ML20041F659
| ML20041F659 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/10/1982 |
| From: | Baxter T METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| Shared Package | |
| ML20041F656 | List: |
| References | |
| NUDOCS 8203170275 | |
| Download: ML20041F659 (132) | |
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UNITED STATES OF AMERICA '82 MTR 15 N035 NUCLEAR REGULATORY COMMISSION f>; i E'E j ,'
BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
LICENSEE'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE ATOMIC SAFETY AND LICENSING BOARD'S PARTIAL INITIAL DECISION ON PLANT DESIGN AND PROCEDURES, SEPARATION, AND EMERGENCY PLANNING ISSUES SHAW, PITTMAN, POTTS & TROWBRIDGE George F. Trowbridge, P.C.
Thomas A. Baxter , P.C.
Robert E. Zahler Delissa A. Ridgway Counsel for Licensee March 10, 1982 8203170275 820310 PDR ADOCK 05000289 G PDR
l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
LICENSEE'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE ATOMIC SAFETY AND LICENSING BOARD'S PARTIAL INITIAL DECISION ON PLANT DESIGN AND PROCEDURES, SEPARATION, AND EMERGENCY PLANNING ISSUES SHAW, PITTMAN, POTTS & TROWBRIDGE George F. Trowbridge, P.C.
Thomas A. Baxter, P.C.
Robert E. Zahler Delissa A. Ridgway Counsel for Licensee March 10, 1982
1 TABLE OF CONTENTS Page(s)
INTRODUCTION................................................. 1 EXCEPTION NO. l....................................,......... 7 Statement of the Case................................... 7 Questions Presented.................................... 15 Argument............................................... 16 A. THE LICENSING BOARD ERRED IN FINDING THAT ADDITIONAL INSTRUMENTATION IS NECESSARY AT TMI-l TO DETECT INADEQUATE CORE COOLING................................. 16 B. THE LICENSING BOARD ERRED IN FINDING THAT REACTOR WATER LEVEL INSTRUMENTATION OR ITS EQUIVALENT IS NECESSARY FOR THE LONG-TERM OPERATION OF TMI-1................. 28 C. THE LICENSING BOARD'S DECISION SHOULD NOT BE READ TO REQUIRE ANY PARTICULAR DESIGN CRITERIA FOR ANY REACTOR COOLANT LEVEL OR OTHER INSTRUMENTATION INSTALLED AT TMI-1........................... 43 EXCEPTION NO. 3.............................................
45 Statement of the Case.................................. 45 Questions Presented.................................... 50 Argument............................................... 51 A. THE LICENSING BOARD ERRED BY PLACING UNDUE RELIANCE ON GENERIC GUIDANCE DOCUMENTS THAT DO NOT CONSIDER OR EVALUATE THE ADEQUACY OF LICENSEE'S EMERGENCY RESPONSE STAFFING AND ORGANIZATION AT TMI-1........................ 51 B. THE LICENSING BOARD EITHER MISUNDERSTOOD OR IGNORED LICENSEE'S REASONS FOR RETAINING DECISIONMAKING AUTHORITY ONSITE WITH THE EMERGENCY DIRECTOR DURING THE EARLY HOURS OF AN EMERGENCY....... 54 C. THE LICENSING BOARD'S DIRECTION THAT DURING THE EARLY HOURS OF AN EMERGENCY ~
DECISIONMAKING AUTHORITY BE TRANS-FERRED OFFSITE TO THE EMERGENCY SUPPORT DIRECTOR CONSTITUTES AN IMPERMISSIBLE INTRUSION INTO THE LEGITIMATE INTERNAL MANAGEMENT RES PONSIBILITIES OF LICENSEE. . . . . . . . . . . . . . . . 60 CONCLUSION.........................................'......... 64 APPENDIX TABLE OF AUTHORITIES Page(s)
Cases:
Consumers Power Company (Midland Plant, Units I and 2), ALAB-379, 5 N.R.C. 565, 570-71, n.18 (1977)............................................ 30 Duke Power Company (Catawba Nuclear Station, Units 1 and 2), ALAB-3 55, 4 N.R.C. 397, 402-05 (1976).................................. ....... 43 Federal Tort Claim of General Public Utilities Corp., CLI-81-10, 13 N.R.C. 773, 775 (1981)............ 63 Metropolitan Edison Company, et al. (Three Mile Island Nuclear Station, Unit No. 1)
