Technical Advisor to the Shift Supervisor must be present on all shifts and available to the control rocm within tai minutes. Applicant's current staffing proced-
"res require two licensed reactor operators in the control room instead of one.
The second reactor operator may be absent for short periods of time.
(App.
Panel 1 at 2).
The technical and diagnostic capability in the control room has been 19.
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increased by adding a Technical Advisor to advise the Shift Supervisor.
Shift Technical Advisors have been selected from among the group of licensed Senior Reactor Operators at McGuire, all of whom have "eceived additional
. simulator training and have received additional academic training, includ-ing instruction in heat transfer, fluid flow, thermodynamics, and plant transients. The Shift Technical Advisor provides additional evaluation and assessment of both normal and unanticipated transients. The Senior Reactor Operator must be in the control room at all times the plant is above the cold shutdown mode of operation. The Assistant Shift Supervisor has been assigned administrative duties, relieving the. Shift Supervisor of duties which could detract from his management responsibility for safe operation of the plant. These changes provide additional excertise in the control An ECCS termination decision will be made by the Senior Reactor room.
Operator with input available from the Shift Technical Advisor and the Reactor Operator. This change provides greater assurance that the ECCS will not be prematurely terminated by operator action. There are an acceptable number of both reactor operators and senior operators to run the Ms, aire facility. The Applicant is engaged in a long-tenn hiring and training program to obtain additional reactor operators and senior operators, including additional personnel to account for attrition.
(App. Panel I, Staff Exh. I, pp. 22-32; Tr. 2880-4, 3007-13, 3028-32).
20.
Equipment modifications have been made at the McGuire plant subsequent to the TMI-2 accident. These include installation of a subcool-ing monitor to monitor the approach to an inadequate core cooling situation.
Alarms are also provided to warn the operator of an approach to a potential inadequate core cooling condition. Applicant is planning to install a reactor vessel level measurement system at McGuire which is designed to monitor the water level in the reactor and provide further indication of an approach to an inadequate core cooling situation. These modifications and additions will increase the assurance that operation of the plant will be done in a safe manner. Pressurized water reactors are operated at tem-peratures below the saturated temperature. This is the temperature at which water will boil at a corresponding pressure.
Equipment modifications were designed to warn the operator of operation beyond the normal operation approaching the saturated temperature that could lead to inadequate core cooling. New instrumentation such as a subcooling monitor and associated equipment would monitor the approach to inadequate core cooling conditions.
The subcooling monitor is part of the McGuire in-plant computer system.
It will display to the operator on the computer video screens, the actual system conditions on a temperature vs. pressure graph.
It will indicate to the operator what the core conditions are relative to a saturation curve.
(App. Panel I, Tr. 2925-31,2983,3009-12)
- 21. The subcooling monitor is basically software, which utilizes hard-wired inputs that are scanned at a minimum of once per minute and gives the operator a video display of the information. The problem with computer Lcklog due to the delay in the printers that occurred at TMI, which did not have a video display, will not occur at McGuire. The operators are thorcughly trained through the use of steam tables to calculate margins to saturation.
Thus, they compute margins even if the monitor were to fail. There is no problem similar to that which,ccurred at TMI due to computer backlog with respect to input to the control room via the subcooling monitor.
(Tr. 2884-7, 3009-11,3032-4)
- 22. On redirect, it was shown that a hydrogen bubble would not occur in a
, the reactor vessel and would not interfere with natural circulation.. Appli-cant has installed a redundant reactor coolant venting system located directly off the top of the reactor vessel head. This system alicws for detection of any hydrogen bubble fomed in the reactor vessel and for venting any hydrogen off into the pressurizer relief tank rather than directly into the containment atmosphere.
(Tr. 3061-3, 3092-3).
- 23. Operator training has been substantially enhanced. Subsequent to the TMI-2 accident, t.ie training program has been revised and expanded to reflect the lessons learned from the accident.
Revisions include the addition of a TMI-type accident scenario in the simulator portion of the training program, priority is given to the avoidance of an inadequate core cooling situation. Candidates for the cperator training program are employed by Applicant for several years before being selected.
Screening of candi-dates for the training program includes consideraticn of aptitude test results, senicrity, training grades throughout employment, and empicyment evaluations.
The operator training program is lengthy, ever 2i years, and intensive. The for,al program uses an effective mix of formal classrcem presentations, research reactor training, on-the-job training using task lists, simulater operation and both written and oral examination. The prcgram includes instruction in pcwer plant operating practices and r.uclear theory.
Specific course material includes electric theory, heat transfer, fluid ficw, themo-dymanics, chemistry, physics, mathematics, health physics, reacter theory, nuclear systems, transients, radioactivity theory, radiatien detection and instrument control theory. The operators receive training in monitoring and reading of indicators in the control recm and on actions to detect and
. A demonstration of a TMI accident, and'also respond to false readings.
natural circulation and ECCS operation have been added to the simulator The instructors conducting the operator training program at program.
McGuire are, with the exception of one Reactor Operator, all holders of a Senior Reactor Operator license. Prior to receipt of a Reactor Operator license, each candidate is required to comply with all appropriate provisions of 10 CFR Part 55, to include successful completion of an extensive operator examination.
Each licensed operator is required to undergo annual requali-fication training that consists of formal classroom presentations and simulator operations. The requalification training includes approximately.
