ML20117P422
| ML20117P422 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/03/1985 |
| From: | Glasspiegel H CLEVELAND ELECTRIC ILLUMINATING CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | |
| References | |
| CON-#285-292 OL, NUDOCS 8506060219 | |
| Download: ML20117P422 (106) | |
Text
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DOCKETED June 3,Ul %
UNITED STATES OF AMERICA WN -5 All :00 NUCLEAR REGULATORY. COMMISSION
- GFFICE cr SECRr:ThR>
00CMEilNG & SEPVM,F' BEFORE THE ATOMIC SAFETY AND LICENSING BOARD . BRANCH In the Matter of )
THE CLEVELAND ELECTRIC Docket Nos. 50-440 b ILLUMINATING COMPANY, ET AL. ) 50-441'
)
(Perry Nuclear Power Plant, )
Units 1 and 2) )
APPLICANTS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF A PARTIAL INITIAL DECISION (HYDROGEN CONTROL)
Jay E. Silberg, P.C.
Harry H. Glasspiegel SHAW, PITTMAN, POTTS & TROWBRIDGE Counsel for Applicants y 1
7 O b TABLE OF CONTENTS Page OPINION..................................................... 2 I. HISTORY OF THE CASE.................................... 2 II. THE CONTENTION......................................... 3 A. Background and Introduction....................... 3 B. The Hydrogen Rule................................. 10 C. Witnesses......................................... 15 D. Hydrogen Control System........................... 21 E. Igniter System Selection and Preliminary Evaluation............................ 22 F. Igniter System Design and Operation............... 25 G. Ultimate Structural Capacity of Containment....... 27 H. Containment Response and Equipment Survivability..................................... 29 I. Summary and Conclusion............................ 36 FINDINGS OF FACT............................................ 37 I. CONTENTION AND WITNESSES............................... 37 II. HYDROGEN CONTROL SYSTEM................................ 46 A. Combustible Gas Control System.................... 46 B. BWR System Features to Minimize Risk of Large Hydrogen Releases........................... 48-C. Igniter System Selection and Preliminary Evaluation............................ 51 D. Igniter System Design and Operation............... 56
o[c 5 Page E. Ultimate Structural Capacity of Containment....... 62 F. Containment Response and Equipment
. Su rviv ab i li ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 CONCLUSION OF LAW........................................... 91 ORDER.......................................................92 APPENDIX A: Written Testimony Received Into Evidence APPENDIX B: . Exhibits
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June 3, 1985 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION.
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
'In'the-Ma'tter of
)
)
THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY, ET AL. ) 50-441
)
(Perry Nuclear Power Plant, )
Units 1 and 2) )
APPLICANTS' PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF A PARTIAL INITIAL DECISION (HYDROGEN CONTROL)
Pursuant to 10 C.F.R. 52.754(a)(1), The Cleveland Electric Illuminating Company ("CEI"), acting for itself and as agent for Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and The Toledo Edison Company (collectively re-ferred to herein as " Applicants"), submits in the -fonn of a partial initial decision Applicants' proposed findings of fact and conclusions of law relating to Issue No. 8 (hydrogen con-trol) in this proceeding., The proposed findings of fact and conclusions of law follow the form prescribed by the-Atomic Safety and Licensing Board (the " Board").1/
1/ Memorandum and Order (Proposed Findings and Conclusions)
(April 18, 1985).
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OPINION I. HISTORY OF THE CASE This is the third partial initial decision in this con-tested proceeding on the application for operating licenses for the Perry Nuclear Power Plant ("PNPP"). The Board has, in its second partial initial decision, set out a detailed description of the history of the case.2/ There is no need to restate the history of the case here, except as to Issue No. 8 (hydrogen control), which is the subject of this partial initial deci-sion. Direct testimony on Issue No. 8 was filed on April 1, 1985. The hearing was held on April 30 and May 1-3, 1995, in Perry, Ohio.3/
The decisional record of the proceeding for the third phase consists of the testimony and exhibits filed by the par-ties, and the other evidence contained in the transcripts of the hearing.4/ In preparing this decision, the Board reviewed 2/ See Applicants' Proposed Findings of Fact and Conclusions of Law in the Form of a Partial Initial Decision (Emergency Planning and TDI Diesel Generators) (May 13, 1985), at 2-7. ,
3/ On May 1, 1985, the Board conducted a site tour to observe the PNPP igniters, diesel generators, control room, and related areas of the plant. See Tr. 3214 (Board).
4/ Appendix A to this partial initial decision identifies, by witness, the location of written testimony in the transcript.
Appendix B lists the exhibits identified, indicates the Board's ruling on any offer of an exhibit into evidence, and identifies the location of admitted exhibits in the transcript.
_ o. 4 and' considered the entire record and the proposed findings of fact and conclusions of law submitted by the parties. Those proposed findings.and conclusions that are not incorporated di-rectlyLor by inference in this partial.' initial decision are re-
.jected as being unsupported by the record of the case or as being unnecessary to the rendering of this decision.
This Board's jurisdiction is limited to a determination of findings of. fact and conclusions of-law on matters put into controversy;by the parties to the proceeding or found by the Board to involve a serious safety,. environmental, or common de-fense and security question.5/ The Board has made no such ad-ditional determinations in this case.
II. THE CONTENTION A. Background and Introduction Issue No. 8 has a long and somewhat complicated history in this' proceeding.6/ The_ contention was originally submitted by Sunflower Alliance, Inc. (" Sunflower") in its March 5, 1981 Pe-tition for Leave to Intervene. Sunflower's proposed contention 5/ 10 C.F.R. $2.760a. See Houston Lighting and Power Co.
(South Texas Project, Units 1 and 2), LBP-81-54, 14 N.R.C. 918, 922-23 & n.4 (1981).
6/ A summary of the background of the hydrogen control issue is contained in-the Board's Memorandum and Order (Motions on Hydrogen Control Contention) (March 14, 1985), at 1-3. This section of the' partial initial decision expands somewhat on that description.
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0 t sought to raise a contention on the ability of PNPP "contain-ment structures" to' withstand a hydrogen explosion such as oc-curred at Three Mile Island, Unit 2. Both Applicants and the NRC Staff opposed the admission of the contention. In the spe-cial prehearing conference memorandum and order,7/ the Board denied the contention's admission on the grounds that a "cred-ible loss-of-coolant accident scenario entailing hydrogen gen-eration and combustion" had not been alleged by Sunflower as was required by Metropolitan Edison Co. (Three Mile Nuclear Station, Unit 1), CLI-80-16, 11 N.R.C. 674, 675 (1980)
("TMI-1"). The Board invited the Intervenor to file a late contention-addressing the TMI-1 requirements.g/
Sunflower subsequently moved to resubmit its hydrogen con-trol contention in an attempt to meet the TMI-1 standard.
Applicants and the NRC Staff both opposed admission of Sunflow-er's resubmitted late contention. They argued that Sunflower failed to specify a detailed accident scenario which met the TMI-1 standard, and failed to comply with the Commission's late-filing criteria in 10 C.F.R. $2.714. The Board granted Sunflower's motion over these objections, and reworded Sunflower's resubmitted contention to read as follows:
7/ Special Prehearing Conference Memorandum and Order Con-cerning Party Status, Motions to Dismiss and to Stay, the Ad' missibility of Contentions, and the Adoption of Special Discov-ery Procedures, LBP-81-24, 14 N.R.C. 175, 207-209 (1981).
g/ LBP-81-24, 14 N.R.C. at 208.
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o o Applicant has not demonstrated that the manual operation of two recombiners in each of the Perry units is adequate to as-sure that large amounts of hydrogen can be safely accommodated without a rupture of the containment and a release of substan-tial quantities of radioactivity into the environment.9/
The Board's admission of the contention was influenced by the
. Commission's issuance of a proposed rule dealing with hydrogen control for Mark III BWR plants such as PNPP. 46 Fed. Reg. 62281 (December 23, 1981).
Applicants filed a motion with the Appeal Board for directed certification of the Board's decision in LBP-82-15 ad-mitting the hydrogen contention. Applicants argued that the admission of the contention was contrary to the Commission's holding in TMI-1, and was inconsistent with Potomac Electric Power Co. (Douglas Point Nuclear Generating Station, Units 1 and 2), ALAB-218, 8 A.E.C. 79, 85 (1974). The Appeal Board de-nied Applicants' motion as interlocutory.10/
By Memorandum and Order (Concerning Procedural Motions),
dated September 17, 1982, the Board designated Ohio Citizens for Responsible Energy ("OCRE") as the lead intervenor for Issue No. 8.
9/ Memorandum and Order (Concerning Late-Filed Contentions:
Quality Assurance, Hydrogen Explosion, and Need for Increased Safety of Control System Equipment), LBP-82-15, 15 N.R.C. 555, 563 (1982).
10/ ALAB-675, 15 N.R.C. 1105 (1982).
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9 Discovery on this issue closed on September 30, 1982. See Tr. 753. As of that date, OCRE submitted one set'of interroga-tories and request for production of documents to Applicants and one set to the NRC Staff. Applicants submitted one set of interrogatories and request for production of documents to OCRE.
. In December 1982, the NRC Staff moved for reconsideration and dismissal of Issue No. 8, on the grounds that OCRE had not yet specified a credible accident scenario which could lead to a TMI-2 type event at PNPP, as required'by the Commission's TMI-1 decision. The Board denied the Staff's motion as untime-ly.11/
In February 1983, the NRC Staff and the Applicants re-quested the Board to set a deadline for OCRE to identify the TMI-2 type scenario which it intended to litigate under Issue No. 8. The Board denied the request.12/ The Board, noting the upcoming QA hearing, determined that the specification of a scenario at that time was not necessary to assure the orderly progress of the proceeding.
11/ Memorandum and Order (Concerning Reconsideration and Dis-missal of Hydrogen Control Contention), LBP-82-110, 16 N.R.C.
1895 (1982).
12/ Memorandum and Order (Staff's Motion to Establish a Dead-lina Concerning a Hydrogen Generation Scenario) (March 3, 1983).
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.Also in February 1983, OCRE moved to reword its contention and to defer its consideration until after_the then pending
, rulemaking. dealing with hydrogen control was completed. Appli-l cants and the NRC Staff opposed OCRE's motions. The Board deferred further action.on OCRE's motions.13/ The Board be-lieved that a final hydrogen control rule was about to issue, I
suus saw no reason to take -further action.
In November 1983, OCRE filed a motion to reopen discovery, l which was opposed by Applicants and the NRC Staff. The Board i
L denied OCRE's motion, finding that the. discovery schedule was I. fair and reasonable, and that OCRE had'not shown good cause for extending it. The Board did require Applicants to file supple-
! mental answers to OCRE's earlier interrogatories.14/
In July 1984, OCRE again moved to reopen discovery. OCRE L
filed extensive new interrogatories on Applicants with its mo-tion, many of which were answered by Applicants on a voluntary basis.
l The Commission's final rule containing hydrogen control requirements for BWR plants with Mark III containments was pro-
, mulgated on January 25, 1985. 50 Fed. Reg. 3498 (1985) (to be
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codified at 10 C.F.R. Il 50.44(c)(3)(iv)-(vii)) (the " hydrogen l 13/ Memorandum and Order (Applicant's Answer to Procedural Mo-l tion Concerning Hydrogen Control) (March 31, 1983).
1 14/ Memorandum and Order (OCnE Motion to Reopen oiscovery)
(December 20, 1983).
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. rule"). On January 23, 1985, OCRE moved to reword Issue:No. 8, and proposed new language based on specific provisions of the hydrogen rule. Applicants.and the NRC' Staff opposed OCRE's mo-tion. -The Staff also filed a motion for summary disposition of Issue No. 8, which OCRE opposed. The Board denied summary dis-position but approved OCRE's motion to reword Issue No. 8, mod-ifying OCRE's proposed wording to read'as follows:
The Perry hydrogen control system is inade-quate to assure that large amounts of hy-
'drogen can be safely accommodated without a rupture of the containment and a release of substantial quantities of radioactivity to 5
the environment._1_5/
In a subsequent opinion explaining its rulings,11/ the Board provided the basis for its admission of the reworded con-tention. The Board explained the reworded contention as basi-cally alleging that Applicants' hydrogen control system does.
not conform to the new regulatory requirements of 10 C.F.R.
$50.44, as-enacted in the hyd agen control rule.12/ The Board recognized that the new rule included certain implementation schedules, but concluded that it was not necessary to decide at that time which of these provisions were applicable.lg/
11/ Memorandum and Order (Motions) (March 13, 1985), at 2.
lj/ Memorandum and Order (Motions on Hydrogen Control Conten-tion) (March 14, 1985).
12/ Id. at 6-7.
11/ Id. at 6.
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On March 18, 1985, OCRE moved to compel the appearance at the hearing of Dr. Marshall Berman, of Sandia National La-boratories. Sandia has been engaged in ongoing research on hy-drogen control for the NRC. OCRE's motion was premised on a
-1983 report by Sandia on the Grand Gulf Nuclear Station (" Grand Gulf") igniter system, the design of which was the basis for the PNPP igniter system design.
In response to OCRZ's motion, the NRC Staff furnished an affidavit from Dr. Berman. That affidavit indicated Dr.
Berman's acceptance of the Grand Gulf igniter system design (which had been approved by the NRC Staff subsequent to the Sandia report). The affidavit also indicated that Dr. Berman had no disagreements with the Staff's position on the adequacy of the PNPP igniter system. The Board thus saw no basis to compel Dr. Berman to testify about the PNPP igniter system, and denied OCRE's motion.19/
In response to a motion for continuance filed by OCRE, the Board delayed the start of the evidentiary hearings for Issue No. 8 until April 30, 1985.20/ Applicants and the NRC Staff submitted pre-filed written testimony on the hydrogen control issue. OCRE filed no testimony and presented no witnesses at the hearing. OCRE cross-examined both Applicants' and Staff's witnesses.
19/ Memorandum and Order (Motions for Continuance of Hearing and to compol Appearance of NRC Witness) (March 29, 1985).
20/ Id.
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. -o B. The Hydrogen Rule The Board has carefully considered the Commission's new hydrogen control requirements, contained in 10 C.F.R. 55 50.44(c)(3)(iv)-(vii), as well as the Commission's explanation of the rule, contained in the " Supplementary Information" sec-tion which accompanied the rule, 50 Fed. Reg. 3498-3505 (1985).
This partial initial decision is the first licensing board de-cision arising under the new regulation.
For the PNPP Mark III containment design, the new rule in 10 C.F.R. 5 50.44(c)(3)(iv)(A) requires a hydrogen control sys-tem justified by a suitable program of experiment and analysis.
The system must be capable of handling, without loss of con-tainment structural integrity, an amount of hydrogen equivalent to that generated from a 75% metal-water reaction. The hydro-gen rule, as distinguished from the Commission's separate rulemaking on severe accidents (see 45 Fed. Reg. 65474 (1980)),
deals only with recoverable degraded core scenarios. 10 C.F.R. 5 50.44(c)(3)(vi)(B)(2); 50 Fed. Reg. at 3499 (1985); see Tr.
3657-58 (Richardson), 3696-97, 3731 (Pratt), 3750 (Notafrancesco). The rule does not specify the type of hydro-gen control system that must be used. The hydrogen control system must be installed and operational prior to operating in excess of 5 percent power. See 10 C.F.R. 5 50.44(c)(3)(vii)(B); 50 Fed. Reg. at 3500 (1985).
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, o Under 10 C.F.R. $ 50.44(c)(3)(vii)(B), Applicants are re-quired to provide a " preliminary analysis" of their hydrogen control system to support an NRC Staff finding that the plant is safe to operate at full power (in excess of 5% power). Sec-tion 5 50.44(c)(3)(vii)(B) provides that " completed final anal-yses are not necessary" for such a Staff determination, pro-vided a preliminary analysis has been submitted and found acceptable by the NRC Staff. See 50 Fed. Reg. 3500 (1985).
Under 10 C.F.R. $ 50.44(c)(3)(vii)(A), Applicants are required to submit to the Staff by June 25, 1985, a proposed schedule for completing the final analysis. Under 10 C.F.R.
$ 50.44(c)(3)(vii)(D), a final schedule for meeting the requirements of the rule will be established by the NRC Staff within 90 days of receipt of Applicants' proposed schedule.