Commission Order, 44 Fed. Reg. 40461 (1979)............. 2 Order and Notice of Hearing, CLI-79-8, 10 N.R.C. 141 (1979)........................... 2,3,4 1
Order, CLI-81-3, 13 N.R.C. 291, 295 (1981).............. 4 Order, CLI-81-19, 14 N.R.C. 304 (1981).................. 4 Order, CLI-81-34, 14 N.R.C. (December 23, 1981).............................................. 5 l
Licensing Board " Memorandum and Order on Effect of New Emergency Planning Regulations," March 23, 1981...................... 52 Licensing Board " Partial Initial Decision
, (Background and Management Issues),"
! LBP-81-32, 14 N.R.C. 381 (1981)................. 1,12 Licensing Board " Partial Initial Decision (Plant Design and Procedures, Separation, and Emergency Planning Issues),"
LBP , 14 N.R.C. December 14, 1981)............................................... passim Licensing Board " Memorandum and Order Modifying Partial Initial Decision of December 14, 1981," LBP , 15 N.R.C.
(January 26, 1982).............................. 1,21
-iii-
l Licencing Board "M2morcndum cnd Order Satting Preliminary Hearing," March 2, 1982................ 6 Niagara Mohawk Power Corporation (Nine Mile Point Nuclear Station, Unit 2), ALAB-264, 1 N.R.C. 347, 357 (1975)............................... 42 Northern Indiana Public Service Company (Bailly Generating Station, Nuclear 1), ALAB-303, 2 N.R.C. 858, 867 (1975)............................ 42,43 Public Service Company of New Hampshire, et al.
(Seabrook Station, Units 1 and 2),
CLI-76-17, 4 N.R.C. 451, 462 (1976).................... 30 Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station), ALAB-655, 14 N.R.C. 799, 815-816 (1981), and LBP-81-12, 13 N.R.C. 557, 566 (1981)................... 14 Tennessee Valley Authority (Fartsville Nuclear i Plant, Units lA, 2A, 1B and 2B), ALAB-367, 5 N.R.C. 92, 94 n.4 (1977)............................. 43 Vermont Yankee Nuclear Power Corporation (Vermont Yankee Nuclear Power Station),
ALAB-229, 8 A.E.C. 425, 440 (1974)..................... 30 i
Statutes:
l Atomic Energy Act of 1954, as amended, SS 103(a), 161 42 U.S.C. SS 2133(a), 2201............................. 61 Regulations:
i 10 C.F.R. S 50.46........................................... 27 10 C.F.R. S 50.46(b)(1)..................................... 27 ,
10 C.F.R. S 50.47(b)........................................ 61 10 C.F.R. S 50.47(b)(2)..................................... 47 10 C.F.R. S 50.47(b)(8)..................................... 47 10 C.F.R. S 50.54(s)(2)(ii)................................. 61 10 C.F.R. Part 50, Appendix A............................... 61 10 C.F.R. Part 50, Appendix E............................ 47,61 10 C.F.R. Part 50, Appendix E, S IV.E.8..............'....... 47
-iv-
l March 10, 1982 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
LICENSEE'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE ATOMIC SAFETY AND LICENSING BOARD'S PARTIAL INITIAL DECISION ON PLANT DESIGN AND PROCEDURES, SEPARATION, AND EMERGENCY PLANNING ISSUES INTRODUCTION On December 14, 1981, the Atomic Safety and Licensing Board in this proceeding ("the Licensing Board") issued a Partial Initial Decision on Plant Design and Procedures, Separation, and Emergency Planning Issues. LBP , 14 N.R.C. (1981). The procedural history of this case is 1 The Licensing Board's " Memorandum and Order Modifying Partial Initial Decision of December 14, 1981," amends the Partial Initial Decision in several respects. LBP , 15 N.R.C. (January 26, 1982).