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per year on the simulator responding to simulated emergencies which includes the TMI-accident scenario. Applicant testified that, during operations, McGuire would use five operational shifts:
3 in shift operation, 1 off duty and 1 in training.
(App. Panel I, p. 6-8, Tr. 2852, 2872-80, 2993-4, 3000-7, 3028-9, 3039 FSAR Sect. 13.2)
- 24. The direct, prefiled written " Testimony of Jesse L. Riley Regarding Hydrogen Generation, Combustion, and Containment Response" with attached "Frofessional Qualifications of Jesse Riley", bound into the record following Tr. 3780, was rejected by the Licensing Board as evidence in this proceed-ing. (Tr. 3967-69 ). The Licensing Board also rejected Mr. Riley's o*al direct testimony given on the record at Tr. 3767-3811, 3816-3824, 3864-3875.
(Tr. 3967-69). The proferred testimony concerned the generation and combastion of hydrogen, effects of such combustion on the containment structure, the McGuire containment design loading, chemical reactions, reactor systems, and reactor operator performance.
The Board concluded,
. that Mr. Riley was not qualified to testify as an expert on matters relating to strer.;th of containment structure, particularly structural engineering aspects, and was also not qualified on hydrogen burning or detonation.
(Tr. 3967, 3969).See also (Tr. 3875-3967).
- 25. CESG's direct case regarding operator training consisted of the testimony of five psychologists. CESG subpoenaed Dr. John Philip Brockway, Dr. Gary Thomas Long, Dr. James Richard Cook, Dr. Edward Leo Palmer, and Dr. John Edward Kello. Their testimony was directed toward general psychological phenomena under certain work conditions, such as boring tasks, information overload, fatigue, stress, cognitive dissonance, group think, risky-shift, obedience to authority, massed and distributive learning, all or none learning, overlearning and amnesia. According to their profes-sional qualification statements, these witnesses had no background with respect to operation of nuclear power reactor facilities and did not relate the general phencmena discussed to nuclear power plant operations or to control room activities.
(Fol. Tr. 3624, 3835, 3845, 3853 and 3978; Tr. 3624-39, 3635-8, 3977-82; App. Exh. 8, Tr. 4674).
26.
In rebuttal, Applicant presented a panel of psychologists, Dr.
Lewis F. Hanes, Dr. Julian ti. Christensen, Dr. Eric F. Gardner, Mr. Robert M. Koehler, and Mr. Richard J. Marzec. These witnesses were able to provide a link between theory and practice. The testimony of this panel of psychologists demonstrated that the psychological phenomena mentioned by CESG's psychologists had been taken into account in the structure of its
4 operator training program, the structure of the control rcom, the organization of the operating personnel, and in the function and duties of the control room operator, or that those conc 3 pts do not apply to nuclear power plant operators or operation. The testimony showed that there is little risk of operater error in a nuclear oower plant that would affect safe operation of the plant from the effects of such psychological phenomena as, cognitive dissonance, risky shift, group think, forgetfulness, information overload, boredom, or mental or physical fatigue. Obedience to authority is considered to be a positive attribute for an operator who is It required as a matter of safety to follow established safety procedures.
was pointed out that application of psychological concepts and theories, developed on an experimental basis, to real world operation of nuclear Dr. Hanes power plant operations must be done with extreme caution.
testified that due to " chunking", an individual with training can handle increasingly greater amounts of information that may seem complex to the outsider. Chunking involves an ability to grasp increasingly large chunks of complex sets of information as training and experience is increased.
(Tr. 4715-4847)
- 27. Stress on operators is not considered to affect the safe operation of the McGuire station. Stress levels in operators of nuclear plants are generally low and the effects of stress on safe operation are reduced to even lower levels by operator training and experience. An, operator is required to follow specific procedures.
Individuals who do not perform well l
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under stressful conditions during the testing and training phases of the operator training and qualification program are eliminated.
Supervisors of operating personnel are trained to recognize any aberrant behavior due to such things as fatigue that might occur in its operators. The training program which mixes training, work experiente, and simulator training is considered to be a good example of the learning concept of " articulation" and how Applicant includes such phenomena as " distributive learning" in its training process.
(Tr.4740-72,4781-99).
Changes have been made in administrative and operating procedures.
28.
They include: (1) a redefining of the Shift Supervisor's primary responsi-bilities and duties to emphasize safe operations of the plant; (2) new shift turnover checklists; (3) more stringent restrictions on overtime; and The.:e (4) more stringent controls on verification of system availability.
changes are intended to reduce the possibility of any operator error which could lead to premature termination of the ECCS. They are designed to enhance the ability of the reactor operators to operate the plant in a safe Operating lines of authority and management responsi-and efficient manner.
bility have been clarified and formalized. procedures have been provided to require verification of safety system availability when systems are removed from or returned to service and notification of operators when Detailed shift safety systems are removed frem or returned to operation.
turnover procedures have been provided, including the use of detailed checklists to assure that current information is provided to the oncoming shift.
(App. panel I, Tr. 2351,2986-93).
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- 29. Significant changes have been made in Applicant's emergency procedures which will reduce the likelihood of premature terminatior of ECCS and the possibility of inadequate core cooling. The revised e.nergency procedures provide specific criteria ;* terminating ECCS operation.