The hydrogen rule requires Applicants to perform a number of analyses to demonstrate the ability of the hydrogen control system to deal with accident scenarios producing large amounts of hydrogen. For example, under paragraph (c)(3)(iv)(B),
Applicants must demonstrate, using analytical techniques ac-capted by the NRC Staff, that PNPP's containment structural in-tegrity would be maintained following the generation of large quantities of hydrogen. Paragraph (c)(3)(v)(A) requires Appli-cants to demonstrate the survivability of equipment in the con-tainment necessary to establish and maintain safe shutdown and to maintain containment integrity. See 50 Fed. Reg. at 3501 l
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7 (1985). . Paragraph (c)(3)(vi)(B) requires Applicants to evalu-ate'"the consequences of large amounts of hydrogen," including
" consideration of hydrogen control measures as appropriate."
Applicants' evaluation must "[ijnclude the period of recovery from the degraded condition." The evaluation must use "acci-dent scenarios that are accepted by the NRC Staff," which are
" accompanied by sufficient supporting justification to show that they' describe the behavior of the reactor system during and following an accident resulting in a degraded core." Id.
Applicants are required by the same paragraph to "[s]upport the design of the hydrogen control system selected . . . ." Id.
The hydrogen rule, although not a model of clarity, strongly suggests that the Commission did not intend'that a preliminary analysis under paragraph (c)(3)(vii)(B) address the substantive requirements of paragraphs (c)(3)(iv)(B), (v) and (vi). The Board's interpretation is based on the fact that the final analysis schedule requirement of paragraph (c)(3)(vii)(A) expressly references paragraphs (c)(3)(iv), (v) and (vi),
whereas the preliminary analysis requirement in paragraph (c)(3)(vii)(B) only references paragraph (c)(3)(iv)(A), the general requirement for a hydrogen control system. The legal doctrine "expressio unius est exclusio alterius" (expression of one thing is the exclusion of another) suggests that the Com-mission excluded the detailed requirements of paragraphs (c)(3)(iv)(B), (v) and (vi) from the preliminary analysis N
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, e requirements'of the rule. At a minimum, it is clear that the Commission did not intend to require Applicants to demonstrate
' final compliance with all aspects of the hydrogen rule in order to receive a full power operating license. To read the rule otherwise would vitiate the rule's distinction between " prelim-inary" and " final" analyses.
It is also apparent that the Commission intended the Board to place weight on the NRC Staff's judgment of what constitutes a satisfactory preliminary analysis. Paragraph (c)(3)(vii)(B) of the rule states:
4 Completed final analyses are not necessary for a staff determination that a plant is safe to operate at full power provided that prior to such operation an applicant has provided a preliminary analysis which the staff has determined provides a satisfacto-ry basis for a decision to support interim operation at full power until the final analysis has been completed.
Id. (emphasis added). Similar discretion is explicitly ac-corded the Staff in other parts of the rule, e .' q . , paragraph (c)(3)(iv)(B) (" Containment structural integrity must be demon-strated by use of an analytical technique that is accepted by the NRC Staff"), and paragraph (c)(3)(vi)(B)(3) ("Use accident scenarios that are accepted by the NRC staff"). While these statements do not remove the preliminary analysis issue from the Board's jurisdiction, the Board concludes that the Commis-sion intended the Board to place at least some reliance on the Staff's judgments in these areas.
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The Board has concluded that, in the circumstances of this c a s e', there is no need for the Board to delineate the precise line between " preliminary" and " final" analyses under the new rule, or to-determine whether compliance with the rule's pre-liminary analysis provision, paragraph (c)(3(vii)(B), requires at'least a preliminary showing under paragraphs (c)(3)(iv)(B),
-(v) and (vi) of the rule. Applicants' preliminary analysis, as described in the Preliminary Evaluation Report and in Appli-cants' testimony, and as approved by the Staff and discussed in Staff's testimony, does address in detail the substantive pro-visions of the hydrogen rule, including the requirements in paragraphs (c)(3)(iv)(B), (v) and (vi) of the rule. For the reasono set forth in this partial initial decision, the Board has concluded that Applicants' preliminary evaluation, as re-viewed and accepted by the Staff, provides ample basis to sup-port the Staff's determination that PNPP is safe to operate at full power during the interim period prior to the completion of the final analysis.
The Board is satisfied based on the evidence that the PNPP hydrogen control system, using distributed igniters, is a safe and viable hydrogen control concept, based on well understood combustion phenomena and-extensive industry and NRC-supported research and analysis. In addition, the Board has concluded that conservative plant-specific analyses, and comparisons to Grand Gulf Nuclear Station's similar design and analyses, E
i provide sufficient preliminary information to conclude that (1) the PNPP containment ultimate capacity is far greater than the pressures resulting from hydrogen combustion following a de-graded core accident; thus, the structural integrity of the containment is' maintained during and following hydrogen burn-ing; and (2) the temperature for which the equipment is quali-fied to perform its function is greater than the expected equipment thermal response to the environment defined in the preliminary analysis of hydrogen combustion following a de-graded core accident; thus, necessary systems and components will be capable of performing their functions during and after hydrogen burning.
Therefore, in the unlikely event that the hydrogen control system would be called upon, Applicants' preliminary evaluation of the-system demonstrates that it would safely accommodate large amounts of hydrogen without a rupture of the containment and release of radioactivity to the environment. For these reasons, the Board has concluded that OCRE'S contention is without basis.
C. Witnesses Applicants' witnesses at the evidentiary hearing on Issue No. 8 were Eileen M. Buzzelli, John D. Richardson, Kevin W.
Holtzclaw, Roger W. Alley, Bernard Lewis, Bela Karlovitz, G.
Martin Fuls, and James H. Wilcox. Finding 3. As indicated I
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below, the Board found Applicants' witnesses to be well-qualified to address the~various technical areas encom-passed by the' hydrogen control issue.
Dr. Bernard Lewis and Bela Karlovitz of Combustion and Ex-plosives Research, Inc. (Combex) are distinguished experts in the field of combustion phenomena. Dr. Lewis has conducted re-search on hydrogen combustion for over 50 years and has pro-duced over 250 publications on combustion, flames and explo-sions, including basic textbooks used in many countries. Dr.
Lewis and Mr. Karlovitz have studied and published intensively in this area since performing research together at the U.S. Bu-reau of Mines. Findings 4, 5.
Dr. Lewis and Mr. Karlovitz impressed the Board during the hearings with their knowledge and understanding of the combus-tion phenomena that would occur following a degraded core event at PNPP, including the mitigative effect of the PNPP distri-buted igniter system. The Board notes that Dr. Lewis and Mr.
Karlovitz previously_ testified in the McGuire operating license proceeding about the distributed igniter system planned for that plant. The Appeal Board has recognized the special compe-tence of Dr. Lewis'and Mr. Karlovitz to address the hydrogen control issue. See Duke Power Company (William B. McGuire Nuclear Station, Units 1 and 2), ALAB-669, 15 N.R.C. 453, 471 &
nn.37 & 39 (1982).
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For these reasons, the Board has placed significant weight on the testimony of Dr. Lewis and Mr. Karlovitz in this pro-ceeding.
The Board found Dr. G. Martin Fuls to be similarly well-qualified to testify in his area of expertise. Dr. Fuls developed the CLASIX computer programs which have been used at PNPP.and other nuclear plants to address hydrogen burning issues in response to the TMI-2 accident. He has long and im-pressive credentials, including over 20 years experience in the development of analytical techniques for nuclear plant systems.
Finding 6. With Dr. Lewis and Mr. Karlovitz, Dr. Fuls was one of the originators of the distributed igniter concept. Tr.
3637-39 (Fuls, Karlovitz, Richardson).
John D. Richardson has worked closely with the Hydrogen Control Owners Group ("HCOG"), which was formed in 1981 to address generic hydrogen control NRC licensing issues common to BWR 6/ Mark III nuclear plants such as PNPP. Mr. Richardson chaired HCOG for several-years. Before joining Enercon Ser-vices, Inc., his present employer, Mr. Richardson worked for Mississippi Power & Light and was responsible for the licensing and technical issues relating to the Grand Gulf Nuc lear Station hydrogen ignition system. Mr. Richardson assisted Applicants in the preparation of the PNPP Preliminary Evaluation Report, "The Cleveland Electric Illuminating Company Preliminary Evalu-ation of the Perry Nucler Power Plant Hydrogen Control System"
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(" Preliminary Evaluation Report") (App. Ex. 8-1). Finding 7.
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I Eileen M. Buzzelli is Applicants' senior licensing engi-neer.in' charge of licensing issues relating to the PNPP distri-buted igniter system. Ms. Buzzelli was responsible for preparing the Preliminary Evaluation Report. Finding 8. Dur-ing the hearing, both Ms. Buzzelli and Mr. Richardson demon-strated a detailed understanding of the PNPP containment design
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and of-how the containment would be affected by a hydrogen gen-eration event. Ms. Buzzelli and Mr. Richardson also provided the Board with the necessary overview of various elements of the hydrogen control issue, which is a hybrid issue spanning a number of different technical disciplines, including structural analysis, combustion, thermodynamics, hydraulics, computer mod-elling, and other areas.
Mr. Holtzclaw of General Electric ("GE") has spent nearly 17 years in the nuclear power industry. He has been responsi-ble for GE programs related to NRC degraded core rulemaking, and has been the GE Program Manager of the GE Severe Accident Program. Mr. Holtzclaw gave authoritative testimony explaining various design'and operational features of the GE-designed PNPP BWR 6/ Mark III containment. He also testified about the probabilities of hydrogen-generation events, and explained the basis for the selection of the scenarios used in Applicants' preliminary evaluation of the PNPP igniters. Finding 9.
The Board also found Roger W.~ Alley to be a highly compe-tent witness. Mr. Alley has over 16 years experience in L
structural design of nuclear power plants and is the Project Structural Engineer for PNPP. Mr. Alley is employed by Gilbert / Commonwealth, Inc. ("G/C"). He was responsible for the analysis of the ultimate structural capacity of the PNPP con-tainment, as contained in " Cleveland Electric Illuminating Com-
.pany Perry Nuclear Power Plant Units 1 and 2 Ultimate Structur-al Capacity of Mark III Containments," transmitted to the NRC oon February 11, 1985 (" Ultimate Capacity Report") (App. Ex.
8-4). Finding 10.
OCRE cross-examined Mr. Alley at length concerning his analysis of the PNPP containment structural integrity. The Board has carefully considered Mr. Alley's testimony and his report on the PNPP ultimate capacity. The Board has concluded that Mr. Alley's analytical techniques and conclusions are appropriate and meet the requirements of the hydrogen rule.
Applicants' rebuttal witness, James H. Wilcox, is a welding engineer with nine years experience in the field. Mr.
Wilcox is Supervisor of the Test Support Group at PNPP. He gave persuasive testimony concerning welding issues raised by
.. OCRE , including the applicability of certain American Welding l
Society standards cited by OCRE. Finding 11.
The NRC Staff's witnesses at the evidentiary hearing were Allen Notafrancesco, Li Yang, Hukam C. Garg and Dr. William Trevor Pratt. Finding 12.
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OCRE questioned Mr. Notafrancesco's qualifications and moved to strike his testimony. Tr. 3671-74 (OCRE). The Board denied the motion. Although Mr. Notafranceso does not have the long experience of witnesses such as Dr. Lewis and Mr.
Karlovitz, he has been associated with hydrogen research for over three years, and is the individual at the NRC with respon-sibilityffor reviewing.the PNPP igniter system. Tr. 3674-75.
Mr. Notafrancesco's subsequent responses to questions during the hearing confirmed to the Board that Mr. Notafrancesco.is knowledgeable and competent to address the adequacy of the PNPP igniter system. Finding 13.
Dr. William ~Trevor Pratt has impressive credentials and publications in the area of core melt phenomena, including hy-drogen production and containment response. Finding 16. The
. Board found Dr. Pratt's testimony regarding the MARCH and HECTR codes, core melt phenomena, and related areas, to be of great assistance to the Board's understanding of.the hydrogen issue.
The Staff's-other two witnesses were Li Yang, who addressed Applicants' containment structural capacity analysis, and Hukam C.'Garg, who testified about equipment survivability-in containment following the production and burning of hydro-gen. - Findings 14, 15. The Board has found that the reviews-and testimony of the-Staff in these two areas address in;an appropriate manner Applicants' compliance with the hydrogen rule.
z D. . Hydrogen Control System The hydrogen control system at PNPP consists of a combus-tible gas control system ("CGCS") and a distributed igniter system. Findings 17, 20. The CGCG, which includes the hydro-gen recombiners.and several other subsystems, is designed to meet the original provisions of 10 C.F.R. $ 50.44 (unchanged by the hydrogen rule) for combustible gas control following a de-sign basis loss-of-coolant accident ("LOCA"). Findings 18, 19.
The distributed igniter system is in addition to the CGCS and is.used to control amounts of hydrogen beyond those covered by design basis accident requirements in the original hydrogen rule. Finding 20.
During the evidentiary hearing OCRE did not question the adequacy of the CGCS, but instead, focused its cross-examination on the distributed igniter system. The Board has concluded that the CGCS is adequate to handle the amounts of hydrogen which it was designed to control.
Applicants presented uncontradicted evidence at the hear-ing that PNPP's BWR 6/ Mark III design has a number of inherent safety systems -- in addition to the igniter system -- that re-duce the likelihood and effects of degraded core events, such as occurred in the TMI-2 PWR containment. Findings 21-24.
Applicants pointed to authoritative studies indicating that the risk of generation of large amounts of hydrogen and consequent offsite releases is extremely remote. Findings 25-28.
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OCRE argued that this~ evidence is not relevant to OCRE's contention. See Tr. 3224-32. The Board disagrees. OCRE's contention raises the spectre of a containment rupture and re-lease of substan'tial radioactivity to the public. The evidence demonstrated that even without a distributed igniter system this is unlikely to occur. Applicants-must of course still comply with.the new hydrogen ~ rule, and provide a hydrogen con-trol system capable of handling large amounts of hydrogen. As indicated below, the Board has concluded, wholly apart from the evidence on accident probabilities, that PNPP is safe to oper-ate at full power until Applicants' final-analysis has been completed, and that Applicants meet.the rule's requirement for-a preliminary analysis. Nonetheless, the evidence on accident probabilities referred to above adds to the Board's assurance that PNPP will be safe to operate without releases of substan-tial radioactivity to the public as claimed by OCRE.
E. Igniter System Selection and Preliminary Evaluation Applicants indicated to the NRC Staff in March 1982 that
.they.would be installing a distributed igniter system to im-prove PNPP's hydrogen control capability in response to the TMI
- accident. Applicants informed the Staff that the system con-
. cept would be the same concept as that being implemented at the time by all ice-condenser and Mark III plants. The igniter concept had been thoroughly analyzed at the other plants at the time Applicants selected it for PNPP. Finding 29.
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After selecting the igniter system, Applicants joined in a research and analysis effort being conducted by the Hydrogen Control Owners Group to address the use of igniters in Mark III containments such as PNPP. J.pplicants informed the Staff of this in 1982, and stated that they would be submitting plant-specific information in support of PNPP's operating license ap-plication. Finding 30.
In July 1984, Applicants. submitted to the NRC Staff a pro-gram plan describing the generic and plant-specific activities undertaken to resolve the hydrogen control issue for PNPP.
Finding. 31. In December 1984, HCOG submitted for the Staff's approval a plan for research'and analysis to be used for the final, long-term analysis of the distributed igniter system.
Finding 34.
Soon after the final hydrogen rule was promulgated, Appli-cants submitted for Staff approval a proposed scope of informa-tion to be submitted for a preliminary analysis under the rule.
Finding 32. The Staff approved the scope of information.as acceptable to support full power operation pending completion of the final analysis. Finding 35. It was agreed that Appli-cants would submit a preliminary evaluation which included a description of the igniter system, an analysis of PNPP contain-f ment ultimate capacity, a containment response analysis, and a comparison of the significant design features of PNPP to those
- of Grand Gulf. Finding 35. The Staff had previously I
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I-
determined that Grand Gulf's igniter system was acceptable to i
support safe interim operation at full power until completion of the final analysis, and had issued a full power license for Grand Gulf in mid-1984. Finding 33.
At the time Applicants submitted to the Staff their pro-posed scope of information for the preliminary evaluation, Applicants indicated that they would' endorse the HCOG long-term program when it was approved by the Staff and would then apply program results to the PNPP final analysis as necessary. Find-ing 34.
In March 1985, Applicants submitted to the Staff their preliminary evaluation of the PNPP distributed igniter system (App. Ex. 8-1), covering the areas previously outlined to the Staff. Finding 37. Applicants selected distributed igniters as.the most viable concept to control large amounts of hydro-gen. The system is designed to handle, without loss of con-tainment integrity, an amount of hydrogen equivalent to that generated from a metal-water reaction involving up to 75% of fuel cladding surrounding the active fuel region. Finding 38.