2 On August 27, 1981, the Licensing Board issued a Partial Initial Decision on Procedural Background and Management Issues. LBP-81-32, 14 N.R.C. 381 (1981). The Licensing (continued next page)
i documented well in the Licensing Board's first Partial Initial Decision. See I.D., 11 1-36, 14 N.R.C. at 386-399.
Consequently, our summary here of the underlying framework for this appeal may be exceedingly brief.
Licensee holds Facility Operating Licenses DPR-50 and DPR-73 for the Three Mile Island Nuclear Station, Units 1 and 2, respectively ("TMI-1" and "TMI-2"). Each plant employs a pressurized water reactor ("PWR") designed by Babcock & Wilcox Company ("B&W") and is located at Licensee's site 10 miles southeast of Harrisburg, Pennsylvania. I.D., 11 1 and 2, 14 N.R.C. at 386.
At the time of the TMI-2 accident on March 28, 1979, TMI-1 was in a power ascension mode after completing a refuel-l l ing outage and was immediately shut down oy Licensee. On July 2, 1979, the Commission issued an immediately effective order l directing that TMI-l remain shut down until further order of the Commission itself, and announcing its determination that it is in the public interest that a hearing precede restart of the facility. I.D., 14, 14 N.R.C. at 386; 44 Fed. Reg. 40461 (1979).
On August 9, 1979, the Commission issued an Order and Notice of Hearing, CLI-79-8, 10 N.R.C. 141 (1979), in which it (continued)
Board's findings on management issues are under review by another Atomic Safety and Licensing Appeal Board, which includes two members of this Appeal Board -- Chairman Edles and Judge Buck. The Licensing Board's two partial initial
. decisions employ a single, sequential paragraph numbering system, however, so that we may cite to them as an Initial Decision -- e.g., "I.D., 1 .
j -
ordered a hearing and established the Licensing Board to rule on petitions to intervene, conduct the hearing, render an initial decision and to certify the record to the Commission itself for final decision. The Order and Notice include a list of "short-term actions" recommended by the Director of Nuclear Reactor Regulation ("NRR") to be required of Licensee to resolve the concerns discussed by the Commission and to permit a finding of reasonable assurance that TMI-l can safely resume operation. CLI-79-8, 10 N.R.C. at 144-145. The Commission also included a list of "long-term actions" recommended by the Director of NRR to be required of Licensee to resolve addi-tional concerns which, though they need not be resolved prior to resumption of operation at TMI-1, must be ss:isfactorily addressed in a timely manner and to permit a finding of reasonable assurance of the safety of long-term operation. Id.
at 145.
The Commission provided that the subjects to be considered at the hearing shall include:
(1) Whether the "sbort term actions" recommended by the Director of Nuclear Reactor Regulation (set forth in Section II of this Order) are necessary and sufficient to provide reasonable assurance that the Three Mile Island Unit 1 facility can be operated without endangering the health and safety of che public, and should be required before resumption of operation should be permitted.
(2) Whether the "long-term actions" recom-mended by the Director of Nuclear Reactor Regulation (set forth in Section II of this Order) are necessary and
sufficient to provide reasonable assurance that the facility can be operated for the long term without endangering the health and safety of the public, and should be required of the licensee as soon as practicable.
Id. at 148. The Commission further held that if the Licensing Board should issue a decision authorizing resumption of operation upon completion of certain short-term actions by Licensee and a finding that in its judgment Licensee is making reasonable progress toward completion of certain long-term actions, the Commission would issue an order within 35 days after such decision, in which the Commission would decide whether the portion of the order suspending operation would remain immediately effective. Id. at 149, as modified by CLI-81-3, 13 N.R.C. 291, 295 (1981). The Commission specifi-cally stated that its decision on that question would not affect direct appellate review of the merits of the Licensing Board's decision. CLI-79-8, 10 N.R.C. at 149.
On August 20, 1981, the Commission modified its Order and Notice of Hearing to provide that an Atomic Safety and Licensing Appeal Board be established to hear initial appeels in this proceeding, rather than having the record certified by the Licensing Board directly to the Commission for final decision. The Commission emphasized that it intends to decide whether any Licensing Board decision authorizing operation of TMI-l should be effective during the pendency of any appeals.
g.