Applicant testified that prior to the iMI-2 accident, emergency procedures required an operator to first identify the accident "?.ich was in progress and then take corrective actions based on the particular procedure for that accident. Subsequent to the TMI-2 accident new emergency procedures have been developed at McGuire requiring operator actions in response to the operator's subjective determination of the accident in progress. Emergency procedures now require that before the operator can terminate the ECCS, four specific criteria must be verified as being within acceptable limits.
The specific criteria are:
(1) Raactor coolant system pressure is greater than a specified m6nimum value and increasing, and l
(2) Pressurizer level is greater than a specified minimum value, and 0
(3) The reactor coolant system is subcooled by greater than 50 F, anc (4) Adequate auxiliary feedwater flow for core heat removal is injected into at least one ncn-faulted steam generator.
If any of the four criteria are not met, the procedure directs that ECCS operations cannot be terminated.
If the four criteria are met, an inadequate core cooling situation cannot exist and generation of exces-I sive hydrogen is impossible. All licensed operators have been thoroughly instructed in the use of such procedures. Operator compliance with these procedures precludes the premature operator termination of ECCS operations thereby preventing inadequate core cooling and excessive hydrogen l
generation.
(App. Panel I, p. 5, Tr. 3013-16) l
] 30. Applicant testified that in the unlikely event of operator premature termination of ECCS operations, emergency p-ocedures require that readings of the four ECCS tennination/reinitiation criteria parameters be continuously checked and recorded in a log every 15 minutes for a two-hour period follow-ing such termination. These log entries must be independently verified.
If any of the parameters are not within acceptable ranges, the procedures require that ECCS operation be reinitiated.
In the event of a TMI-type accident at McGuire, if ECCS operation was prematurely tenninated, the operator would have over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to reinitiate ECCS operation before generating an amount of hydrogen in excess of that produced by a 2% zirconium-water reaction.
Applicant concluded that even in the ire.redible event that ECCS operation was prematurely terminated, it is incredible to further assume that withiit the two-hour period ECCS operation would not be reinitiated prior to genera-tion of hydrogen in excess of that produced by a 2% zirconium-water reaction.
(App. Panel II, Canady, Muench and Barron folicwing Tr. 3045).
31.
In its cross-examination CESG asked a wide range of questions concerning the cause of the TMI event, selection of personnel and manning, operating procedures, and equipment and instrumentation, in addition to operator training. The questioning explored the above areas in some detail.
The cross-examination failed to establish in any of these areas any material basis for concluding McGuire would be operated in such a manner that ECCS would be improperly terminated in a TMI-2 accident.
(Tr. 2865-91, 2920-78, 4385-4480,4520-27).
. 32. The Staff agrees with Applicant and testified that equipment changes, enhanced operator training, technical competence and improved operating procedures have substantially reduced the likelihood of recurrence of an event at McGuire such as TMI-2.
(Staff Ex. K at 3).
In its proposed findings. CESG generally disagrees with these conclusions by Applicant and Staff.
- 33. The Board finds that actions taken by Applicant, subsequent to the TMI-2 accident, are such that in the event of a TMI-type accident at McGuire the likelihood of ECCS operations being prematurely terminated by the control room operating staff is so remote that such an accident ic not credible.
In the unlikely evA't of premature termination of ECCS, the Board finds that current emergency procedures provide reasonable assurance that ECCS will be reinitiated within sufficient time to prevent the generation of hydrogen in excess of the design basis of 10 CFR HEO.44.
IV. ADDITIONAL EVIDENCE PRESENTED 34.
Considerable additional evidence was received in the record of this reopened proceeding on which separate findings of fact have not been made because : Jch findings have now been determined to be unnecessary to the decision. We recognized during the evidentiary hearing at the close of the first phase of Applicant's case that if we held then, as we now hold here, that CESG's failure to establish a credible accident scenario resulting in hydrogen generation leading to offsite doses in excess of 10 CFR Part 100, should preclude CESG from further litigating the issue of excessive hydrogen generation.
Such a ruling would have terminated the hearing at that point. However, in order to build a complete rec.ord and provide for the contingency that careful reflection of the evidence after the close of the hearing would lead us to a finding adverse to Applicant regarding CESG's burden of establishing a credible TMI-type accident scenario at McGuire, we elected to receive the additional evidence offered concerning the McGuire containment response to a TMI-type accident. Pertinent portions of such evidence are summarized below.
Hydrogen Control 35.
Evidence regarding the McGuire Plant response to a TMI-type accident involved five major areas:
containment structural capability; containment systems that mitigate the effects of excessive hydrogen generation; the ignition and burning of hydrogen; transitions to detonation; and pyrolysis of polyurethane foam.
A.
Containment St"uctural Capability
- 36. The McGuire ;ontainment vessel is a freestanding welded steel structure witti a vertical cylinder to which are welded horizontal and vertical stiffeners, a hemispherical dcme and a flat base. The vessel is 115 feet in diameter and 171.25 feet high. ( App. Ex. 5B at 4-4 and 4-7).
- 37. Three independent structural analyses were conducted to determine the McGuire containment vessel functioral capability (the maximum point at which the containment can be reasonaoly assured of retaining its leak tight integrity). Tr. 3746-49. One was undertaken by Applicant and reported in Priory Testimony following Tr. 3654 and in Applicant's Exhibit 5B, Chapter 4.
Based on its analysis, Applicant concluded that the functional capability of its McGuire containment is 67.5 psig.
(Priory Testimony
. following Tr. 3654 at 2).
A second analysis was conducted by Applicant's Consultant, 38.