This is accomplished by burning hydrogen at low concentrations, thereby maintaining the concentration of hydrogen below levels which could potentially threaten containment integrity. Id.
In selecting the igniter system, Applicants considered a number of technical criteria, such as mitigation effectiveness, consequences of intended or inadvertent operation, reliability, 4
testability, and availability of design and equipment. Finding
- 39. There'is an industry consensus that igniters constitute an adequate system for controlling large amounts of hydrogen. Id.
For_these reasons, the Board has concluded that Applicants applied appropriate criteria in selecting a distributed igniter system for PNPP, and that the igniter system concept is indeed the most viable and logical-system available to meet the requirements of the new hydrogen rule.
F. Igniter System Design and Operation The distributed igniter system uses glow plug igniters such as those commonly used in diesel engines. Finding 40.
The igniters have been extensively tested and have demonstrated reliable ignition. Id.
The igniters are spaced in 102 locations throughout the drywell,.wetwell and upper containment regions of the plant.
i Findings 41, 44. Spacing criteria were selected to assure re-dundancy and complete coverage of all containment areas. Find -
l l ing 44. The igniters are maintained at a high surface tempera-I ture to assure reliable ignition. Finding 41.
The igniter assemblies are safety-grade, and are powered
-from safety-grade sources. Findings 42, 43. The igniters are distributed between two separate electrical power divisions, f
and can be powered from the diesel generators. Id.
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Dr. Lewis and Mr. Karlovitz have reviewed the placement of the igniters and believe they will be able safely and effec-tively to burn large amounts of hydrogen. Finding 46. They have concluded that the placement of igniters near ceilings or walls will not_ inhibit burning, and that hydrogen that accumu-lates can be expected to burn at low (8%) concentrations.
Findings 47, 48, 100. Dr. Lewis and Mr. Karlovitz have also determined that Applicants' distributed igniter system will prevent. local detonations from occurring, and that there is no danger of a transition from deflagration burning to detonation.
Findings 49-51. The NRC Staff has reviewed the system design and found it acceptable. Finding 53.
The Board has concluded, based on the views of the Staff",
the views of preeminent hydrogen combustion experts such as Dr.
Lewis and Mr. Karlovitz, and the absence of contrary evidence, that the igniters will burn large quantities of hydrogen reliably and at low concentrations,. and will preclude detonation from occurring inside the containment.
The Board has also concluded that the igniter system can be reliably operated, Findings 54, 55, and that it will be sub-jected to sufficient preoperational and functional testing.
Finding 52.
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G. Ultimate Structural Capacity of Containment
, Applicants' preliminary evaluation establishes the ulti-mate structural capacity of the PNPP containment, and evaluates the PNPP drywell pressure capability, to confirm that the con-tainment integrity would be maintained in the event large quan-tities of hydrogen are generated and burned. Findings 56-58.
Applicants have performed a number of analyses of the PNPP containment since 1981 to satisfy NRC requests and the require-ments in the hydrogen rule. Findings 59-61. G/C analyses using mean and lower-bound values and ASME Service Level C and D limits are included in the G/C Ultimate Capacity Report (App.
Ex. 8-4). Finding 61.
The analytical techniques and results in G/C's ultimate capacity analysis have been reviewed and approved by the NRC Staff. Finding 62. These analytical techniques and results adequately describe the containment response to the structural loads involved. Finding 63. In G/C's ASME Service Level C analysis, G/C conservatively used minimum specified materials in all but one case. In that case, G/C used actual material properties with suitable margins. Finding 64.
-Applicants' analysis demonstrates, and the Board has con-cluded, that the PNPP containment meets the Service Level C requirements of the ASME Code. Finding 65. Further, Service Level C limits are conservative, and there is additional pres-sure capacity in the PNPP containment beyond Service Level C limits. Finding 66.
i
c.- .
'G/C's Ultimate Capacity Report shows that the limiting pressure capacity of the PNPP containment using Service Level C limits.is conservatively calculated to be 50 psig. Findings 67, 68. This is over 3 times greater than the design level, and well exceeds the peak pressure of 21 psig after hydrogen 4
combustion calculated in the PNPP preliminary evaluation.
Findings 67,-69.
i Applicants have also evaluated the negative pressure capa-bility of the containment, and the positive and negative pres-sure capability of the drywell. These evaluations show that
.the expected pressures from hydrogen. combustion can be ade-
-quately handled. Findings 70, 71.
The Board has concluded, based on its consideration of the above, and using the ASME Service Level C limits and other criteria referenced in the hydrogen rule, that the PNPP con-tainment structural integrity would be maintained in the un-5 likely event of a degraded core accident generating large amounts of hydrogen.
During the hearing, OCRE raised questions about certain inaccessible welds containing potentially rejectable indica-tions under the ASME Code. See Finding 74. These welds have been thoroughly analyzed by Applicants, and the Board has con-
.cluded that the welds would not impair the integrity of the PNPP containment in the event of a recoverable degraded core event. Findings 76-78. OCRE raised other welding concerns I'
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which the Board has concluded are unfounded and do not call into question the structural analysis in Applicants' Prelimi-nary Evaluation. Findings 74, 75.
OCRE also cited tests conducted on models of other con-tainments in which the models were tested to failure. Finding
- 79. The Board has concluded that these tests have no relevance to the preliminary evaluation and are outside the scope of the hydrogen rule. Id.
OCRE also asked a number of questions during the hearing about the potential for leakage through drywell or containment wall penetrations in the event of significant hydrogen genera-tion and burning. See Findings 80-84. The Board has examined each of these issues and has concluded that OCRE's concerns are-unfounded. Id.
H. Containment Response and Equipment Survivability Applicants' preliminary evaluation analyzes selected acci-dent scenarios using the MARCH and CLASIX-3 computer programs i
previously used to analyze Grand Gulf's igniter system. Find-l ing 85. Applicants analyzed combustion from postulated hydro-gen releases, and the resulting thermal and pressure environ-l ments, and effects on equipment survivability. Id.
l Applicants analyzed two scenarios: (1) a drywell break, loss-of-coolant accident with temporary failure of emergency core cooling system injection, and (2) a transient with a l
l E
stuck-open relief valve accompanied by a temporary failure of emergency core cooling system injection. Finding 86. The two scenarios are based on' accident event initiators previously an-alyzed by the industry and NRC. Findings 86, 87. The NRC has accepted the scenarios for use in the PNPP preliminary'evalua-tion. Finding 88. The Board has concluded that the scenarios represent the bahavio of the reactor system for a postulated accident resulting in a degraded core, see 10 C.F.R. 5 50.44(c)(3)(vi)(B)(3), and are appropriate scenarios to use for a preliminary analysis.
The MARCH Code was originally developed by Battelle Columbus Laboratory for the NRC. Finding 89. MARCH provided a method to obtain hydrogen release rates for the preliminary evaluation. Id. -A'ssumptions made in MARCH about hydrogen and steam. releases are consistent with the hydrogen rule and are
_ conservative. Findings 90-92. .The Staff has reviewed and ac-cepted Applicants' use of the MARCH Code results, concluding that the MARCH results are conservative and. postulate the most severe challenge to the-hydrogen control system and maximize the rate of heat added to containment. Finding 92. The Staff
'has identified some shortcomings in the MARCH Code, but they do not affect the acceptability of the Code for use in the PNPP preliminary evaluation. Findings 93, 94.
Based on the above, the Board has concluded that Appli-cants use of the MARCH Code results in the preliminary 4
b
1 evaluation is acceptable and is consistent with criteria in the hydrogen rule.
The PNPP preliminary evaluation used the CLASIX-3 computer program.to analyze the PNPP containment response to hydrogen combustion. Finding 95. CLASIX-3 was used in the Grand Gulf
-containment response analysis, and was adopted from the CLASIX code.used to analyze igniter systems in ice condenser. plants.
Id. There are three compartments in the PNPP CLASIX-3 model:
the drywell, wetwell and containment. The code models MARK III containment plant features while tracking the distribution of oxygen, hydrogen and steam. Findings 95-99.
The CLASIX-3 model uses accepted engineering input assump-tions, and makes conservative assumptions about the behavior of the reactor system during and following the two hydrogen re-lease scenarios analyzed in the PNPP preliminary evaluation.
Id-CLASIX-3 models deflagration burning, which is the propa-gation of a slow, discrete-type flame through a flammable mix-ture; CLASIX-3 does not model diffusion (continuous) flames, which will be addressed in Applicants' final analysis. Finding 100. Dr. Lewis and Mr. Karlovitz have reviewed and confirmed the flame speed and ignition assumptions used in CLASIX-3. Id.
They have also confirmed that the high degree of turbulence following a hydrogen generation event will promote rapid and complete mixing of hydrogen, which will result in uniform L-
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l hydrogen concentrations in any given compartment. Id. The l l
Board has concluded, in the absence of evidence to the con-trary, that the ignition; flame speed, and mixing assumptions in the PNPP CLASIX-3 analysis are appropriate.
The CLASIX-3 results for PNPP show that the most severe temperatures and pressures from hydrogen burning occur in the wetwell. Finding 101. For a few burns, the temperature in the wetwell reaches approximately 1760* F and pressures reach ap-proximately 21 psig for a short period of time. Id. The PNPP-peak pressures and temperatures calculated by CLASIX-3 are com-parable to those calculated for Grand Gulf. The exception is PNPP's peak wetwell temperature, which is higher than Grand Gulf's. This higher peak wetwell temperature is not expected to have a significant effect on the overall equipment tempera-ture rate. Finding 102. This is because the burns are few in rundoer and of short duration when compared to the time required for heat transfer to the equipment and the resulting tempera-ture increase. Finding 112. Burns at PNPP are fewer in number and of shorter duration than those of Grand Gulf, and are ex-pected to result in lower equipment temperatures. Id.
The Staff has reviewed and approved Applicants' use of CLASIX-3 in the PNPP preliminary evaluation. Finding 103. The Staff found that the analysis is conservative in that it over-
. predicts the number of burns that will occur in the wetwell.
Id. The Staff also concluded that the Grand Gulf containment analysis should bound the analysis at PNPP. Id.
L . __._ __
OCRE asked a number of questions about early Staff re-quests for information about CLASIX-3 in 1982 and 1984 (before the final hydrogen rule was promulgated). See Findings 104, 105. The Board has concluded that the Staff's earlier re-quests, as explained during the evidentiary hearings, exceed the scope of the preliminary analysis requirements in the hy-drogen rule and do not call into question the adequacy of Applicants' preliminary evaluation. Findings 104, 105.
OCRE was also concerned that Applicants did not use the HECTR Code to model the PNPP containment analysis, citing Sandia's earlier evaluation of Grand Gulf using HECTR. See Finding 106. For the reasons noted during the hearing, the Board has concluded that the HECTR analysis performed by Sandia for Grand Gulf is now out-of-date and does not in any way call-into question Applicants' PNPP analysis. Id.
OCRE also raised concerns about possible drywell bypass
' leakage in the drywell break scenario. See Finding 107. The Board has concluded, based on the uncontradicted evidence presented during the hearing, that the effect of drywell bypass will be minimal and is of no concern to the operation of the PNPP distributed igniter system. Id.
The Board has also considered OCRE's claims that suppres-sion pool cooling might be diminished in the accident and re-covery scenarios analyzed by Applicants. See Finding 108.
There was no credible evidence that this would occur, and the 4
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Board has concluded that OCRE's concern is without basis. Id.
OCRE also raised a concern about the possibility of reduced suppression pool mixing due to the operation of the sprays, citing an analysis by John Humphrey, a former GE engineer. See Finding 109. Mr. Humphrey's concerns related to design basis accidents and not to hydrogen generation events. The Board has concluded that Mr. Humphrey's concerns do not call into ques-tion the effectiveness of the distributed igniter system to re-spond to a recoverable degraded core event producing hydrogen.
Id.
Applicants have performed a preliminary assessment of the ability of PNPP equipment to survive deflagration burning based on comparison to NRC-approved survivability analyses previously performed for Grand Gulf. Finding 110. Applicants have developed a preliminary list of equipment required to survive a hydrogen event, using criteria that were used at Grand Gulf and that are consistent with the hydrogen rule. Id.
The preliminary evaluation compared the temperature pro-files and equipment thermal responses of similar equipment at PNPP and Grand Gulf. The comparison demonstrated that the thermal response of the PNPP equipment remained below its envi-ronmental qualification temperature, and was lower than the thermal response of similar equipment at Grand Gulf. Finding 111. Applicants also performed evaluations demonstrating the ability of PNPP equipment to survive the pressures predicted by CLASIX-3. Finding 113.
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Based on the above, the Board has concluded that essential equipment at PNPP would survive expected temperatures and pres-sures as-defined by CLASIX-3, and would be available to estab-lish and maintain safe shutdown and to maintain containment in-tegrity in the event of a recoverable degraded core event producing large quantities of hydrogen. Although Applicants have not completed equipment survivability tests for diffusion flames, diffusion flame burning is not expected to be a threat to containment. Finding 114. The use of more realistic hydro-gen release rates in the final analysis.is expected to result in lower temperatures and a less severe thermal environment than those evaluated by Applicants to date. Id. Shielding of components in the area of the diffusion flames also minimizes the potential effect on equipment. Id. For these reasons, the Board concludes that PNPP is safe to operate at full power pending completion of Applicants' final analysis.
At the hearing, OCRE questioned whether electrical cable within the containment would survive the effects of hydrogen burning, citing a paper by Dr. Pratt and others and the results.
of tests performed by Fenwal and at the Nevada Test Site.
Finding 115. The Fenwal tests do not apply to PNPP's equip-ment. Id. The Nevada Test Site results support Applicants' conclusion that electrical cable would survive the thermal en-vironment of hydrogen burning in a recoverable degraded core sequence. Id. For these and other reasons set forth in L - _ _ _ __ _
Finding 115, the Board has concluded that electrical cable would survive the effects of hydrogen burning. Similarly, the Board has considered and rejected as without basis concerns ex-pressed by OCRE about the survivability of polymeric seals used in PNPP lower personnel airlock and equipment hatches. Finding 116.
I. Summary and Conclusion As indicated in the foregoing sections, the Board has carefully considered the various elements of Applicants' pre-liminary evaluation of.the PNPP distributed igniter system,
' including Applicants' preliminary analysis of system design and operation, containment structural integrity, containment re- ,
sponse and equipment survivability. The Board has also consid-ered concerns expressed by OCRE about the adequacy of Appli-cants' analyses in these areas. For the reasons noted, the Board has concluded that Applicants have met the hydrogen rule by providing a preliminary analysis of the PNPP distributed ig-niter system which assures that PNPP can be safely operated at full power until Applicants' final analysis has been completed.
The Board has concluded that the igniter system could safely handle large amounts of hydrogen in the unlikely event of a re-coverable degraded core event at PNPP, without release of sub-stantial quantities of radioactivity to the environment.
FIND'INGS OF FACT I. CONTENTION'AND WITNESSES
- 1. Issue No. 8, as reworded by the Board, was admitted as a contention in this proceeding by Memorandum and Order (Mo-tions) (March 13, 1985). The Board explained the basis for ad-mitting the reworded contention in Memorandum and Order (Mo-tions on Hydrogen Control Contention) (March 14, 1985). OCRE
- . s..
is the lead intervenor on-the issue and-the only intervenor which participated in.its litigation. OCRE presented no wit-nesses.
- 2. Issue No. 8, as litigated,-reads as follows:
V The Perry hydrogen control systera is inadequate to assure that large amounts of hydrogen can'be safely accommcdated without a rupture of the containment and a release of substan-tial quantities of radioactivity to the environment. ,
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3 '. Applicants'-direct te dimony~atzthe' hearing was
~
presented by a panel of seven witnesses. " Applicants' Direct
' Testimony of Eileen M. Buzzelli, John D. Richardson, Kevin W.
Holtzclaw, Roger W. Alley, Bernard Lewis,'Bela Karlovitz and G.
Martin Fuls on the Preliminary Evaluation.of the Perry Nuclear Power Plant Hydrogen Control System (Issue >&B)," ff. Tr. 3241
" Applicants Testimony"). Applicants also presented oral rebut-tal testimony by James H. Wilcox to address specific welding
\
issues raised by OCRE duringfits cross-examination. Tr.
3751-59 (Wilcox).