CLI-81-19, 14 N.R.C. 304 (1981).
I _
On December 23, 1981, the Commission issued a further Order providing that in this proceeding the Atomic Safety and Licensing Appeal Board does not have the authority to stay the Licensing Board's decision, when and if it becomes effective.
"The Commission is the exclusive administrative body with the power to determine whether Unit one may restart during the pendency of any possible appeals of a Board decision before the Atomic Safety and Licensing Appeal Board." CLI-81-34, 14 N.R.C. _ (1981).
The hearing before the Licensing Board recognized a grouping of intervenor contentions and other hearing issues into four major categories:
Plant design and procedures.
Separation of TMI-l and TMI-2.
Management qualifications of Licensee.
Emergency planning.
I.D., 1 29, 14 N.R.C. at 396. This appeal addresses the Licensing Board's decision only on plant design and procedures, separation, and emergency planning issues.
The Licensing Board concluded its December 14, 1981 Partial Initial Decision as follows:
In Part II relating to plant design and procedures, in Part III relating to the separation of the Three Mile Island nuclear units, and in Part IV relating to emergency planning in the vicinity of Three Mile Island, we have found various deficiencies in design, procedures and planning which must be corrected before restart. These corrections in the form of Licensee commitments, NRC Staff requirements and Board-imposed condi-tions provide reasonable assurance that, with respect to the issues decided in this Partial Initial Decision, Three Mile Island Unit No.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I
1 .
I 1 can be operated in the short term without f endangering the health and safety of the public. The Board has also found in Parts II, III, and IV that the Licensee has made reasonable progress with respect to various necessary and sufficient long-term actions which, relative to the issues decided, provide reasonable assurance that Three Mile Island Unit No. I can be operated in the long '
term without endangering the health and safety of the public.
I.D., 1 2024. The Licensing Board recommended that the pendency of the reopened proceeding on operator license examination cheating (a management issue) should not be a bar to the restart of TMI-l up to five percent of design power.
Id. at 1 2026.
Several matters related to plant design and proce-dures and separation issues are now pending before the Licensing Board. One is the issuance of a final Licensing Board decision on which of the Licensee commitments, Staff requirements, and Licensing Board conditions should be made license conditions. See I.D., 1 1217. The motions by intervenors Steven C. Sholly and Union of Concerned Scientists
("UCS") to reopen the evidentiary record on plant design and procedures issues is under active consideration. See Licensing Board " Memorandum and Order Setting Prelininary Hearing," March 2, 1982. Finally, Licensee is moving the Licensing Board to reconsider and/or clarify one of its proposed requirements for the separation of TMI-l and TMI-2 (fuel handling building ESF filter system).
EXCEPTION NO. 1 1
The decision by the Licensing Board to require additional instrumentation to detect inadequate core cooling, in the form of reactor coolant level instrumentation or its equivalent, is not based on the whole record and is not supported by reliable, substantial and probative evidence. See PID 11 630-705.
Statement of the Case The Licensing Board admitted three intervenor I
contentions on the subject of detecting inadequate core cooling. ANGRY (Anti-Nuclear Group Representing York)
Contention No. V(B) stated as follows:
The NRC Order fails to require as conditions for restart the following modifications in the design of the TMI-l reactor without which there can be no reasonable assurance that TMI-l can be operated without endangering the public health and safety:
(B) Installation of instrumentation pro-viding reactor operators direct informa-tion as to the level of primary coolant in the reactor core.
Mr. Sholly withdrew his Contention 6(b) in a written memorandum dated December 23, 1980, and UCS withdrew its Contention 7 by letter dated January 5, 1981.3 Subsequently, no intervenor participated in the evidentiary sessions at which the Licensee and Staff testimony on this issue was heard; nor did any 3 Licensee and the NRC Staff both responded to all three contentions with direct testimony filed prior to the withdrawals. See I.D., 1 631.
intervenor submit proposed findings on inadequate core cooling issues. Consequently, the Licensing Board's decision on the detection of inadequate core cooling is not directed toward the intervenor contentions, but rather toward the issues raised by the Commission's August 9, 1979 Order and Notice of Hearing.