Mr. R. S. Orr and produced a functional capability figure of 68 psig, essentially the same as the Applicant's calculation.
(Testimony of R. S Orr Regarding the McGuire Structural Integrity, following Tr. 3654 at 1 and 2, Tr. '656).
The third independent analysis was undertaken by Ames Laboratory" 39.
of Iowa State University acting as consultant to the NRC staff and is identified as the Staff Analysis. The predicted ultimate strength of the containment shell for the uniform static pressure case was calculated to have a mean value of 84 psig and a standard deviation of 12 psig.
Because of anticipated deformations in the shell at 84 psig, the Staff considers the mean pressure minus 3 standard deviations,48 psig,is the appropriate lower bound capacity at which leak tightness will be assured.
(Staff Ex. K following Tr. 4353 at 27 through 33).
The probability of containment failure at 48 psi, the lowest of 40.
the functional capacity estimates, was calculated to be 4 x 10-5 per-occurrence. (Tr. 4894,4942-43). This multiplied by the probability of a TMI type accident (10 5 to 10-6 per year) results in an overall probability of failure due to a TMI-type accident of 10-10 to 10-11 per reactor year.
(CESG Ex. 61; Tr. 4943-45).
i CESG presented testimony of Joe E. Lanford, a structural engineer, 41.
and former Duke Power Company employee, who testified that he observed excessive grinding around a weld on one of the McGuire containment structures causing a gauge in the base metal of about 1/8 inch deep and
23 -
several inches long. He brought this to the attention of Duke officials but did not know whether any corrective action was taken. Mr. Lanford opined that if such a gauge had not been repaired, "the containment would be, to some degree large or small, depending on the damage, weakened from the design load capacity."
(Tr. 3827-32).
- 42. Testimony frcm Applicant and Staff witnesses failed to specifi-cally identify Mr. Lanford's alleged faulty weld. Staff witness Herdt opined that it was not established whether the weld was completed or not.
If the weld was not completed (i.e., pre-inspection) and a gouge was noticed, the welder could repair that section and no records of a gouge and/or repair would appear on any Q/A records.
(Tr.4980-81).
43.
Both Applicant and Staff witnesses described the welding inspection and Q/A procedures.
(Tr. 4847-4853; Tr. 4971, 4972).
Significant among the procedures are the required visual and radiographic inspections of every butt weld.
(Tr. 4847-4852). Visual inspection is conducted with an instru-ment capable of detecting width changes in the metal thickness to 1/32 of an inch.
(Tr. 4859). Radiographs would detect changes in metal as small as 15/1000 of an inch.
(Tr. 4861).
Both Applicant and Staff testified that excessive grinding as described by Mr. Lanford would or should have been detected in both visual and radiography inspections and repaired.
(Tr. 4850-2; 4972,4981-2).
f
+
-._o
. The record further shows that, even if the gouge described by 44.
Mr. Lanford as 1/8 inch deep and several inches long was, for whatever reason, undetected and not repaired, it would have an insignificant effect on containment capability. Because of its ductility, steel can tolerate small imperfections such as the gouge described by Mr. Lanford.
(Tr. 4896).
B.
Hydrogen Mitigation System Description During the TMI event, hydrogen released to containment was ignited 45.
This resulted in a by existing ignition sources within containment.
pressure spike of 28 psig. The emergency hydrogen mitigation system (EHM) installed at McGuire consists of 62 igniter assemblies (46 in the lower compartment, 8 in the ice condenser upper plenum, and 8 in the upper containment). The igniters are designed to initiate hydrogen gas burning at low concentrations thereby preventing gas buildup to the detonable range.
The igniters work in combination with other containment systems, including the ice condenser system, the containment air return system, the hydrogen skint:er system and the containment spray system.
(Testimony of David L. Canup following Tr. 3488).
C.
Ignition and Burning of Hydrogen Applicant's witnesses testified that the ignition and burning of 46.
hydrogen generated in a TMI-type accident will occur by: (a) a continuous burn at the top of the ice condensers; (b) a series of burns initiated in the lower containment; or (c) a combination of (a) ani (b). They further said that the Case (a) scenario is the likely scenario and the pressure rise in containment resulting from that would be only a few psi. (Tr. 3353-7).
y y
...-,,.9
. Of the three scenarios, the peak pressure in containment would result from a Case (b) scenario (multiple burns of hydrogen in the lower compart-men t).
(Testimony of William Rasin, David Goeser, Bela Karlovitz, Bernard Lewis and Edward McHale Regarding Hydrogen Generation and Ignition following Tr. 3144 (Lewis Panel) at 10).
- 47. Applicant analyzed the Case (b) scenario. (Id. at 2, 3). The actual conditions assumed by Applicant for this scenario are described in the testimony and involve a small-break loss-of-coolant accident assuming failure of ECCS and a 75-80% zirconium-water reaction (1550 pounds of hydrogen). (ld.; Applicant's Ex. 5A at 2-3 through 2-5 and Ex. 5B at Table 6, Accident Scenario JVD 12; Tr. 3202-3).
Applicant's analysis of the Case (b) scenario, the most severe of 48.
the three cases considered, resulted in a peak pressure of less than 16 psig.
(Lewis Panel at 2, 3). This is considerably below the containment shell functional capability as determined by three independent structurcl analyses. See Containment Structural Capability SIV. A. Suora.
D.
Transitions to Detonation 49.