- 4. Dr. Bernard Lewis and Bela Karlovitz of Combex have long_and distinguished backgrounds in the area of combustion,
-flame, and explosives research and application. Dr. Lewis holds degrees from Massachusetts Institute of Technology (B.S.,
Chemical Engineering), Harvard (Masters, Physical Chemistry),
and Cambridge University, England (Ph.D, Physical Chemistry and ScD Honorary Degree). Dr. Lewis was for many years the Chief of the Explosives and Physical Sciences Division of the U.S.
Bureau of Mines, where he was in charge of all combustion, flame, and explosives research. He holds many distinctions and titles, such as Founder and Honorary Chairman of the Combustion Institute, an international scientific society. Dr. Lewis has produced over 250 publications, chiefly on combustion, flames, and explosions, including basic combustion textbooks used in many countries. He has studied hydrogen combustion intensively since 1930. Applicants Testimony at 2, 5-6 (Lewis), attached qualification of Dr. Bernard Lewis; Tr. 3673 (Hiatt).
- 5. Bela Karlovitz spent many years conducting experimen-
-tal research on the applications of flames in the generation of electricity. He originated a process by which electrical power is produced directly from flames (magnetohydrodynamic power generation), and has originated and developed other U . _ - _ _ _ _ _ _ _ _ -_ _ _ .
=
applications based on fundamental turbulent flame theory. Mr.
Karlovitz worked with Dr. Lewis at the Bureau of Mines and was the Chief of the Flame Research Section from 1947-1953. He has authored numerous publications, particularly in the area of turbulent flames. Mr. Karlovitz and Dr. Lewis have been in-volved with the' development of the igniter system as a hydrogen control system-in nuclear power plants since the' issue first received consideration following the TMI-2 accident. Mr.
Karlovitz and Dr. Lewis testified about their evaluation and approval of the PNPP distributed igniter system. Applicants Testimony at 3, 6-8 (Karlovitz), attached _ qualifications of Bela Karlovitz; Tr. 3256-58 (Karlovitz).
- 6. Applicants also sponsored the direct testimony of Dr.
G. Martin Fuls. Dr. Fuls holds degrees from Carnegie Mellon University (B.S. and M.S., Mechanical Engineering), and from the University of Pittsburgh (Ph.D, Mechanical Engineering).
Dr. Fuls has over 20 years experience in the development of analytical techniques for analyses of nuclear plants. He worked extensively with computer. programs at Bettis Atomic Power Laboratory, and then with Offshore Power Systems (Westinghouse). After the TMI-2 accident, Dr. Fuls began
! developing analytical methods to evaluate the consequences of a hydrogen burn in a reactor plant containment. As an outgrowth of this work, Dr. Fuls developed the CLASIX-3 model used to
,, 4 l
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, e evaluate the'PNPP containment pressure and temperature re-sponses.to degraded core' events with hydrogen release and deflagration. Dr. Fuls testified about his-CLASIX-3 analysis at Perry, " Containment Pressure and Temperature Response to Hy-drogen Combustion for Cleveland Electric Illuminating Perry
. Nuclear Power Plant," dated October 7, 1982 (" OPS-38A92") (Ap-pendix.A to App. Ex. 8-1). Applicants Testimony at 3, 7-8 (Fuls), attached qualifications of Dr. G. Martin Fuls; Tr. 3255 (Fuls).
- 7. John D. Richardson, another member of Applicants' witness panel, is employed by Enercon' Services, Inc. Mr.
Richardson was previously employed as a nuclear engineer for Westinghouse Electric Corporation from 1974 to 1976, and then by Mississippi Power & Light Co. from 1976 to 1983. In the latter position Mr. Richardson.was responsible for licensing and technical issues relating to the Grand Gulf Nuclear Station Hydrogen Ignition System. He served for over two years as Chairman of the Mark III Hydrogen Control Owners Group
'("HCOG"). HCOG was formed in 1981 to address generic hydrogen control NRC licensing issues common to BWR 6/ Mark III nuclear plants. Since joining Enercon, Mr. Richardson has worked with Applicants on the preliminary evaluation of the PNPP igniter system. Mr. Richardson testified about the igniter system de-sign, and about analyses of the igniter system that have been L.
performed-for Mark'III containments under HCOG's direction.
Applicants Testimony at 2-4 (Richardson), attached qualifica -
tions of John D. Richardson; Tr. 3248-49 (Richardson).
- 8. Eileen M. Buzzelli is employed by CEI as a Senior Li- -
censing Engineer. Ms. Buzzelli has been responsible for all licensing issues relating to the PNPP Preliminary Evaluation Report (App. Ex. 8-1). Ms. Buzzelli's testimony provided information on Applicants' evaluation, selection, and imple-
' mentation of the igniter system for hydrogen control at Perry.
Applicants Testimony at 1, 3 (Buzzelli), and attached qualifi-cations of Eileen M. Buzzelli; Tr. 3244-45 (Buzzelli).
- 9. Kevin W. Holtzclaw also provided direct testimony on behalf of Applicants. Mr. Holtzclaw is a principal licensing engineer with General Electric Company, which has designed the BWR 6/ Mark III' plants including PNPP. Mr. Holtzclaw spent nearly 17 years as an engineer in the nuclear power industry, and has worked with GE since 1969. Since 1980, he has been ac-tive on behalf of GE in programs related to degraded core rulemaking. He has served as the GE Program Manager of the GE
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Severe Accident Program. Mr. Holtzclaw testified about the de-sign and operation of the BWR 6/ Mark III system, about post-TMI improvements in the design, and about the probabilities of var-ious accident scenarios potentially involving hydrogen produc-tion. Applicants Testimony at 2, 4-5 (Holtzclaw), attached qualifications of Kevin W. Holtzclaw; Tr. 3246-47 (Holtzclaw).
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- 10. Roger W. Alley also testified for Applicants. Mr.
Alley is employed by Gilbert / Commonwealth, Inc. ("G/C"), the
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architect-engineer for PNPP. He has over 16 years experience in structural design of nuclear power plants and is GK's Project Structural Engineer for PNPP. Mr. Alley has the over-all responsibility at G/C for the PNPP structural design and supervises approximately 25 G/C structural engineers in that task. He was involved in developing the original ASME contain-ment design sp'ecification for PNPP, and has been responsible for both finite element and stress analysis of PNPP's shell-type structures, including the containment, drywell and shield building. Mr. Alley was responsible for preparing the Ultimate Capacity Report (App. Ex. 8-4). Mr. Alley provided extensive testimony about his evaluation of the PNPP contain-ment structural capacity, and whether containment integrity would be maintained in the event hydrogen were released and burned. Applicants Testimony at 2, 5 (Alley), attached quali-fications of Roger W. Alley; Tr. 3253-54 (Alley).
- 11. Applicants presented James H. Wilcox as a rebuttal witness to address welding issues raised by OCRE. Mr. Wilcox is employed by Applicants as the Supervisor of the Test Support Group at PNPP. He holds a Bachelors of Science degree in welding engineering from the Ohio State University, with nine years experience as a welding engineer. Mr. Wilcox testified
{
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about various requirements contained in the American Welding Society, Welding Handbook, to which OCRE had referred during cross-examination. Mr. Wilcox also addressed other welding questions raised by OCRE, encompassing certain QA/QC reports, the impact of multiple weld repairs, the adequacy of Appli-cants' analysis of certain inaccessible containment weld de-facts, as addressed by OCRE Ex.-13 (the Aptech Report)', and other welding issues. .Tr. 3751-59.(Wilcox).
- 12. The NRC' Staff's. testimony at the evidentiary hearing was presented by a panel of four witnesses. " Testimony of Allen Notafranceso Regarding Issue #8 (Hydrogen Control)"
("Notafrancesco I"); " Testimony of Li Yang Regarding Issue #8
-(Hydrogen Control)" (" Yang"); " Testimony of Allen Notafrancesco on 'he t Hydrogen Control Issues Contained in the Licensing Board Contention #8" ("Notafrancesco II"); "Testimon) of Hukam C.
Garg Regarding Issue #8 (Hydrogen Control)" ("Garg"); and " Pro-fessional Qualifications of William Trevor Pratt;" ff. Tr.
3676.
- 13. Allen Notafrancesco is employed by the NRC staff as a Containment Systems Engineer in the Containment Systems Branch
("CSB") of the Division of Systems Integration. CSB is the lead NRC branch for review and' evaluation of hydrogen control systems. As part of-his responsibilities, Mr. Notafrancesco
~
has been associated with hydrogen combustion research and L - . - . __ _ . . _ . -_
hydrogen issues for over-three years. Mr. Notafrancesco is the lead NRC. staff reviewer.for the Perry hydrogen control system.
He holds a Bachelor of Science degree in Nuclear Engineering-and'a Master of Science degree in Mechancial Engineering from
, the Polytechnic Institute of New York. Mr. Notafrancesco testified about the NRC Staff's overall- review and approval of
<. 'the PNPP distributed ignition system. His testimony addressed the Staff's evaluation of the-igniter system design, the MARCH j and CLASIX-3 computer models used to analyze hydrogen produc-tion and containment response following a degraded core event, and.the-Staff's review of other aspects of Applicants' prelimi-nary evaluation of the PNPP hydrogen control system.
Notafrancesco I and II; Tr. 3666-68, 3672-75 (Notafrancesco).
- 14. Li Yang provided direct testimony for the NRC Staff f'
i discussing the Staff's review and acceptance of the structural
(
l capacity portion of Applicants' Preliminary Evaluation Report.
Mr. Yang is employed by the NRC as a structural engineer in the Structural and Geotechnical Engineering Branch, Division of En-gineering, Office of Nuclear Reactor Regulation. He holds Bachelors and Masters degrees in civil engineering from universities in Taiwan, China, and British Columbia, Canada.
Before' joining the NRC, Mr. Yang worked for over 15 years as a
. civil / structural engineer with architect / engineering firms in-volved in the design of industry buildings, fossil and nuclear
- power plant facilities. Yang.
[
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- 15. The NRC Staff also presented the direct. testimony of Hukam C. Garg. Mr. Garg testified about his review of Appli-cants' Preliminary Evaluation Report, insofar as it addressed the survivability _of equipment in the PNPP containment environ-ment following the production and burning of large amounts of hydrogen. Mr. Garg is employed by the NRC as an electrical en-gineer~in the Division of Engineering. His responsibilities include the review of licensee and license applicant environ-mental. qualification programs for safety-related equipment in nuclear plants. Prior to joining the NRC in 1981, Mr. Garg worked for architect / engineering firms from 1969 to 1980 on nuclear plant instrumentation and controls, including equipment qualification, and nuclear plant electrical system design.
Garg; Tr. 3668-69, 3721 (Garg).
- 16. The last member of the NRC Staff's witness panel was Dr. William Trevor Pratt. Dr. Pratt holds Bachelor of Science and Doctor of Philosophy degrees in Mechanical Engineering from the University of Strathclyde, Glasgow, Scotland. He is cur-rently employed by the Brookhaven National Laboratory as Group Leader (Principal Investigator) of the Accident Analysis Group, Division of Engineering and Risk Assessment. Among his other duties, Dr. Pratt provides technical assistance to the NRC
, Staff in assessing potential core meltdown accidents in nuclear reactors. Dr. Pratt has published extensively on the subject l
v
of core melt phenomena, including hydrogen production and con-tainment response issues. Professional Qualifications of William Trevor Pratt ff. Tr. 3676. Dr. Pratt did not sponsor any pre-filed direct testimony. He testified that he was fa-miliar with the contents of all the Staff's testimony, particu-larly the testimony in Notafrancesco II concerning hydrogen generation, and attested to its accuracy. Tr. 3670-71 (Pratt).
Dr. Pratt provided testimony regarding computer codes such as MARCH and HECTR used to analyze hydrogen production and burning following degraded core events. See, e.g., Tr. 3688-91, 3726 (Pratt). Dr. Pratt also testified about the differences be-tween degraded-core and severe accident events. See, e.g., Tr.
3698, 3708, 3731 (Pratt).
II. HYDROGEN CONTROL SYSTEM A. Combustible Gas Control' System
- 17. The PNPP hydrogen control system for controlling the concentration of hydrogen which may be generated following a design basis accident is the combustible gas control system.
This is accomplished by mixing volumes of high concentration with those of low concentrations and recombining the hydrogen and oxygen to control the long-term buildup of hydrogen in the containment. Applicants Testimony at 36 (Richardson).
l
- 18. The CGCS is designed to meet original provisions of 10 C.F.R. 5 50.44 (unchanged by the hydrogen rule) for combus-tible gas control following a design basis loss-of-coolant accident. The CGCS consists of the following subsystems: the hydrogen mixing or drywell purge subsystem, the hydrogen recombiner system, the hydrogen analysis subsystem and a backup containment purge subsystem. Id. at 36-37 (Richardson).
- 19. The PNPP hydrogen recombiner system is designed to prevent flammable hydrogen concentrations from forming in the containment or drywell following a LOCA, in*which hydrogen has been generated from the metal-water reaction within the reactor vessel, and from long-term sources of hydrogen such as ra-diolysis. The recombiner system consists of two 100% capacity
.recombiners located in the containment, and a control panel and power supply cabinet located outside containment. Id. at 35-36 (Richardson); Tr. 3251-52 (Richardson).
- 20. The distributed igniter system is in addition to the CGCS and is used to control amounts of hydrogen beyond those covered by design basis accident requirements in the original hydrogen rule. Following the generation of large amounts of hydrogen from a recoverable degraded core accident (equivalent to up to 75% metal-water reaction), the igniter system will be used for controlled combustion ~at low concentrations. The recombiners would be utilized for long-term control of the remaining hydrogen. Applicants Testimony at 35 (Richardson).
L
. .=
B. BWR System Features to Minimize Risk of Large-Hydrogen Releases
- 21. There are a number of inherent design features of BWR 6/ Mark III plants such as PNPP that reduce the likelihood and effects-of degraded core events, such as occurred _in the TMI-2 PWR containment. BWR 6/ Mark III features associated with the plant heat sink, the pressure relief capability, reactor water level measurement, noncondensible gas venting and natural cir-culation, depressurization capability, and core cooling mini-mize the risk of a TMI-2 type of accident from occurring and resulting in degradation of the reactor core. Id. at 10-13 (Holtzclaw).
- 22. The TMI-2 accident involved a metal-water reaction in the core resulting in the generation of amounts of hydrogen be-yond the NRC's design basis accident requirements. Following the TMI accident, the NRC required a number of design improve-ments to all light water reactors (NUREG-0737/TMI Action Plan).
These.TMI-related de~ sign improvements were for the_ purpose of further reducing the likelihood and effects of degraded ~ core accidents, in addition to the inherent design capability of the plants. Id. at'10 (Holtzclaw).
- 23. Several significant improvemente have been or will be implemented at PNPP. Examples of these improvements include:
improved containment instrumentation to monitor containment L
pressure and radiation level and suppression pool water level following an accident; improved post-accident sampling capabil-ity to obtain a post-accident " grab sample" of reactor water and containment air at an accessible location to facilitate as-sessments of core damage; improved emergency procedures guidelines to provide plant' operators with concise, system-oriented procedures to follow during an emergency; and control room improvements which provide man-machine interface in the control room and facilities for responding to an emergency.
Id. at 14 (Holtzclaw).
- 24. Two other post-TMI improvements specific to BWR de-sign are: the automatic restart of High-Pressure Core Spray System at low reactor level in the event the operator takes manual control of the system and subsequently fails to maintain water level; and adding automatic depressurization logic for non-break events (e.g., loss of feedwater) accompanied by fail-ure of all high-pressure cooling systems. Id. at 14 (Holtzclaw).
- 25. .There have been a number of risk and reliability analyses performed to confirm the capability of the BWR 6/ Mark III design to prevent or terminate potential accident sequences and minimize the risk of offsite releases of radioactivity.
GE's analysis concluded that accident sequences resulting in core damage occurred with a frequency of approximately
-6 5 x 10 per reactor year. Approximately one-third of
~
Lthese events did'not involve hydrogen-related failure modes.
Id. at 15 (Holtzclaw).
- 26. An independent review by the NRC Staff and its con-tractors resulted in a core damage frequency value of approxi-mately 2'x 10 per reactor year. Id. at 15 (Holtzclaw).
- 27. These values support the conclusion that core damage events which lead to significant quantities of hydrogen genera-
-tion are very low in probability. These values are well below comparable frequencies calculated in the Reactor Safety Study (WASH-1400) and are also well below the proposed NRC safety goal value for core damage frequency of 1 x 10 ~4 per reac-tor years. Id. at 15 (Holtzclaw).
- 28. In addition, analysis of the frequency of core damage and hydrogen generation have been made as part of the Industry Degraded Core Rulemaking ("IDCOR")-Program. One of the primary technical conclusions of the IDCOR Program was that the probabilities of severe nuclear accidents are extremely low.