See I.D., 1 632.
The Commission's Order and Notice of Hearing identified the short-term and long-term actions recommended by the Director of NRR to be required of Licensee. The Director of NRR has recommended, among other things, that as a short-term action Licensee should be required to comply with the Category A recommendations of NUREG-0578 4 prior to resump-tion of operation, and that as a long-term action Licensee should be required to comply with the Category B recom-mendations of NUREG-0578.
Section 2.1.3.b, entitled " Instrumentation for Detection of Inadequate Core Cooling for PWRs and BWRs,"
includes both short-term and long-term recommendations. As the Licensing Board noted, the TMI-2 Lessons Learned Task Force adopted the following positions:
- 1. Licensee shall develop procedures to be used by the operator to recognize inadequate core cooling with currently available instrumentation. The licensee shall provide a description of the existing instrumentation for the operators to use to recognize these 4 NUREG-0578: TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations (July 1979).
conditions. A detailed description of the analyses needed to form the basis for operator training and procedure development shall be provided pursuant to another short-term requirement,
" Analysis of Off-Normal Conditions, Including Natural Circulation" (see Section 2.1.9 of this appendix).
In addition, each PWR shall install'a primary coolant saturation meter to provide on-line indication of coolant saturation condition. Operator instruc-tion as to use of this meter shall include consideration that is not to be used exclusive of other related plant parameters.
- 2. Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the precedin; section giving an unambiguous, easy-to-interpret indication of inadequate core cooling.
A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.
I.D., 1 633; NUREG-0578 at A-12.5 The Task Force explained the purpose of its recom-mendation as follows:
The purpose of this recommendation is to provide the reactor operator with 5 As the Licensing Board correctly noted, the development of the design and implementation schedule for the new instrumentation under Position 2 was identified as a Category A recommendation in NUREG-0578, with only the installation left over for Category B. The Staff, however, subsequently has treated all of Position 2 as a long-term recommendation. I.D.,
5 636.
_9_ .
Instrumentation, procedures, and training j necessary to readily recognize and implement actions to correct or avoid conditions of inadequate core cooling.
NUREG-0578 at A-11. It is important to note here, as we discuss further below, that the Task Force was concerned not with instrumentation alone, but with procedures and training to detect inadequate core cooling as well.
It is true, as the Licensing Board recited, that the Task Force suggested reactor vessel water level instrumentation as additional means for detection which should be studied.
I.D., 1 634; NUREG-0578 at 8 and A-11. The Task Force also stated that:
A number of ideas have been discussed for the second stage by the NRC Division of Reactor Safety Research, the ACRS, and the reactor vendors. Some of the possibilities include pressure differential cells, con-ductivity probes, heated thermocouples, ultrasonic sounding, as well as gamma and neutron void detectors. However, we conclude that detailed engineering evaluation is required.before design requirements for a direct level measurement system can be specified.
NUREG-0578 at A-12. See also, I.D., 1 635.
Subsequent to the issuance of the Commission's order and Notice of Hearing, Licensee prepared and filed with the Staff its " Report in Response to NRC itaff Recommended Requirements for Restart of Three Mile Island Nuclear Station Unit 1." See Lic. Ex. 1 ("the Restart Report"). The Restart Report later was amended to include Licensee's response to the long-term recommendations of section 2.1.3.b of NUREG-0578 as
l they relate to additional instrumentation. The response included B&W's " Evaluation of Instrumentation to Detect Inadequate Core Cooling, Prepared for 177 Owners Group," August 15, 1980. The following methods of detecting inadequate core cooling were examined in this evaluation: (1) existing core thermocouples; (2) additional axial core thermocouples; (3) ultrasonic reactor vessel level indication; (4) neutron or gamma beam reactor vessel level indication; and (5) differ-ential pressure transmitters for reactor vessel indication.
The B&W evaluation concluded taat none of the proposed methods of detection would meet all of the Staff's criteria.6 The report also concluded that each proposed reactor vessel level measurement system concept fails to provide any additional aid to the operator for detection of inadequate core cooling, and that the potentially ambiguous information provided by such instrument systems could lead to unsafe and incorrect actions if the operator acted on the level indication. Lic. Ex. 1, Supp. 1, Part 2, Answer to 095; Tr. 10,646 (Jones); I.D.,
1 667.