Steam inerting of hydrogen and transitions to detonation received l
considerable attention. Certain of the results of hydrogen ignition tests conducted by Livermore, raised the possibility of hydrogen not burning in lower steam concentrations than generally reported.
(Tr. 3210-18; Staff Ex. K following Tr. 4353 at 15, 16). Dr. Marshall Berman of Sandia National Laboratory, a Staff consultant voiced concern over the possible steam inerting of the lower compartment, followed by removal of the steam in the i
. ice condensers, resulting in high concentrations of hydrogen in the ice condenser possibly resultingin detonable mixtures. (Tr. 4083). Below about 18 volume percent, hydrogen is not detonable.
(Tr.3155,3260).
Both CESG concentration and gecmetry are involved in detonation considerations.
maintained that the configuration of the ice condenser was such that analogies could be made to situations where detonations have been observed to occur.
References were made to the concern of Dr. Roger A. Strehlow (Tr. 3412) and through Dr. Berman, the work of Dr. John Lee of McGill University, a Sandia consultant.
(Tr.4199-200).
Applicant's testimony regarding Dr. Strehlow's concern clearly 50.
points out that even if higher concentrations were present, the geometry necessary for transition to detonation does not exist in the ice condenser.
The testimony reflects that there are no long, narrow confined passageways in the ice condenser; rather the area between baskets is open, there are holes along the baskets and that the lattice frame would not confine any substance from flowing through the condenser. (Tr. 3489-94). CESG also raised the possibility of transition to detonation in the ductwork of the air handling units. The record reflects that the necessary geometry is not there.
(Tr. 3613-7).
Dr. Berman's concerns of detonations in the upper plenum were 51.
strongly influenced by Dr. John Lee's experiments using stoichiometric concentrations of ccmbustible gas (propane or methane) in large tubes open at one end with periodic obstacles (baffles at various distances) inside the tube wherein substantial overpressures (detonations) occurred.
(Tr. 4095-97 as modified at 4243). Applicant's Witnesses, Lewis and l
~ ~..
. Karlovitz, after having consulted with Dr. Lee by telephone, testified that the open, unconfined geometry of the upper plenum of the ice condenser was such that the results produced by Dr. Lee could not be produced in the upper plenum region.
(Tr. 5050, See also 5057-8). Dr. Lewis further stated that if concentrations of hydrogen and steam in the lower compartment began to increase rapidly such that steam inerting occurred, the resulting concentration of hydrogen flowing through the ice condenser would increase correspondingly.
(Tr.5051-6,5102). When the hydrogen / air mixture passing through the ice condenser enters the upper plenum and reaches flammable concentrations, ignition and burning would occur. As the concentration of hydrogen flowing through the ice condenser increased, the flame would settle into the ice condenser at a level where such mixture was just flammable and continue to burn. Thus, due to the presence of igniters in the upper "lenum regions there would never be a detonable mixture of high hydrogen concentra-tion in the upper plenum region. (Id.) The scenario posed by Dr. Berman and his interpretation of Dr. Lee's work is not probable on the bases of hydrogen concentration or necessary gecmetry.
(Tr. 5050-1).
E.
Pyrolysis of Polyurethane Foam CESG raised a question concerning the effect of pyrolysis on 52.
l deccmposition of polyurethane foam insulation as a result of hydrogen burning
(
The ice condenser is insulated from the containment in the ice condenser.
wall with 27,000 pounds of polyurethane covered with sheet steel.
Applicant presented a panel of witnesses consisting of Dr. Lewis, Mr. Rasin and Dr. Leonard S. Edelman. Dr. Lewis testified that, assuming a TMI-type accident which resulted in generation of hydrogen and a continuous r
c
9 burn in the ice condenser, the flame temperature inside the ice condenser would be theoretically about 1400 F(an 81% hydrogen concantration) with a flame height of at most about I centimeter. Within a few feet upward the hot gases would have cooled to about 400-5000F. Dr. Edleman testified to the characteristics of the polyurethane and the effect of heat on the foam.
Applicant perfomed a conservative heat transfer analysis using the flame and hot gas temperature noted by Dr. Lewis. The analysis was conservative because it ignored the effect of ice in the condenser and assumed a 6 inch flame. The result was the volatization of 250 pounds of foam and the addition of 3 x 106 Btu of heat energy, which would have an insignificant effect on containment pressure.
(Tr. 5136).
Applicant testified that if the foam was not enclosed but totally exposed to oxygen and a flame, it would burn as long as there was sufficient oxygen to support combustion.
In the sealed ' configuration at McGuire there is little oxygen to support combustion and the foam itself does not generate free oxygen for such support.
An experiment conducted in such a configuration resulted in an inability to sustain ccmbustion even when a large opening was present. (Tr. 5041, 5068-77, 5106-41,5145-60,5180-92,5215-23).
53.
In cross-examination, CESG detemined that the amount of oxygen initially present in the containment is only suf ficient to burn about 9,000 lbs. of the 27,000 lbs. of polyurethane initially present in the i
containment. The net volume of the ccmbustion gas would be 100,000 cf at standard conditions.
If the heat of combustion were removed, the presence I
of this gas would cause a pressure increase of about 8% in the containment.