IDCOR's quantification of the probability of core damage and hydrogen generation for a BWR 6/ Mark III plant ranged from
-6 ~
1.9 x 10 to 8.3 x 10 per reactor year. Id. at 15-16 (Holtzclaw).
L
C. Igniter System Selection and Preliminary Evaluation
- 29. In a letter dated- March 22, 1982, and in Amendment 8 to the PNPP Final Safety Analysis Report, Applicants responded to a request by the NRC Staff for a description of PNPP's pro-gram to improve its hydrogen control capability in response to the TMI accident. Applicants indicated.that a distributed ig-niter system had been selected for hydrogen mitigation at PNPP, and that the design would be based on igniter systems that had been developed at the Grand Gulf, Sequoyah, McGuire and D.C.
Cook nuclear stations. The distributed igniter system selected for PNPP was the same post-TMI hydrogen mitigation concept being implemented by all ice-condenser and Mark III containment plants. The igniter design concept had been thoroughly ana-lyzed and reviewed by the utilities at these plants and by the NRC at the time it was selected for PNPP. Id. at 16-17 (Buzzelli).
- 30. For BWR 6/ Mark III plants, research and analysis to support the use of an igniter system to accommodate large amounts of hydrogen was being performed through the Hydrogen Control Owners Group. Applicants indicated to the NRC Staff in 1982 that they were participating in the HCOG program to address the use of igniters in Mark III containments and would be submitting plant-specific information in support.of Appli-cants' operating license application. Applicants' commitments b _ _
\
.- 1.
were reflected in Section 6.2.7 of the NRC Staff's Safety Eval-uation Report (NUREG-0887),-dated May 1982. Id. at 17
- (Buzzelli).
~1. In July 1984, Applicants provided an update to the NRC Staff on Applicants' ongoing efforts to resolve the de-graded core hydrogen control issue. At that time, Applicants submitted a program plan describing the generic and plant-specific activities undertaken to resolve the hydrogen control issue for PNPP. Id. at 17 (Buzzelli).
- 32. _Following the publication'of the final hydrogen rule, Applicants sought clarification from the NRC Staff regarding the scope of information that was acceptable to the Staff for a preliminary analysis under the rule. In a letter to the NRC dated February 5, 1985 (App. Ex. 8-2), Applicants identified the information on hydrogen control which it planned to provide in a preliminary evaluation. The letter indicated that Appli-cants' preliminary evaluation would cover a description of the igniter system, an analysis of containment ultimate capacity, a containment response analysis, and a comparison of the signifi-cant design features of PNPP to those of Grand Gulf. Id. at 18-19 (Buzzelli); App. Ex. 8-2.
- 33. Prior to Applicants' February 5, 1985 letter, and as
- noted in the letter, the Staff had determined for a similar b
plant, i.e., Grand Gulf, that similar. systems were acceptable to support safe interim operation at full power until comple-tion of the final analysis. The NRC issued the Grand Gulf full power license in mid-1984. App. Ex. 8-2; 3742-44 (Notafrancesco).
- 34. On December 14, 1984, HCOG submitted for the Staff's review a program plan of research and analysis to be used to justify the distributed igniter systems on a final, long-term basis. At the time of Applicants' February 5, 1985 letter, the HCOG long term program was still under review by the Staff.
Applicants stated in the February 5 letter that they will com-plete the PNPP final analysis on a schedule consistent with HCOG's program. Applicants will endorse the NRC-approved HCOG plan, and as the results of the HCOG program become available, Applicants will address their applicability to PNPP as neces-sary. App. Ex. 8-2.
- 35. The NRC responded to Applicants' February 5, 1985 letter by letter dated February 20, 1985 (App. Ex. 8-3). The Staff stated that the scope of the preliminary information Applicants was providing in support of full operating license for PNPP was acceptable. App. Ex. 8-3; Applicants Testimony at
- 19. The NRC Staff found that the scope of the PNPP preliminary analysis was consistent with the scope of similar information provided for the Grand Gulf igniter system, which the Staff had L. .
reviewed and accepted. Tr. 3742-43 (Notafrancesco); Applicants Testimony at 30 (Buzzelli).
'36. Applicants submitted their Preliminary Evaluation Re-port (App. Ex. 8-1) on March 1, 1985, and provided supplemental information on March 21, 1905. The Preliminary Evaluation Re-port covers four major areas of the PNPP design and analysis.
These include a hydrogen control system description, a descrip-tion of the structural capabilities of containment and drywell, a containment response analysis, and a design comparison to significant plant features-of Grand Gulf. App. Ex. 8-1; Appli-cants Testimony at'19' (Buzzelli).
- 37. Applicants' Preliminary Evaluation Report, in Section 2.0, discusses the design of the distributed igniter system, including-the design criteria, igniter locations, power sup-plies, testing and system operation. The next area, addressed in Section 3.0 of the Report, describes the containment ulti-mate capacity analysis and evaluations of the containment nega-tive pressure capability and the drywell positive and negative pressure capabilities. The third area, covered in Section 4.0, describes the analysis of the containment pressure and tempera-ture response to hydrogen combustion, including the PNPP re-suits provided as Appendix A to the Report. Finally, Section 5.0 of the Report provides a comparison of the PNPP design and supporting analysis with that of Grand Gulf. This comparison
establishes the similarity of systems and analytical results, and provides additional basis for an NRC staff decision to sup-
+.
port full power operation'for PNPP. Supplbhentaryinformation
-to Applicants' preliminary evaluation in the areas of preoperational testing and equipment survivability, and clari-fication of the containment response analysis information, was-submitted on March 21, 1985. App. Ex. 8-1; Applicants Testi-mony at-19-20 (Buzzelli).
- 38. Applicants selected a distributed ignition system as the,most viable concept to control large amounts of hydrogen
_ generated.during a postulated degraded core event. The system-is designed to handle, without loss of containment structural
. integrity, an amount of. hydrogen equivalent to that generated from a metal-water reaction involving up to 75% of fuel cladding surrounding the active fuel region. This is accom-plished by_ burning hydrogen at low concentrations, thereby maintaining the concentration of hydrogen below levels which could potentially threaten containment integrity. Applicants l
Testimony at 29 (Buzzelli); Tr. 3258 (Karlovitz); Notafrancesco p I at 3.
- 39. The technical criteria used in the selection of the l
igniter system for PNPP considered the mitigation effective-ness, consequences of intended or inadvertent operation, reliability, testability, and availability of design and '
i i
4
(_ .
. equipment.- - Following the TMI accident, and prior to Appli-cants' selection of an igniter-based hydrogen control system, the use of igniters to burn hydrogen at low concentrations to control large amounts had been impl'emented as the hydrogen con-trol system at ice condenser plants (McGuire, Sequoyah, D.C.
Cook) and at the first NRC licensed Mark III containment plant (Grand Gulf). Applicants Testimony at 29-30; Tr. 3249-50 (Richardson); 3637-39 (Fuls, Karlovitz, Richardson). There is a general technical consensus today within the nuclear industry that an igniter system constitutes an adequate system for con-
. trolling large amounts of. hydrogen. Applicants Testimony at 30 (Buzzelli).
D. Igniter System Design and Operation
- 40. The thermal glow plug igniters used in the PNPP sys-tem are the same as the igniters commonly used in diesel en-gines. Notafranceso I at 3; Tr. 3249 (Richardson). The glow-plug igniters have been extensively tested and have demon-strated reliable ignition. Applicants Testimony at 30-31 (Lewis and Karlovitz); Tr. 3639-40 (Fuls, Karlovitz).
- 41. The PNPP igniters are spaced throughout the drywell, i ~
l wetwell and upper containment regions of the plant. The ig-niters are maintained at a high surface temperature (1700*)
which assures ignition of hydrogen in a controlled manner at or i
near its lower combustion limit as it is released. Applicants Testimony at 32 (Richardson); Notafrancesco I at 3.
l l
b
- 42. The hydrogen' igniters are powered from safety grade
-power distribution panels and motor control centers through 15 KVA transformers. Each transformer is capable of being powered from one of the e'mergency diesel ~ generators. The 102 igniters are divided into six~ groups of approximately equal number, three-groups in Division 1 and.three groups in Division II.
'Each group is powered from a separate distribution power panel.
2 Applicants Testimony at 32 (Richardson); Notafrancesco I at 4.
4
- 43. The PNPP hydrogen ignition system is designed with suitable redundancy to assure that no single active component failure,. including power supply. failure, will prevent func-tioning of.the system.
The system is designed as a safety i . grade system, and is capable'of operating for the duration of the hydrogen generation event. The igniter assemblies are i
l classified and designed as electrical Safety Class lE and l Seismic Category 1. Applicants Testimony at 33 (Richardson);
see Finding 42.
- 44. Igniter locations were based on criteria that consid-ered potential' hydrogen release locations, appropriate spacing n 'incopen areas, redundancy and potential for pocketing in en-
! closed regions. Based on these criteria, igniters are located i
- in a ring above the suppression pool and throughout the con-tainment and drywell. The igniters are located approximately every 30 feet with alternating divisional power supplies, such 1
l E.:
that a distance of approximately 60 feet may exist if only one emergency power division is available. Two igniters,. one from each power division, are located in enclosed containment areas
.which could accumulate hydrogen. The number and arrangement of igniter assemblies are similar to those at the Grand Gulf.
Applicants Testimony at 33-34 (Richardson); Tr. 3642 (Buzzelli, Richardson).
~45 . OCRE conducted cross-examination concerning a draft .
~
description of the PNPP igniter locations (OCRE Ex. 16). The document was prepared in July 1984. Tr.'3648-49 (Buzzelli).
It has been superceded by the location descriptions contained in the Preliminary Evaluation Report (App. Ex. 8-1). Many of the locations in App. Ex. 8-1 are different than the loc.ations
-in the draft document, and App. Ex. 8-1 contains more precise location descriptions. The changes were made to meet spacing criteria and to ensure availability of support structures. Tr.
3503-05, 3607-08, 3640-42 (Buzzelli).
- 46. Dr. Lewis and Mr. Karlovitz have physically observed the containment and pertinent systems inside the containment at the McGuire PWR ice-condenser plant, and at the Grand Gulf and PNPP BWR 6/ Mark III plants, and have reviewed related documen-tation as part of these reviews. Their review of the PNPP ig-niter system indicates that it will be able safely and effec-tively to burn large amounts of hydrogen. Applicants Testimony at 30-31 (Lewis and Karlovitz).
I i
- 47. Dr. Lewis and Mr. Karlovitz concluded that the ig- ,
niter assembly design and locations at PNPP will ensure an ade-quate flow of hydrogen-air mixtures to the igniters. Appli-cants Testimony at 31 (Lewis and Karlovitz). Neither the spray shield nor the placement of igniter assemblies underneath ceil-ings or near walls will inhibit burning. Tr. 3513-14 (Lewis).
Dr. Lewis and Mr. Karlovitz believe the PNPP igniter system de-sign concept as implemented is adequate to prevent significant accumulations of hydrogen-air mixtures. Applicants Testimony at 31 (Lewis and Karlovitz).
- 48. Dr. Lewis and Mr. Karlovitz concluded that the PNPP igniter system will assure that hydrogen at low (8%) concentra-tion in the presence of air will be ignited by the large number of igniters located at various points in the containment and burn without abrupt pressure rise. This burning would be re-peated as succeeding flammable mixtures are formed. This con-clusion is substantiated by experimental and theoretical data.
Applicants Testimony at 31 (Lewis and Karlovitz); see Finding 100.
- 49. The number and placement of igniters at PNPP are such that local detonations are unlikely to occur. Applicants Testimony at 31. Even if the concentration hydrogen needed to cause a local detonable could occur, the geometry necessary to produce such a local detonation does not exist. Tr. 3635 (Karlovitz).
L _ -_-._ _ _
- 50. There is also no danger of transition from hydrogen deflagration to detonation. For the'PNPP containment design, large volumes of hydrogen-air mixtures with a composition with-
'in the detonable range cannot accumulate in geometrical config-urations condusive to transition. Applicants Testimony at 31 (Lewis.and Karlovitz); Tr. 3634-36 (Karlovitz).
- 51. During the hearing, OCRE expressed a concern about experiments conducted by Dr, John H.S. Lee of McGill University which produced a detonation of hydrogen -at a concentration of 13.8%. Tr. 3522-23. Dr. Lee's experiment.used high explosive e charges (petrol or TNT) to produce the detonation in a tube 28 feet in diameter. The experiment has no application to the PNPP igniter system. Tr. 3524 (Lewis).
- 52. Both preoperational and surveillance testing will be performed on the PNPP igniters to verify operability of the system. Preoperational testing will include energizing one of the two divisions from the control room and verifying that all igniters powered from the associated panel are functional.
Identical procedures will be followed for the remaining ig-niters powered from the other division. Functional testing of the system will verify that the surface temperature'of the ig-niters is adequate, and that the power supply transformers and the igniter transformers are providing satisfactory voltages.
With the testing described above, reliable ignition of hydrogen
-60 ,
J N
at low concentrations will be assured using the PNPP distri-buted igniter system. Applicants Testimony at 34-35 (Richdrdson).
- 53. The NRC Staff reviewed the design of 'he t PNPP distri-buted igniter system. The Staff concluded that the PNPP~ design arrangement ensures that hydrogen will burn reliably and effec-tively by maximizing the number of potential ignition points.
The Staff found the preliminary analysis of the PNPP igniter system design acceptable pending the results of Applicants' completed final analysis. Notafrancesco I at 3-4; Staff Ex. 8 at 6 6-3.
54'. 'The PNPP igniter system is manually placed in service when the reactor water level reaches the top of the active fuel. Actuation at this time allows the operator sufficient time to manually energize the igniters by two on-off handswitches located on a. control room panel. As indicated in Appendix A to the Preliminary Evaluation Report, hydrogen burn-ing occurs no sooner than one hour after the onset of the acci-dent. This is consistent with other generic analyses of the more probable degraded core accidents. Applicants Testimony at 34 (Richardson); App. Ex. 8-1, Appendix A.
- 55. Emergency procedures addressing the operation of the igniter-system, and its relationship to the other elements of L
the combustible gas control system, are currently being developed on a generic basis by HCOG and the BWR Owners Group.
PNPP will have plant specific emergency inst' ructions 'in place prior to exceeding five percent power. Tr. 3424-27 (Buzzelli).
-E. Ultimate Structural Capacity of Containment
- 56. Applicants' preliminary evaluation submitted under the hydrogen rule contains an analysis of the PNPP containment ultimate structural capacity, as well as an evaluation of the pressure capability of the PNPP drywell. Applicants Testimony at 24-25; (Buzzelli); App. Ex. 8-1, Sections 3.1, 3.2 and 5.3; App,. Ex. 8-4.
- 57. The internal pressure capacity of the PNPP contain-ment is analyzed in the G/C Ultimate Capacity Report (App. Ex.
8-4). The Report addresses the internal pressure capacity of the PNPP containment vessel, including all components. App.
Ex. 8-4.
- 58. The G/C Ultimate Capacity Report contains analyses of the internal pressure capacity of the PNPP containment using ASME Service Level C and D limits, as well as mean and lower bound yield values using actual material certifications. App.
Ex. 8-4; Tr. 3253-54, 3283-85, 3583-85 (Alley).
- 59. G/C began its ultimate capacity analysis in 1981 in response to a request to Applicants from the NRC Staff. The
Staff requested Applicants to establish the pressure capacity of the PNPP containment, including the general shell and key components such as the personnel access air lock, equipment hate'hLand the main's' team penetration. The'NRC' Staff requested that this analysis consider as-built material strengths and consider lower-bound and upper-bound material properties. Tr.
3253, 3583 (Alley).
- 60. In 1982, G/C performed additional analyses to address a specific question from the NRC Staff relating to the struc-tural integrity of the PNPP containment for a 45 psig loading at ASME Service Level C limits. At that time, G/C also estab-lished the PNPP containment's maximum pressure capacity at ASME Service Level C and Service Level D limits. Tr. 3253, 3583-85 (Alley).
- 61. The Ultimate Capacity Report contains the results'of all the foregoing G/C analyses using mean and lower-bound val-ues and ASME Service Level C and D limits. App. Ex. 8-4; Tr.
3253-54, 3283-85, 3583-85 (Alley); Yang at 2; Staff Ex. 8 at 6-5.
- 62. _The analytical techniques and results in G/C's ulti-mate capacity analyses have been reviewed and accepted by the NRC Staff. The Staff concluded that these analytical results demonstrate the adequacy'of the as-built pressure capacity of
~
A
the PNPP containment steel shell. Yang at 2; Staff Ex. 8 at 6-5; Applicants Testimony at 24 (Buzzelli); Tr. 3588-89 (Alley).
- 63. The analytical techniques used in G/C's. analysis ade-quately describe the containment response to the structural loads involved. Tr. 3589 (Alley).
- 64. The analysis of the PNPP containment's maximum pres-sure capacity at ASME Service -Level C limits presented in the -
Ultimate Capacity Report used minimum specified material prop-erties in all but one case. In that case, G/C used actual ma-terial properties with suitable margins. Tr. 3589-90 (Alley);
, App. Ex. 8-4, pp. 20-21, Table 10.
- 65. G/C's analysis established that the PNPP containment vessel and all key components meet the Service Level C require-ments of the ASME Code,Section III, Division I, Subarticle NE-3220, considering pressure and dead load alone. Applicants Testimony at 26 (Alley); Tr. 3590 (Alley); App. Ex. 8-4 at 1-2; Yang at 2.
- 66. There is additional pressure-retaining capability in the containment beyond the ASME Service Level C limits refer-enced'in the hydrogen rule. The ASME Code permits higher Ser-vice Level D limits to be used where the priraary intent is to assure that violation of the pressure retaining boundary will L
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i not' occur. Also, the ASME code states that Service D limits s
are appropriate-for extrem, sly low probability postulated 4 ~
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events. 1Due; postulated degraded core accident addressed in the hydrogen; rule would fall in t$is category.
Therefore, use of '
y , w y Service Level C limits to define the PNPP I prsssureyretaining
< capability, rather than the higher, more realistic Service
- Level D limits, represents a conservative approach to assuring Jthat containment integrity will be maintained. Applicants V
-Testimony at 27 ( Alley); Staff Ex. 8 at 6-5; - see Tr. 3236-37 (Board).
- 67. LThe Ultimate Capacity Repcrt indicates that' using-ASME Service Level.C limits, the controlling lower bound pres-sure_ capacity for the PNPP containment is 50 psig for Penetra-
. tion _414. Using the more realistic Service Level D limits, the
- y-Report shows +ttat the controlling containment pressure capacity is 57 psig-for the same penetration. Thus, the conse{vitively-
- calculated ASME Service Level C limits calculated by G/C demon-1.
strate an ample margin above the peak pressure of 21 psig after hydrogen combustion. Applicants.' Testimony at 24, 28 (Buzzelli,
~
_. s .,s Alley); Tr. 3.,585-86 (Alley); App. Ex. 8-4 at 20-23, Table 10; see Finding 101.
\
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- 68. As.part of its analyses, G/C considerec the variation
,of PNPP containment materie.ls strength due to the elevated tem-
, perature from hydrogen burns. For the temperature ranges
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.- =d., ,,__.._._,_m,_.. ,m.,,_,,,,._..-,_m,, , , , , . , , , , , _ , , , , . , , . , . _
a w expected to result following a hydrogen event, the ASME Code provides for a reduction of stress allowables of approximately 10 percent. This reduction applies to the limiting containment-component, penetration, P 414. 'However, in determining the pressure capacity of P 414, G/C's analysis used minimum specified material strength rather than actual material strength. The capacity of P 414 would be 30% higher using its actual material strength. Thus, the 10% reduction is not sig-nificant, and G/C's analysis is conservative. Tr. 3286, 3586-87 (Alley).
- 69. As demonstrated in Applicants' preliminary evalua-tion, the actual pressure capacity of the containment is sig-nificantly (over 3 times) greater than the design level of 15 psig. This is primarily because of the substantial conserva-tisms in the ASME Code-allowable stress limits used in the de-sign of the containment. Applicants Testimony at 26 (Alley);
-Yang at 2;. Staff Ex. 8 at 6-5.
- 70. The Preliminary Evaluation Report demonstrates that the PNPP containment design is adequate to handle negative pressures following the combustion of hydrogen resulting from postulated degraded core accidents. This is primarily because PNPP has redundant vacuum breakers which would alleviate nega-
! tive pressures in the containment. Applicants Testimony at 25 (Buzzelli); App. Ex. 8-1 at 13-15.
L
- 71. Applicants have also evaluated the positive and nega-tive pressure capabilities of the PNPP drywell. Section 3.2 of
'the Preliminary Evaluation Report establishes that the PNPP drywell pressure capability has substantial margin over design
' levels, and can adequately handle calculated pressures from hy-drogen combustion (which are below design levels).
This fact is confirmed by previous NRC-approved evaluations of the Grand Gulf drywell structure, which are applicable to PNPP's similar-drywell design. Applicants Testimony at 25-26 (Buzzelli); App.
Ex. 8-1 at 15-17.
- 72. The NRC Staff reviewed and approved Applicants' PNPP drywell pressure capability evaluation. The Staff concluded that Applicants' referencing of the Grand Gulf drywell analyti- ,
cal results, including the Staff's acceptance of'those results,
'is' appropriate for demonstrating that substantial margin and -
capability above required capacities are expected for-tlie PNPP drywell structure. Yang at 3; Staff Ex. 8 at'6-5.
- 73. When compared with the pressures predicted for a hy-drogen event at PNPP, see Finding 101, Applicants' ultimate ca-
.pacity analyses discussed above demonstrate that the PNPP con-tainment structural integrity would be maintained in the-unlikely event of a degraded core accident generating large.
amounts of hydrogen. Applicants Testimony at 24 (Buzzelli).
p j
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- 74. OCRE expressed a concern at the hearing about the ef-fact on the PNPP containment structural integrity of multiple repairs of certain weld joints, and whether the repairs could cause embrittlement of the base metal. See,'e.g., Tr.
3299-3303. Multiple repairs of the welds in question do not cause degradation of the welds. The type of material used is very ductile and forgiving. It has a.very high weldability and it is easy to weld with no embrittlement from repeated weld re-x pairs. Tr. 3302-03 ( Alley)', 3755 (Wilcox).
- 75. OCRE expressed a concern with the fact that the gen-eral courses of welds in the containment vessel were not sub-
'jected to post-weld heat treatment. Tr. 3303. Such treatment was not required by the fabrication code or ASME Section III.
Tr. 3303 (Alley), 3755 (Wilcox).
- 76. OCRE was concerned about the effect on containment structural integrity of a limited number of inaccessible welds in the PNPP containment which contain potentially rejectable indications based on the ASME Code. See, e.g., 3303-43. The welds were thoroughly reviewed by Aptech Engineering Services under the direction of G/C. Aptech is a specialist in the area of fracture fatigue analyses of the type performed in its re-view of the inaccessible welds. Tr. 3304, 3591-92 (Alley).
e c
- 77. The results of Aptech's review are contained in a re-port entitled " Analysis of Inaccessible and Potentially Re-jectable Defects in Perry Nuclear Power Plant" (July 1983)
(OCRE Ex. 13). Aptech concluded'in its report that the weld indications were acceptable through the life of the plant for the stress levels expected. The Aptech analysis was conserva-tive in that it did not take credit for the concrete in the containment annulus. The design basis loading stresses evalu-ated by Aptech were conservative and significantly exceed the stresses which the welds would experience following the genera-tion and burning of.large amounts of hydrogen. Tr. 3313, 3316, 3592-93 (Alley), 3326 (Buzzelli), 3755 (Wilcox). The Aptech conclusions are not affected by considerations of the elevated temperatures expected following a hydrogen event. Tr. 3590-91 (Alley).
- 78. The NRC Staff reviewed and accepted the Aptech re-port. The Staff's findings are documented in the PNPP SER Num-ber 4 (February 1984). The Staff concluded that the defective welds in question would not impair the integrity of the PNPP containment structure.- Tr. 3732-33 (Yang).
l 79. OCRE expressed a concern about tests conducted by Sandia National Laboratory, in which various containment models were tested to failure. See, e.g., Tr. 3356-58, 3382-87, 3392-3402. The test-to-failure experiments are nct relevant to E
=o . e
'l G/C's ultimate-capacity analyses, which are-based on ASME Ser-l vice Level C. limits well within the. elastic range. The test-to-failure experiments involved significantly higher pres-sures than those applicable to G/C's analysis. The experiments.
involved'non-linear plastic type analyses, which are outside the. range of the applications of'the G/C analyses. Id., Tr.
.3629-30 (Alley).
- 60. OCRE expressed concern with whether G/C's ultimate -
capacity analysis had correlated containment-leakage with in-ternal pressure. Tr. 3271. The ASME Service Level C limits to which'the PNPP containment was analyzed take leakage into con-sideration and assure that there will be a pressure containing barrier with'the primary purpose of containing within leakage limits the results of various accident phenomena. The ASME Code assures that containment. vessels.are designed and' built to safely contain all radioactive ' substances that may be released into the containment atmosphere. Tr. 3271 (Alley). ,
- 81. OCRE expressed a concern about the potential for leakage from equipment. hatch. flange separation. .G/C's analysis considered potential leakage from the equipment hatch, and-foundzthat ASME Service Level C limits were met.
~
G/C's analy-
' sis also looked ~at.the precompression across the seals, and the gap at the flanges.of the equipment hatch. The analysis estab-
.lished'that there is very adequate precompression of the double k' .__ _ _. _ _ _ _ _ .
]
l O ring seal of the equipment hatch to assure springback and to prevent leakage. Thus, equipment hatch flange separation will I
not cause leakage and is not an issue. Tr. 3273-76 (Alley).
- 82. OCRE expressed a concern about the potential for com -
pression set of the PNPP equipment hatch O ring seals due to thermal.and radiation aging, and due to the high temperatures associated.with hydrogen burning. Tr. 3277-80. The evidence shows that the o rings are qualified up to 300' F, and that compression set is not a problem for the environmental condi-tions associated with hydrogen burning. Tr. 3581-82_(Alley),
3664 (Buzzelli). Also,.there are maintenance requirements for the O ring seals which will establish that they have adequate springback. Tr. 3278 (Alley).
- 83. OCRE also expressed a concern about the potential for containment leakage due to wear, roughness,,or irregularity of the seating surfaces between the PNPP equipment hatch flanges.
Tr. 3278-80. G/C reviewed the manufacturer's drawings for the mating surfaces and found that the smoothness,of the surfaces is more than adequate to facilitate the leak tightness of the seals. .Tr. 3583 (Alley)'. The smoothness cf the surfaces is also'one of the maintenance requirements. Tr. 3279 (Alley).
- 84. OCRE also expressed a concern that the inflatable seals, which are used in the PNPP personnel airlock to prevent i
k
leakage,-have poor resistance to severe accident conditions.
OCRE~ cited studies by Argonne National Laboratory in this re-gard. Tr. 3362-73~. The Argonne study dealt with environmental conditions due to long-term, severe accidents, and not with the results of recoverable, degraded-core events,-which are much less severe. The seals will be qualified and Applicants will confirm that the seals can survive the environmental conditions caused by hydrogen burning. Tr. 3362-76, (Buzzelli, Richardson).
F. Containment Response and Equipment Survivability
- 85. As part of its preliminary evaluation of the PNPP distributed igniter system, Applicants analyzed selected acci-dent scenarios for PNPP using the MARCH and CLASIX-3 computer -
programs. Applicants analyzed combustion from postulated hy-drogen releases, and the resulting thermal and pressure envi-ronments and effects on equipment survivability that can be ex-pected in a postulated degraded core event producing large I amounts of hydrogen at PNPP. Applicants Testimony at 37 (Buzzelli); App. Ex. 8-1, Section 4.0 and Appendix A.
- 86. The preliminary evaluation of the PNPP igniter system is based on the analyses of two recoverable degraded-core sce-narios: (1) a small drywell break ("DWB") loss-of-coolant I
accident ("LOCA") with temporary failure of emergency core 1 cooling system ("ECCS") injection, and (2) a transient with a
-i4
stuck-open relief valve ("SORV") accompanied by a temporary failure of ECCS. The selection of these two scenarios was based on industry and NRC considerations of accident event initiators, which have been divided into two general categories
-- those resulting from postulated pipe breaks, and those caused by' plant transients compounded by multiple failures.
Because of analyses showing that scenarios which dominate plant risk'are transient-initiated, the SORV scenario was chosen as the base case recoverable-core event for the PNPP preliminary evaluation. The DWB LOCA scenario was included in the prelimi-nary evaluation in order to consider the potential consequence of' hydrogen release directly to the drywell. Applicants Testi-mony at 37-39 (Holtzclaw); App. Ex. 8-1 at 18-19.
- 87. The two scenarios used in the PNPP preliminary evalu-ation represent the behavior of the reactor system for a postu-
. lated accident resulting in a degraded core, and have been ac-cepted by the NRC staff in previous plant licensi.,o reviews.
The combination of mass and energy releases from these postu-lated events is representative of a wide variety of postulated degraded core situations in which hydrogen generation may be a factor. Applicants Testimony at 39 (Richardson); App. Ex. 8-1, Section 4.2.
- 88. The NRC Staff reviewed and accepted the accident sce-narios which Applicants selected for the PNPP preliminary k -
- A. -
evaluation. She Staff, concluded that the scenarios are repre-
, sentative of-the challenges to the PNPP containment and are
) .therefore acceptable. Notafrancesco II at 2-3; Staff Ex. 8 at
~6-3 to 6-4.
- 89. The:PNPP preliminary evaluation used the MARCH com-
~
puter code to calculate hydrogen releases. The MARCH code was developed by Battelle Columbus Laboratory for the NRC to pro-t- ' vide.the capability te analyse the thermal-hydraulic rea,nnnna of the reactor core, primary. coolant system, and containment to lthe initial' phases of core melt accidents. The code provides a method to obtain hydrogen release rates. The PNPP analysis was based on the reactor coolant system response and releases using results from the MARCH code. Applicants Testimony at 40 (Fuls); Notafrancesco II at 3; Staff Ex. 8 at'6-4.
- 90. . MARCH models the release of hydrogen with the steam
- from whatever openings in the primary system may be appropriate.
for the scenario. The specific transient analyzed under MARCH was the SORV event. For this event, the hydrogen and steam.re-leases are directly introduced into the suppression pool through the-safety relief valves. The PNPP. preliminary evalua-tion of the DWB event modeled the SORV hydrogen and steam re-leases as entering the drywell through the break, and into the suppression pool through the safety relief valve. Using the.
mass and energy releases calculated for the SORV event in the 7
L Q. .
i DWB evaluation is equivalent to assuming a' break the size of the SORV opening. This is conservative since the SORV opening is larger than the normally analyzed small line break LOCA which is the basis for the DWB case. Applicants Testimony at 40 _(Fuls).
- 91. In the PNPP containment response analysis the MARCH results were modified to be consistent with the hydrogen rule.
. Since the original MARCH transient analysis went well beyond a -
recoverable degraded core, the MARCH results used for the PNPP
-preliminary evaluation were modified to simulate core recovery.
The quenching and recovery were not mechanistically calculated.
Instead, the hydrogen release rate was held constant at the previous peak rate, and the hydrogen reaction was terminated when 75% oxidation of the active cladding was reached. Appli-cants Testimony at 40-41 (Fuls); Notafrancesco II at 5.
19:2 . The NRC Staff reviewed and accepted Applicants' use of-the MARCH Code'in the PNPP containment response analysis, including the hydrogen release rates. The Staff found that Applicants' analytical approach provides the most severe chal-
-lenge to the hydrogen control system and maximizes the rate'of heat added to the containment. The Staff accepts the hydrogen releases calculated by Applicants' analysis. The Staff con-cluded that the MARCH Code was an acceptable code to use for the preliminary analysis, and that MARCH was the best I
r l
( '
mechanistic code available at the time Applicants performed their analysis. Further, the Staff has noted that Applicants
.used the same hydrogen and steam release rates as those used for Grand Gulf, d'espite the fact that PNPP has fewer fuel. bun-dies and less zircaloy material to oxidize and would therefore produce less hydrogen than Grand Gulf. The Staff concluded that the PNPP hydrogen and steam rates are bounded by the rates calculated by Grand Gulf, and that the PNPP postulated hydrogen release rates are overestimated. Notafrancesco II~at 4, 6, 8; Staff Ex. 8 at 6 6-4, 6-7.
- 93. The Staff has identified and evaluated a number of shortcomings in the version of the MARCH Code (MARCH 1.1) used in Applicants' preliminary evaluation. Some of the shortcom-ings, such as the treatment of the zircaloy channel boxes, pro-duced conservative results. These outweighed nonconservative results from other identified shortcomings of the Code, such as not including the contribution of actinides. Another shortcom-ing identified, the time step instability of MARCH 1.1, was recognized by Battelle and only stable time steps were used.