If, instead of burning, the polyarethane wore to completely gasify, it would result in 250,000 cf at standard conditions and an increase in containment
. pressure of about 20%, about 3 psi. The air handling ducts are not leak-tight and some of the volatilized gases frem the pyrolysis were assumed to be released into the ice condenser atmosphere where they could contribute to the burning already occurring there. The heat of the combustion of the foam at 12,000 Btu's/lb was used as an approximation of the energy contribu-tion. The resulting total energy contribution to the containment would be a maximum of 3 million Btu's, which is compared to the 80 million Stu's contri-buted by hydrogen burning under the S2D scenario. This additional energy contribution will not significantly increase the total pressure rise in containment.
(Tr. 5121-60, 5218-25).
- 54. At the close of the evidentiary hearing, the Staff asked for additional time to review the record on pyrolysis and, if necessary, file testimony on that subject. The Staff filed an affidavit which has been admitted as Staff Exhibit Q.
The Board permitted the parties the opportunity to respond to Staff Exhibit Q.
Applicant filed an affidavit which has been admitted as Duke Ex. 9.
CESG filed an affidavit (Ex. 63) on the same date as Staff but not in response to Staff Ex. Q.
In Ex. Q, the Staff found the information sufficiently complete in the record and did not have substantial information to add. The Staff has no "urther concerns about the pyrolysis of foam.
(Tr.5245-6,5252).
I
. V.
10 CFR Part 2 - Apoendix B
- 55. The Atomic Safety and Licensing Board has heard and decided, as necessary, all issues that have come before it. But for the provisions of Appendix B to 10 CFR Part 2, we would authorize the Director, Office of Nuclear Reactor Regulation, upon making requisite findings witn respect to matters not embraced in our Initial Decision of April 18, 1979 or in this Supplemental Initial Decision, to issue full-term, full-power, operating licenses to McGuire Nuclear Station, Units 1 and 2.
- 56. In our analysis of the evidence, we have not identified any serious, close questions which we believe may b> crucial to whether a license should become effective before full appellate review is completed. However, the Board has heard considerable testimony regarding ongoing research concerning hydrogen mitigation systems and has taken official notice of the contents of the program instituted by TVA to demonstrate by January 31, 1982 that an adequate hydrogen control system is installed at the Sequoyah facility and will perform its intended function in a manner that provides adequate safety marginsU. The staff has taken the position that this requirement should also apply to McGuire, and we concur that such a requirement is appropriate with respect to the matters which we have considered in this proceeding.
Since i
1 the decision to be reached by January 31, 1982, however, also entails matters beyond the scope of this proceeding (e.g., censideration of results of refined CLASIX calculations, results of verifications of equipment survivability, etc.),
- / "Research Program on Hydrogen Combustion and Control Quarterly Progress Report", Tennessee Valley Authority - Sequoyah Nuclear Plant, December 15, l
1980 (Tr. 5227).
l
. we consider such a requirement to be within the purview of the Cirector of the Office of Nuclear Reactor Regulation.
VI. CONCLUSIONS OF LAW
- 57. In an operating license proceeding, the Board is called upon to decide only the issues in controversy among the parties. 10 CFR 52.760a. Other matters required to be determined prior to the issuance of an amendment to the zero-power operating license for Unit I authorizing full-power operation or of an operating license for Unit 2 are entrusted to the Director of the Office of Nuclear Reactor Regulation. 10 CFR 552.760a, 50.57.
- 58. Based upon the foregoing findings of fact, which are supported by reliable, probative, and substantial evidence in the record, and upon considera-tion of the entire evidentiary record in this reopened proceeding, the Board makes the following Conclusions of Law in supplementation of the Conclusions of Law reached in its April 18, 1979 Initial Decision:
(1) There is reasonable assurance tnat in the event of a TMI-type accident at McGuire, substantial quantities of hydrogen (in excess of the design basis of 10 CFR 550.44) will not be generated.
(2) As to Contentions 1 and 2, the actions taken and the procedures adopted by Duke Power Company subsequent to the TMI accident, provide reasonable assurance that (a) in the event of a TMI-type accident at McGuire, the likelihood of ECCS operations being prematurely terminated by the control rocm operating staff is so remote that such an accident scenario is not credible; (b) in the unlikely event of premature termination of the ECCS, operations will be reinitiated within sufficient time to prevent the generation of hydrogen in excess of 10 CFR 550.44; and (c) the McGuire facility can be operated wittnut undue risk to the public health and safety with respect to possible hydrogen generation resulting from accidents of the type which occurred at TMI-2.
i (3) Because the Board has found that quantities of hydrogen in excess of the design basis of 10 CFR 650.44 will not be generated, breach of containment and offsite doses in excess of 10 CFR Part 100 guideline values resulting from hydrogen combustion in a TMI-type accident at McGuire are not credible events. Accordingly, the premise for CESG Contentions 3 and 4 has not been established and there is no need to make specific findings with respect to those contentions.
(4) The NRC Staff has issued Supplement 3 to the McGuire Safety Evaluation Report (Staff Exhibit H) which addresses the significance of the unresolved generic safety issues as they relate to the McGuire facilities and has provided a reasonable foundation for its several conclusions in conformity with the Board's April 18, 1979 Initial Decision.