The Staff's concern about code sensitivity to user-chosen val-ues of input variables is being addressed to the Staff's satis-faction as part of the final analysis. None of the MARCH 1.1
. shortcomings identified by the Staff affect the Staff's conclu-sion.that MARCH 1.1.was acceptable for the PNPP preliminary evaluation. Notafrancesco II at 4-6.
k
z.; 'i ,
- 94. Dr. Pratt des' tibed difficulties'which have been ex-perienced in-other anal res with the MARCH sub-routine MACE.
Tr.-3700-(Pratt). Howe 3r, Applicants' PNPP preliminary evalu-ation did not use the M':E sub-ro'u tine in its analysis. Tr.
3726 . ( Pratt) ~.
- 95. The PNPP prel ainary evaluation used.the CLASIX-3 computer program to ana /ze the PNPP containment response to
. hydrogen combustion. C \SIX-3 is a multi-volume contai'nment I
code, modified from the 3riginal CLASIX Code used to analyze
- th'e' effects of hydrogen :ombustion on ice condenser plants.
c There are three compart ants in the PNPP CLASIX-3 model: the drywell, wetwell and cc :ainment. CLASIX-3 has the capability to-model MARK-III conta ament plant features (including the suppression pool ~,-: refue ing pool, vacuum breakers, and drywell-
-purge system) while tra -<ing the distribution of the atmosphere constituents, i.e., ox) an, hydrogen and steam. The CLASIX-3 Code also has the capar lity of modeling containment sprays.and structural heat sinks. CLASIX-3 was used in the Grand Gulf containment response ar lysis. Because of the similarity of. -
the Grand Gulf and PNPI plants, Applicants used the same
~CLASIX-3 program,'with adification of input parameters as nec-G 'essary~to account for a ecific PNPP design features. Appli-F cants Testimony at 41-4., 44 (Fuls); App. Ex. 8-1 at 20-21 and Appendix A; Notafrancer o I at 4-5.
~.
r
- 96. CLASIX-3 applies the laws of thermodynamics and uses standard engineering equations for the conservation of mass and energy, chemical' reactions, heat transfer and fluid flow. It uses standard engineering ~ assumptions and practices such as treating-non-condensible gases and highly superheated steam as perfect-gases. The code also uses finite differential heat transfer equations, perfect mixing and a single temperature within each control volume, and a single temperature in each finite element in passive heat sinks. Applicants Testimony'at 42 (Fuls).
13 7 . In the CLASIX-3 analysis, mass and energy released to the containment atmosphere in-the form of steam and hydrogen are provided as input to the code. The burning of hydrogen is calculated in.the code with provisions to vary the conditions.
under which hydrogen is assumed-to burn and conditions at which the burn will propagate to other compartments. At twenty
' minutes into the transient, the igniters and two combustible gas control system compressors are activated and begin pumping-gas from the containment into the drywell. At 30 minutes into the transient, the upper pool begins dumping water into the suppression pool and at 6500 seconds into the transient, drawdown of the suppression pool (reinstatement of injection systems) to refill the reactor vessel begins. Applicants Testimony at 42-43 (Fuls); App. Ex. 8-1 at 20-21 and Appendix A.
k
. o
- 98. In the drywell break case, all reactor system re-leases are initially to the drywell. At twenty minutes into the transient, the Automatic Depressurization System is. assumed to be~ actuated'and 50% of all subsequent releases are assumed to exit the reactor system through the safety relief valves di-rectly into the. suppression pool. All other assumptions are the same as for the stuck open relief valve transient. Id.
- 99. Both scenarios are continued until hydrogen equiva-a lent to 75% of the clad in the active fuel region has been re-acted. At this time, the stuck open relief valve transient is terminated. In order to assess the consequences of the-large accumulation of hydrogen in the drywell at the end of hydrogen release, the drywell break scenario was continued until at least one burn occurred in the drywell. Id.
100. The CLASIX-3 code models deflagration burning, which is the propagation of a slow, discrete-type flame through a flammable mixture. CLASIX-3 does not model diffusion (continu-ous) flames, which will be addressed in Applicants' final anal-
'ysis. Tr. 3551 (Fuls); Applicants Testimony at 46 (Lewis and Karlovitz); Notafrancesco I at 5. Dr. Lewis and Mr. Karlovitz have reviewed the input parameters used in CLASIX-3 to repre-sent ignition and combustion of hydrogen. They have concluded that the parameters are conservative based on theoretical and experimental data. This includes the conservative assumption t
i i
=~
)
in the CLASIX-3 code that there will be ignition and propaga-tion of hydrogen in concentrations at 8.0%. Dr. Lewis and Mr.
Karlovitz confirmed the conservatism in the CLASIX-3 assumption of a flame speed of 6 ft/sec. Dr. Lewis and Mr. Karlovitz also have concluded that the high degree of turbulence in the sce-narios covered by Applicants' preliminary evaluation will as-sure rapid and complete mixing of hydrogen which will result in uniform hydrogen concentrations in any given compartment, as assumed in tne-rNPP CLA5IX-3 analysis. Applicants Testimony at 44-47 (Lewis and-Karlovitz);. Tr. 3520-22, 3616-17, 3626-28 (Lewis), (Richardson).
101. The results of the PNPP CLASIX-3 analysis are de-scribed in Appendix A to App. Ex. 8-1. They show that frequent periodic burns (deflagrations) occur in the wetwell region of the containment for both cases analyzed. In the SORV case, two burns occur in the containment volume. In the DWB case there are no spontaneous burns in the containment volume. Near the end of the analysis of the DWB, the hydrogen concentration in the containment was still below the specified criteria for ig-
[ nition, but for conservatism, the criteria were lowered to pro-duce a burn in order to evaluate the consequences'.of a burn in the containment. In both scenarios there are many brief tem-perature excursions in the wetwell en the order of 800 F, and pressure excursions on the order of 6 psig. For a few burns, l
t l
b
the temperature in the wetwell reaches approximately 1760* F and pressures reach approximately 21 psig for a short period of time due to coincident wetwell and containment burning. Appli-cants Testimony at 47-48 (Fuls); App. Ex. 8-1, Appendix A; Notafrancesco I at 6-7; Staff Ex. 8 at 6 6-7.
102. The peak containment pressures and peak drywell to containment differential pressures for the two cases analyzed are comparable but somewhat lower for PNPP than for Grand Gulf.
The peak temperatures for the PNPP analyses are also comparabl'e in magnitude to those of the Grand Gulf analyses, with the ex-ception of.the wetwell peak temperature in the SORV case. The PNPP wetwell temperature is higher due to a coincident contain-ment and wetwell burn, which did not occur in the Grand Gulf case, due to plant geometry differences. The higher peak wetwell temperature will not have a significant effect on the overall equipment temperature rate because of the duration, number and timing of burns. Applicants Testimony at 48; hpp.
Ex. 8- 1, Sections 4.4.2, 5.5 and Appendix A, Table 18; see Finding 112.
j 103. The NRC Staff has reviewed Applicants' OLASIX-3 con-
[
tainment response analysis used in the PNPP preliminary evalua-tion and has concluded that it constitutes an acceptable demon-strationlof the performance of the PNPP distributed igniter system. Notafrancesco I at 7; Staff Ex. 8 at 6 6-7. The r
k
- e. 9 -
~
Staff concluded that the three-volume model used in the PNPP
~CLASIX-3 analysis is conservative in that it over-predicts the number of burns ~in the wetwell. Tr. 3734, 3749-50
~
The Staff also concluded th'at the containment
^
(Notafrancesco).
-response analysis for Grand Gulf should bound that of PNPP.
Tr. 3690-91 (Notafrancesco); see Applicants Testimony at 48-49 (Richardson).
104. OCRE expressed a concern about a September 1982 re-quest'from the Staff to Applicants for additional information
~
to be used'in the. Staff's evaluation of the-then-planned PNPP hydrogen ignition system. Tr. 3682-84; OCRE Ex. 19' . The information requested in September 1982 goes beyond what the current hydrogen rule requires. In any case, a majority.of the questions in that document have been responded to, and the re-maining questions will be addressed as part of Applicants' i
final analysis. Tr. 3684 (Notafrancesco).
105. OCRE also expressed a concern during the hearing
.,7 about a question to Applicants in an August 1984 NRC informa-tion request. -Tr. 3684-86; OCRE Ex. 20. The NRC question in-
" dicated that CLASIX-3 analyses should not be used as a basis
-for-determining the most severe compartment temperature condi-tions for demonstrating equipment survivability. The Staff be-lieves that although the CLASIX-3 model had a possible non-conservative element, the overall effect of the CLASIX-3 l
i .. P
analysis was conservative. The Staff's concerns about the CLASIX-3 model are being adequately addressed by Applicants as.
part of the final analysis. Tr. 3686-87, 3721-22, 3733-34 (Notafrancesco).
106. OCRE expressed a concern that the PNPP containment response analysis did not utilize the HECTR Code, which had been used by Sandia in analyzing the Grand Gulf containment re-sponse. See, e.g., Tr. 3688-91; OCRE Ex. 21. The HECTR Code used by Sandia did not represent a BWR Mark III configuration.
The version of HECTR used by Sandia in the Grand Gulf review was primitive and crude, it permitted highly conservative as-sumptions to be used, and it over-predicted flame speeds and combustion completeness and under-predicted burn time. More recent HECTR calculations have predicted decreased pressure and temperature. The CLASIX-3 program has been extensively veri-fied as conservative by calculations and comparisons to tests performed by Fenwal and others, and at the Nevada Test Site.
The Staff agrees with Dr. Berman of Sandia that the Sandia's 1983 analysis of Grand Gulf's igniter system does not call into
. question the adequacy of Applicants' containment response anal-ysis in the PNPP preliminary evaluation. Tr. 3621 (Fuls),
3688-91, 3722-26, 3733-46 (Notafrancesco, Pratt). See Finding 114.
L l
107. OCRE also expressed a concern at the hearing that
' Applicants have not adequately analyzed how much hydrogen in the DWB scenario could leak out of the drywell, bypassing the suppression pool. See, e.g., Tr. 3498-3501, 3533-42, 3615-16,
^
3706-13. In a recoverable degraded-core scenario, the vast ma-jority of steam and hydrogen released in the drywell in the DWB scenario would go through the suppression pool, and only a small~ fraction would go out through bypass leakage. GE has an-I' alyzed the potential tor nydrogen Dypass tnrougn tne drywell during a small break LOCA and has found that drywell bypass leakage is of no concern to the operation of the PNPP distri-buted igniter system. Dr. Pratt confirmed that GE's calcula-tions of drywell bypass leakage based on technical specifica-tion limits appear consistent with the types of calculations he has performed, and that GE's conclusions sound reasonable.
Also,.there will be periodic-testing for drywell leakage at PNPP. Finally, even if hydrogen were to leak out through the drywell wall, the hydrogen transport and combustion character-istics analyzed in the PNPP preliminary evaluation would not change. Tr. 3499-3501, 3615-16 (Buzzelli), 3628-29 (Fuls, Holtzclaw), 3726-27 (Pratt).
-108. OCRE also expressed a concern that suppression pool cooling might be diminished if the residual heat removal
("RHR") system is used in the spray mode following the release L
/
of hydrogen. See, e.g., Tr. 3453. There are redundant, safety-grade RHR loops available for suppression pool cooling.
' Active components of the RHR system for decay heat removal are located outside containment and would survive hydrogen burning.
~
Even in the unlikely event that both RHR loops were unavailable for suppression pool cooling, any elevated pressure which might result could be adequately handled by the containment spray, which-is run'through the heat exchanger before being sprayed.
Long term decay heat removal would be ensured. Tr. 3453-54, 3611-13 (Buzzelli, Richardson). In any case, the PNPP prelimi-nary evaluation demonstrates that hydrogen burning occurs early in the event,Las compared with the time of peak suppression pool temperature during the design basis accident events about which OCRE inquired. The hydrogen burning analyzed in the pre-liminary evaluation would have an insignificant effect on the suppression pool temperature. Localized elevated suppression pool temperatures would not affect a hydrogen generation event.
Tr. 3611-13 (Richardson).
109. OCRE also expressed a concern about a 1983 analysis by John M. Humphrey, a former engineer with GE, indicating that 4
containment spray operation might significantly reduce suppres-sion pool mixing effectiveness and lead to pool stratification.
. Tr. 3478-84. Mr. Humphrey's concerns were evaluated for Grand Gulf, and were presented to the NRC Staff and the Advisory Committee on Reactor Safeguards. His concerns were determined not to raise significant safety issues and to be second or third order-effects. In any case, Mr. Humphrey's concerns re-
~
lated to design basis accident considerations and calculations,
. and not to hydrogen generation events. Tr. 3482-85, 3613-14 (Richardson).
110. Applicants have also evaluated on a preliminary basis whether the PNPP containment systems and components necessary to establish and maintain safe shutdown and to maintain con-tainment integrity will be capable of performing their func-tions during and after exposure to the environmental conditions created by the burning of hydrogen. In this evaluation, the thermal environment at PNPP produced by deflagrations has been preliminarily defined using the CLASIX-3 analysis. App. Ex.
1, Section 4.4; Applicants Testimony at 49 (Buzzelli)'. The
- criteria for selecting equipment for the PNPP preliminary eval-uation are consistent with the criteria used in.the NRC-approved Grand Gulf equipment survivability analysis.
Equipment was selected based on its functions during and after postulated degraded core events. The PNPP and Grand Gulf pre-i liminary lists of equipment required to' survive combustion are l
very similar. Applicants Testimony at 50-51 (Buzzelli); App.
Ex. 8-1, Section 4.4.1.
l I-('
7 ,
-111. The PNPP preliminary. evaluation includes a comparison of the temperature profiles for similar equipment at PNPP and Grand Gulf. There was an extensive, NRC-approved program con-ducted at Grand Gulf to evaluate the ability of Grand Gulf equipment to survive hydrogen combustion. This program evalu-ated the ability of equipment to survive the thermal environ-
. ment predicted by the CLASIX-3 computer code. The surface tem-peratures of the Grand Gulf ~ equipment at the end of the hydrogen combustion transient were shown.to remain below tne
. equipment environmental qualification temperature or the subcomponent most sensitive to thermally induced failure was shown to remain below the environmental qualification tempera-ture. The PNPP preliminary evaluation compared the thermal re-sponses of an igniter assembly to the PNPP and Grand Gulf CLASIX-3 environmental temperature profiles. The same igniter assembly heat transfer model and assumptions were used as were used at Grand Gulf. The comparison in the PNPP preliminary evaluation demonstrates that the PNPP CLASIX-3 temperature pro-file will result in lower equipment temperatures than those-produced by the temperatures in the CLASIX-3 profile for Grand
- Gulf. Applicants Testimony at 49-50; App. Ex. 8-1, Section 4.4.2.
112. With regard to the containment temperature predicted by CLASIX-3 for PNPP and Grand Gulf, the higher peak wetwell C
temperature at PNPP (see Finding 102) will not have significant effect on the overall equipment temperature. This is because the burns are of short duration when compared to the time re-quired for heat' transfer'to the equipment and the resulting temperature increase. In addition to peak burn temperature, the number of burns and timing between burns are also important parameters which affect equipment temperatures. Although.com-parable to Grand Gulf, the number of burns at PNPP is lower and the time curation between burns at PNPP is larger than at Grand Gulf, which is expected to result in lower equipment temperd-
.tures. Applicants Testimony at 49 (Richardson); App. Ex. 8-1, Sections 4.4.2, 5.5; see Findings 114, 115.
j 113. Applicants also evaluated PNPP equipment pressure survivability based on pressures calculated in the PNPP con-tainment analysis. The preliminary evaluation demonstrates that the PNPP equipment will survive the peak pressures during hydrogen combustion. App. Ex. 8-1, Section 4.4.3.
114. During the hearing OCRE expressed a concern about whether PNPP essential equipment can be expected to survive diffusion flames produced during a recoverable degraded core event. See, e.g., Tr. 3552. HCCG will be evaluating the ef-fects of diffusion flames on PNEP survivability as part of the final analysis, using a 1/4 scale testing program. The thermal environment due to diffusion flame burning is not expected to L
be a threat to equipment. The use of~more realistic release rates in the final analysis is expected to produce lower tem-
. peratures and a less severe thermal environment than those evaluated by Applicants to date. Tr. 3552-57, 3622-23
-(Richardson), 3749-50 (Notafranesco). In addition, the Staff believes that any potential effect of diffusion. flames is sig-nificantly minimized by shielding of components in the area of
'the diffusion flame environment. Staff Ex. 8 at 6-10.