VII. ORDER WHEREFORE, IT IS ORDERED that the stay of the Licensing Board's April 18, 1979 Initial Decision is lifted and the the Director, Office of Nuclear Reactor Regulation, is authorized, upon making requisite findings with respect to matters not embraced in the Initial Decision of April 18, 1979 or this Supplemental Initial Decision, in accordance with the Comission's regulations, to issue to Duke Power Ccmpany operating licenses (or in the case of Unit 1, an amendment to NPF-9, if appropriate) for a tem of not more than forty (40) years, authorizing operation of the McGuire Nuclear l
Station, Units 1 and 2, at steady state power levels not to exceed 3,411 megawatts thermal; such licenses may be in such form and content as is appropriate in light of such findings.
In view of the Ccmission's Rules of Practice limiting the Board's jurisdiction in a contestac nerating license proceeding, the Scard has I
made findings of fact and conclusions of law on matters actually put into l
t controversy by the parties to the proceeding.
In addition, the licenses will not be issued until the Director, Office of Nuclear Reactor Regulation I
t y
a
33 -
has made the findings reflecting its review of the application under the Atcmic Energy Act, which will be set forth in the proposed licenses, and has concluded that the issuance of the licenses will not be inimical to the comon defense and security and to the health and safety of the public.
Further, the licenses will not be issued until directed by the Commission after the appropriate Appendix B to 10 CFR Part 2 stay review process, if such is applicable.
Exceptions to the Initial Decision of April 18, 1979 and to this Supplemental Initial Decision and requests for a stay may be filed within 10 days after the service of the Supplemental Initial Decision. A brief in su:: port of the exceptions should be filed within 30 days thereafter (40 days in the case of the Staff). Within 30 days after the service of the brief of appellant (40 days in the case of the Staff) any other party may file a brief in support of, or in opposition to, the exceptions.
THE ATOMIC SAFETY AND LICENSING BOARD mnW W.
Jl, Emetn A. Luebke 9
ADMINISTRATIVE JUDGE
!r -
Richard F. Cole ADMINISTRATIVE JUDGE N. <:v2X.
Robert M. Lazo, Chairrfany -
Issued at Bethesda, Maryland, ADMINISTRATIVE JUDGE this 26th day of May, 1981 O
p
APPE?lDIX A Decisional Record The decisional record in this proceeding consists of the following:
1.
The material pleadings filed herein, including the Commission notices, the petitions and other pleadings filed by the parties and the orders issued by the Board during the course of this proceeding; 2.
The transcript in this proceeding. The transcript of testimony at the evidentiary hearings is in fifteen volu ies with pagination from 2674 to 5257; 3.
The decisional record regarding the Initial Decision issued on April 18, 1979.
4.
The exhibits received into evidence at the evidentiary hearing.
These exhibits are identified as follows:
STAFF EXHIBITS Number Identified Received Descriction E
3160-61 Letter from Marshall Berman, I
4441, Sandia flational Laboratories, to Thomas E. Murley and Denwood F.
Ross, February 9,1981 (Appendix B),
l pp. 1-5.
F 3558 Memorandum for Comissioners Ahearne, Gilinsky, Hendrie and Bradford: subj:
OPE Review of Hydrogen Control Measures
@g for Sequoyah (dtd January 22,1981),
y g
4 encl., " Evaluation of the Glow Plug Igniter Concept for Use in the Sequoyah CCcum 2 M
c'uclear Plant, prepared by Roger A.
l 7,
t,33,,~
t
,MY 2 619M
- Strehlow, Consultant ("Strehlow Report A
l January 9, 1981).
d Cfc:3 :f as smin
/
C = c ; ; 2e.cic3
/
s g/
i s
\\ -a "eNL-
' 1l' q tyg/
l
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STAFF EXHIBITS - (Continued)
Number Identified Received Descriotion G,H,I 4006 4006
" Safety Evaluation Report",
Duke Power Company, McGuire Nuclear Station, Units 1 & 2, Supplement No. 2, 3, & 4.
J 4006 4006
" Final Environmental Statement",
Duke Power Company, William B.
McGuire Nuclear Station, Unit 1 & 2, NUREG-0063, addendum, January 1981.
K 4007 4353 "NRC Staff Analysis of Hydrogen Control Measures", William B.
McGuire Nuclear Station, Unit 1 &
2, Docket Nos. 50- 369 and 50-370, February, 1981.
L 4297 Figure 7, " Obstacle Configurations" (Diagram).
- ti 4656 4663 R & D Associates Report (February 1981).
N 4546, 4656 Environmental Qualifications SER, February 24, 1981.
0 5125 5126 Figure 22.2-2B, Supplement #4 to the
" Safety Evaluation Report" for Sequoyah, NUREE-0011.
P 5174 5175 Figure 22.2-2A, Supplement #4 to the Safety Evaluation Report" for Sequoyah, NUREG-0011.
Q Motion to (I.D.)
Joint Affidavit of Vincent S. Noonan, Supplement Harold E. Polk, Krysztof I. Parczewski, Record w/ joint and Charles G. Tinkler (March 27,1981) affidavit of (submitted by staff counsel's letter of staff members March 27, 1981 in accordance with leave Noonan, Polk, granted by the Licensing Board at the Parczewski and close of the hearing on March 19,1981).
i l
Tinkler submitted nunc cro tunc on April 8, E i
i Not admitted for the truth of the matter contained therein, and cannot i
be the basis for proposed findings, Tr. 4663.
(
i
APPLICANT'S EXHIBITS Number Identified Receive Descriotion 5A,B,C,0 3118 3118 "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station",
Volumes 1, 2, 3, & 4.
SE 3298 3298 Applicant's errata sheet to Volume 2 of a 4 Volume document.
6 3336 3337 Three page document entitled " Air Return Flow Paths to Lower Containment".
7 3490 3618 One page document entitled " Lattice Frame".
8 4674 4674 Stipulations Regarding CESG's Psychologists, March 17, 1981.
9 (I.D.)
(I.D.)
April 13, 1981 Affidavit of R. W. Rasin INTERVENOR"S EXHIBITS 40 3215 4654 Sandia Draft letter from Marshall Berman, Sandia National Laboratories, to Thomas E. Murley and Denwood E. Ross, United States Nuclear Regula:ory Commis-sion, February 9,1981, and attached report.
40A 4670 4670 Sandia Final letter frem Marshall Berman, Sandia National Laboratories to Thomas E. Murley and Denwood F.
Ross, United States Nuclear Regulatory Commission, February 9, 1981, and l
attached report published March 6, 1981.
41 3482 3482 Affidavit of Stephen P. West, March 2, 1981.
" Analysis of the Three Mile Island 42 3782 Accident and Alternative Sequences,"
NUREG/CR-1219, pp. v-vi, 1-1 to 1-4, 2-1 to 2-8, prepared by R. O. Wooton, R. S. Denning, P. Cybulskis, Battelle, Columbus Laboratories, prepared for U.S. Nuclear Regulatory Commission (NUREG/CR-1219).
'l i
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1 INTERVENOR'S EXHIBITS - (Continued)
Number Identified Received Description 43 3782 NUREG/CR-1219, Cp. 5, " MARCH Analysis of Alternative Accident Sequences,"
pp. 5-1&5-18; Figs. 5.3, 5.4, 5.5.
44 3784 NUREG/CR-1219, Cp. 8, Analysis of Hydrogen Burning During the TMI-2 Accident, pp. 8-1 to 8-8.
45 3785-87 Proposed testimony of A. D. Miller Regarding Hydrogen Production at TMI (includes Hydrogen Phenomena (Appen-dix HYD) NASAC-1, pp. 1-11, Figs.
HYD-1 to Figs. liYD-6.
46 3786-87 Report from Harmon W. Hubbard, R&D Assoc. to Nuclear Regulatory Ccmission, dated August 4,1980 with encl. entitled
" Hydrogen Problems in Sequoyah Contain-ment."
47 3787
., Memorandum from W. R.
Butler to W. C. Milstead, " Commission Paper - SECY 80-107 'Procosed Interim Hydrogen Control Requirements for Small Containments.'"
48 3789 Figure TH10, Nozzle Locations; Fig. TH9, Primary System Reactimeter Measurement; Fig. THil, Drain Tank Behavior.
49 3790 NSAC-1, Appendix PDS, pp. 12-14 50 3792, Memorandum for R. L.
Tedesco, from W. Butler, subj: Three Mile Island, Unit 2: Analysis and Evaluation of Selected Containment Related Issues, April 25, 1979.
51 3793 Letter from Harman W. Hubbard to Nuclear Regulatory Comission, dated July 25, 1980 with. enc 1.
entitled, "Sequoyah Containment Analysis," July 1980, pp. 1-5, 7-22.
--n p.
e,m s
e + -
INTERVENOR'S EXHIBITS - (Continued)
Numbe-Identified Received Descriotien 52 3793 Memorandum frem P. Tedesco to L.
Rubenstein, datec December 1, 1980.
53 3793-94 248th ACRS Meet'.ng, Tr. 339, 288-90, 404-05.
54 3805 NSAC-1, Appendix ERV, "Electrematic Relief Valve," pp.1-5.
55 3808 NSAC-1, Appendix PDS, " Plant Data Sources for Three Mile Island Unit 2 Accident," pp. 1-6.
56 3808 NSAC-1, Fig. OTSG-1; Fig. OTSG-2 (OTSG Level Indication) (Appendix DTSG); Fig. RCPCS-1 (Pressuri:er Layout) (Appendix RCPCS).
57 3810 Science News, Vol. 118 " Nuclear Accidents: The 10-Minute Myth," p. 37 (withdrawn 4636)
(July 19,1980) and letter fecm Themas B. Sheridan, Massachusetts Institute of Technolcgy, Ce;t. of Mechanical Engineerine to Mr. Jesse Riles [ sic], cated Feo.ary 25, 1981.
58 3818 "NRC Staff Answers to CESG Interroga-tories and Recuests for Cocuments" (January 16,1981).
- 59 3822 4654 Memorandum to R. A. Beri frcm W. T. Pratt, Brookhaven National Laboratories, "Seme Very Preliminary Results of a Short-Term Analysis (3 week study) of Hydrogen Ccmbustion during Cegraded Core Accidents in the Sequoyah Nuclear Plant in the Presence of Glow Plugs", January 15, 1981.
Not acmitted for the truth of the matter contained therein, c..i cannot be the basis for proposed findings, Tr. 4663.
,9
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ItiTERVENOR'S EXHIBITS - (Continued)
Number Iden+.ified Received Descriotion 60 3823 4654 One page from the Encyclopedia Britainnica, describing the meaning of solder.
61 4523 4526 Table 9-1, Dominant Accident l
Sequences - Sequoyah Plant (one page).
62 4881-81 Chapter 8, Accident Process Analysis (Sequoyah) 63 (I.O.)
(I.D.)
March 27, 1981 Affidavit of Jesse L.
Riley 4
4
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