115. OCRE expressed a specific concern during the hearing about whether electrical cable within the containment would be able to-survive the effects of hydrogen combustion or be sub-ject to secondary fires, citing a paper by Dr. Pratt and others (OCRE Ex. 24) and the results of tests performed by Fenwal and at the Nevada Test site. Tr. 3580-81, 3636-37, 3713-18. The Fenwal tests involved very thin teflon wire with a very small heat sink. This type of wire is typically not used at nuclear plants and is not being used at PNPP. Tr. 3734 (Garg). The Nevada Test Site tests involved cable burning only at hydrogen concentrations of 10% and above, and showed no cable ignition for weak (below 10%) hydrogen concentrations. At 10% concen-tration, burning was for a very short time. Most of the cables were still functional despite the burning. The three or four cables that failed were not Class I-E' cables-and are not the type of cable that would be installed in the PNPP containment.
,,, .1 i
The.few-equipment failures that occurred were due to'installa-tion problems and not to hydrogen.related issues. Tr. 3716-17, 3728-29, 3747-48 (Garg). Evaluations conducted for Grand Gulf
~
have sh'own that there is no potential for secondary fires.for the temperatures predicted by CLASIX-3. Tr. 3580-81
'(Richardson). Finally, the temperature peaks predicted for
-PNPP are intermittent, of'short duration, and not uniform throughout the containment. This provides further confidence
= . . . ,
that the equipment would survive. Tr. 3636-37 (Karlovitz, .
Buzzelli,. Richardson), 3747-50 (Garg, Notafrancesco).
116. OCRE also' expressed a concern at the hearing-about I 'the effects of diffusion flames on the polymeric seals used in
.PNPP lower personnel airlock and equipment hatches. Tr.
3558-59. The seals, which will be' evaluated-for survivability'
~
, as part of the final analysis, are expected to withstand the t- thermal environment from any diffusion flames which might occur. Previous equipment survivability analyses conducted for
- deflagration burning show that the seals do not reach rela-tively high temperatures, and that there is significant margin between. equipment response to hydrogen burn temperatures and the qualification temperatures of the seals. The seals are
"~
next to a large mass of metal which provides a large heat sink.
- 1Durs, the seals are' not a limiting component from an equipment survivability standpoint. Further, the equipment hatch ser.ls i
.. s>
are in an area on the outside of the containment'which would not be directly exposed to the hydrogen burn environment. The containment personnel hatch has two doors, inboard and out-board,.and only the inboard door would see the possible high temperatures from the hydrogen combustion. Tr. 3558-59,
-3623-24 (Richardson).
CONCLUSIONS OF LAW The Board has considered all of the evidence submitted by the parties for this third and final phase of the evidentiary hearing, concerning hydrogen control. Based on the findings of fact set forth herein, which are supported by reliable, proba-tive and substantial evidence in the record, the Board decides i all matters in controversy in favor of authorizing operation of the facility. The Board concludes that, as to the matters I
herein, the Director of Nuclear Reactor Regulation should be authorized, upon making requisite findings with respect to mat-ters not resolved in the Board's partial initial decision, to issue licenses to Applicants to operate the Perry Nuclear Power Plant.
l t'
l i
l V
ORDER WHEREFORE, IT IS ORDERED
- 1. Issue No. 8 (hydrogen control) is found to be without merit and is dismissed.
- 2. Pursuant to 10 C.F.R. 5 2.760(a), this is a partial initial decision that will constitute final action of the Com-minnion forty-five (45) days from the date of issuance unless -
exceptions are taken pursuant to S 2.762 or the Commission 4 ' directs that the record be certified to it.
- 3. Exceptions to this decision or designated portions thereof may be filed with the Commission, in the form required by & 2.762(a), within ten (10) days after service of this deci-sion.
- 4. To pursue an appeal, briefs in support of a party's objection also must be filed, within thirty (30) days after filing the exceptions (or forty days in the case of the Staff of the Nuclear Regulatory Commission). Tne brief must comply with the requirements of $ ?.762.
- 5. Within thirty (30) days of the service of the brief of the appellant (40 days for the Staff), parties may file opposing or supporting briefs that comply with the requirements of $ 2.762.
~
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- 6. Filings that'do not comply with the rule governing appeals may be stricken.
Rcepectfully sibmitted,
- s. ,
SHAW,.PITTMAN,'POTTS & TROWBRIDGE 47 W9 -
Jay E. S4lberg, P . C .- 1 (/
Harry H. Glasspiegel
... Counse..for'Acolicants 1800 M Street, N.W.
Washington, D.C. 20036
, (202) 822-1000 s
Dated: June 3, 1985 %
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APPENDIX A WRITTEN TESTIMONY RECEIVED INTO EVIDENCE Following Wit' ness -Transcript Page' Alley, koger W. 3241 Statement of Qualifi cations 'of Robert W. Alley
" Applicants' Direct Testimony of 3241 Eileen'M. Buzzelli,- John D. Richardson, Kevin W.' Holtzclaw, Roger'W. Alley,
. Bernard Lewis, Bella Ka,rlovitz and G. Martin Fuls on the Preliminary .
Evaluation of_the Perry _ Nuclear Power Plant Hydrogen Control System (Issue #8)"-
Buzzelli, Eileen M. 3241 Statement.of Qualifications of Eileen M. Buzzelli l " Applicants' Direct Testimony of 3241 j Eileen M. Buzzelli, John D. Richardson, i
Kevin W. Holtzclaw,-Roger W. Alley, Bernard Lewis, Bella Karlovitz and
'G. Martin Fulsion the Preliminary Evaluation of the Perry Nuclear Power-Plant Hydrogen Control System (Issue #8)"
Fuls, G. Martin 3241 Statement of Qualifications of G. Martin Fuls
" Applicants' Direct-Testimony of 3241 Eileen M._Buzzelli, John D. Richardson, Kevin.W. Holtzclaw, Roger W. Alley, Bernard Lewis, Bella Karlovitz and l
'G. Martin Fuls on the Preliminary Evaluation of:the Perry Nuclear Power Plant Hydrogen Control System
-(Issue.68)"
k
.A-1 l
b . _ _ _ _ _ _ _ _ _ _ _ .
l s LIP Following Witness Transcript Page Garg, Hukam C. 3676 Statement of Qualifications of Hukam C. Garg
" Testimony of Hukam C. Garg 3676 Regarding Issue #8 (Hydrogen Control)"
Holtzclaw, Kevin W. 3241 Statement of Qualifications of Kevin W. Holtzclaw
" Applicants' Direct Testimony of 3241 Eileen M. Buzzelli, John D. Richardson, Kevin W.~Holtzclaw, Roger W. Alley, Bernard Lewis, Bella Karlovitz and G. Martin Fuls on the Preliminary Evaluation of the Perry Nuclear Power Plant Hydrogen Control System (Issue #8)"
Karlovitz,-Bella 3241 Statement of Qualifications of Bella'Karlovitz
" Applicants' Direct Testimony of 3241 Eileen M. Buzzelli,-John D. Richardson, Kevin W. Holtzclaw, Roger W. Alley, Bernard Lewis, Bella Karlovitz and G. Martin Fuls on the Preliminary Evaluation of the Perry Nuclear Power Plant Hydrogen Control. System (Issue #8)"
Lewis, Bernard 3241 Statement of Qualifications of Bernard Lewis
" Applicants' Direct Testimony of 3241 Eileen M. Buzzelli, John D. Richardson, Kevin W. Holtzclaw, Roger W. Alley, Bernard Lewis, Bella Karlovitz and G. Martin Fuls on the Preliminary Evaluation of the Perry Nuclear Power Plant Hydrogen Control System (Issue #8)"
A-2 k
-.:e Following
-Witness . Transcript Page Notafrancesco, Allen 3676 Statement of Qualifications of
. Allen Notafrancesco
'" Testimony of Allen Notafrancesco. 3276 Regarding Issue #8.(Hydrogen Control),"
" Testimony of Allen Notafrancesco. 3276 on the Hydrogen Control Issues Contained in the Licensing' Board-Contention #8" Richardson, John D. _ 3241 _
Statement of Qualifications of John D. Richardson
" Applicants' Direct Testimony of 3241 Eileen M. Buzzelli, John D. Richardson, Kevin W. Holtzclaw, Roger W. Alley, Bernard Lewis, Bella Karlovitz and G. Martin Fuls on the Preliminary Evaluation of the Perry Nuclear Power Plant Hydrogen Control System (Issue-#8)"
Yang, Li 3676 Statement of Qualifications of Li Yang
" Testimony of Li Yang Regarding .3676 Issue #8 (Hydrogen Control)"
l l
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I k
, a p :. .. p .
ORAL TESTIMONY WITHOUT' WRITTEN TESTIMONY
' Witness Introduced.
on Transcript Page
-Pratt, William'Trevor 3676 (Witness for NRC Staff)
Statement of Qualifications of William Trevor Pratt 3676
~
Wilcox, James-~H. 3752 (Witness for Applicants)
A-4 ,
. APPENDIX B Exhibits Admitted atl Following Exhibit Identified at Transcript- . Transcript Number Description Transcript Page Page' Page App. Ex. 8-1 The Cleveland Electric 3219 3243 3243 -
4 Illuminating Company Preliminary Evaluation of the Perry Nuclear Power. Plant Hydrogen l
Control System, March 21, 1985 App. Ex. 8-2 Letter from M. Edelman 3219 3243 3243 to B.J. Youngblood, dated February 5, 1985 re ,
Perry Nuclear: Power Plant Hydrogen Control Evaluation App. Ex. 8-3 Letter from B.J. 3219 3243. 3243 Youngblood to.M.
Edelman, dated February 20, 1985 re Acceptability of the Scope of Hydrogen Control Design ~and Analytical Information to be Provided to Support Full Power Licensing.of Perry Nuclear Power Plant, Unit 1 App. Ex. 8-4 Letter from M. Edelman 3219 3243 3243 Eto B.J. Youngblood, dated February 11, _1985
-re SER Confirmatory Issue (3) Containment.
Ultimate Capacity Analysis t
.B-1
_ J
._ ~ - . . . .
' Admitted at' FolloEing(
Exhibit- . Identified at. Transcript 1 Transcript Number Description : Transcript Page Page Page Staff Ex. 8 " Safety Evaluation 3675' 3677 not. bound Report, NUREG-0887, into record Supplement No. 6" OCRE Ex. 12 Letter from Mr. C.O. 3261 3263 3263 Thomas to'Mr. G.G.
Sherwood, dated-April 13, 1984 re Request for Additional Information Regarding the' Severe Accident Review of Gessar II OCRE Ex. 13 " Analysis of 3305' 3343 3343 Inaccessible and Potentially Rejectable '
Defects in Perry Nuclear .
Power Plant" authored a by Warren P. McNaughton, Jeffrey R. Egan and Jeffrey D. Byron of Aptech Engineering Services, dated July 1983 OCRE Ex. 14 Letter from-M. Edelman 3377 3378 3378 to Mr. B.J. Youngb]ood, dated May 29, 1984 re -
Piping Design Review +
OCRE Ex. 16 Table 2.2-1, Igniter 3508 '
5 3508 3508 locations from the Perry Nuclear Power' Plant Units 1& 2. Interim Report on the Hydrogen Control System 6
B-2 J
p a
Admitted.at Following. 1 Exhibit . Identified at . Transcript Transcript.
Number -Description -Transcript Page Page- Page OCRE Ex. 17 Attachment.A, " Experimental 3562 3562 3562-
-Study of.Hg Diffusion Flames Burning Above a Pool of; Water", from.the Combex Study of Hydrogen Control at Grand Gulf Nuclear Station, dated 1981 OCRE Ex. 18 NRC Memorandum from John 3680 ~3681 3681 Stefano to-B.J. Youngblood, dated May 4,.1983 re Summary Report of meeting with the Cleveland Electric Illuminating Company (CEI) on Perry j Containment Weld Deficiencies OCRE Ex. 19 Letter from A. Schwencer to 3682 3683 3683
, D. Davidson, dated September 16, 1982 re Request for
, Additional Information Regarding Degraded Core Hydrogen Control for the
! Perry Nuclear Power Plant j (Units 1 and 2)-
OCRE Ex. 20 Letter from B.J. Youngblood 3685 3685- 3685 to M. Edelman, dated August 4
30, 1984 re Request for Additional Information Regarding Hydrogen-OCRE Ex. 21 NUREG/CR.2530 3691 3691 3691
. Review ~of the Grand Gulf Igniter System i
, OCRE Ex. :2 NRC Memorandum 3693 REJECTED, from Marc Wigdor, 3696~
l
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.1 Admitted at' Following Exhibit Identified at Transcript Transcript Number Description- Transcript Page Page: ~Page! '
through Jack Rosenthal, to Brian Sheron, dated October 24, 1984 re Hydrogen Control Owners' Group and-NRC Meeting, October 3 and 4, 1984, discussing the use of BWR Heatup Code OCRE Ex. 23 Draft Report,-"An 3701 REJECTED, Assessment of Postulated 3703 Degraded Core. Accidents in the Grand Gulf Reactor Plant," by R.D. Gasst. of Brookhaven National Laooratory, dated June 1982 OCRE Ex. 24 Paper presented at the 3714 3715 3715 Second International Conference on the Impact of Hydrogen on. Water Reactor Safety, " Electrical Cable Insulation Pyrolysis and Ignition Resulting from Potential Hydrogen Burn Scenarios for Nuclear Containment Buildings,"
by A.L. Berlad, R. Juang and W.T. Pratt i
n a
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION.
~
BEFORE THE ATOMIC ' SAFETY AND LICENSING BOARD
- In the-Matter of ) ~00,C[gTEC
)
THE CLEVELAND ELECTRIC . ) Docket Nos. 50-440
-ILLUMINATING COMPANY, ET AL. 50-441
'89 JllN-5 AU :00
)
(Perry Nuclear Power Plant, Units'l and 2)- 0FF E OF SECRtiAn t 000 ETING A SE8vicl.
BRANCH CERTIFICATE OF SERVICE This is to certify that copies of the foregoing "Appli-cants' Proposed Findings of Fact and Conclusions of Law in the Form of a Partial Initial Decision (Hydrogen Control)" were served by deposit in the United. States Mail, first class, post-age prepaid, and by Federal Express to Ms. Sue Hiatt, this 3rd day of-June, 1985, to those on the attached Service List.
Arrb) .
Harry H Glasspiegelf DATED: _ June 3, 1985 i
i J,
r -
UNITED STATES OF AMERICA ~
NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
.In.the Matter of )
v
-)
THE CLEVELAND ELECTRIC ) Docket Nos. 50-440 ILLUMINATING COMPANY- ) 50-441
)
(Perry Nuclear Power Plant, )
Units'l~and 2) )
, . SERVICE LIST 1
~
Jcmes P. Gleason, Chairman Atomic Safety and Licensing
, 513 Gilmoure Drive Appeal Board Panel
' Silver Spring, Maryland 20901 U.S. Nuclear Regulatory Commission Washington,'D.C. 20555 Mr. Jerry R. Kline Docketing and Service Section i
' Atomic Safety and Licensing Board Office of the' Secretary-U.S. Nuclear Regulatory Commission 'U.S. Nuclear Regulatory. Commission Washington,.D.C. 20555 Washington, D.C. 20555 Mr..Glenn O. Bright Colleen P. Woodhead, Esquire Atomic Safety and Licensing Board . Office of the Executive Legal.
U.S. Nuclear Regulatory Commission Director Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Christine N. Kohl, Chairman Atomic Safety and Licensing Terry Lodge, Esquire l Appeal Board .
Suite 105 HU.S. Nuclear Regulatory Commission 618 N. Michigan Street Washington, D.C. .20555 Toledo, Ohio 43624
, Dr. W. Reed Johnson Donald T. Ezzone, Esquire Atomic Safety and Licensing Assistant Prosecuting Attorney i^ Appeal Board Lake County Administration U.S. Nuclear Regulatory Commission Center Washington, D.C. 20555 105 Center Street Painesville, Ohio 44077 Gary J. Edles, Esquire
! Atomic Safety and Licensing Atomic Safety and Licensing Appeal' Board Board Panel
.U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission
- . Washington, D.C. 205a5 Washington, D.C. 20555 Jchn G. Cardinal, Esquire Ms. Sue Hiatt Prcsecuting Attorney 8275 Munson Avenue r
Achtabula County Courthouse Mentor, Ohio 44060 Jcfferson,~ Ohio 44047
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