On September 15, 1980, as directed by the Licensing Board in its Memorandum and Order of August 15, 1980, Licensee filed its written direct testimony in response to the intervenor contentions on detection of inadequate core cooling. I 6 Staff " criteria" for evaluating reactor water level instrumentation were being provided to applicants and licensees by guidance letters.
t Licensee presented a panel of three witnesses which included the Manager of GPU Nuclear Corporation's Systems Engineering Department (Keaten), the TMI-l Supervisor of Operations (M. Ross, a licensed senior reactor operator), and a Supervisory Engineer of B&W's ECCS Analysis Unit (Jones).
Keaten et al., ff. Tr. 10,619. In that testimony, Licensee took the position that reactor vessel water level instru-mentation had not been shown to be needed at TMI-l and conse-quently that a decision should not now be made to install such a system, either prior to restart or thereafter. Id.
The Staff did not file its testimony in response to these contentions on September 15, 1980. On December 1, 1980, the Staff filed the testimony of Mr. Phillipc in response to the contentions on reactor water level instrumentation. In that testimony, Mr. Phillips neither acknowledged nor responded to Licensee's pre-filed testimony of September 15, 1980. He reported that the Staff had found Licensee's justification for no additional instrumentation to be " unacceptable," and stated that "it is likely that a water level measurement system will be required, but not necessarily prior to restart."
Phillips-1,0 ff. Tr. 10,807, at 9.
7 In its decision on management issues, the Licensing Board made a special point of recording its favorable impression of Mr. Ross, garnered from his many appearances in this hearing on a variety of design, operating procedures, and operator training issues. See LBP-81-32, I.D. 1 155, 14 N.R.C. at 439-440 (1981).
8 NRC Staff Testimony of Laurence E. Phillips Regarding Reactor Water Level Instrumentation ("Phillips-1").
On December 16, 1980, Licensee raised with the Licensing Board, at the hearing, its view that the Staff had been unwilling to join issue on whether additional instru-mentation should be required in the long term, and that the Staff testimony of December 1, 1980, neither took a position on that question and defended it, nor explained why Licensee's position was unacceptable to the Staff. After considerable discussion and inquiry by the Licensing Board, the Staff agreed to report a position to the Licensing Board and to explain it.
See Tr. 8459-77. 1 In a second piece of testimony, filed on December 22, 1980, Staff witness Phillips reported the Staff's belief that reactor vessel level information will enhance the operating safety of PWRs. Phillips-2,9 ff. Tr. 10,807, at 5. When he appeared for cross-examination on January 21 and 22, 1981, Mr.
Phillips testified that the Staff still had not made a defini-tive determination that additional instrumentation was needed but, incongruously, testified that no additional instru-mentation was not an acceptable possibility. Tr. 10,840, 42-43 (Phillips).
Following observations by the Licensing Board, after hearing the testimony of Mr. Phillips, that it appeared the Staff had not decided water level indication is necessary, and 9 NRC Staff Testimony of Laurence E. Phillips, Supplementary Testimony to that of Laurence E. Phillips filed December 1, 1980 Regarding Water Level Instrumentation ("Phillips-2").
that the Licensing Board needed a very careful explanation of precisely what the Staff believes and the reasons for it (Tr.
10,886-88), the Staff filed on March 11, 1981, the testimony of Dr. Ross -- the third piece of Staff testimony. Ross, ff. Tr.
15,915. When he appeared for cross-examination on. March 19 and 20, 1981, Dr. Ross acknowledged that his testimony announced for the first time here the position that water level instru-mentation is necessary to provide reasonable assurance of no undue risk to the public health and safety.10 Tr. 15,929-31 (D. Ross). The Staff position still is not without ambiguity, however, since elsewhere in his written testimony Dr. Ross states: "The staff requirement is for additional instru-mentation for detection of ICC. The preferred technique is monitoring of the reactor coolant system inventory." Ross, ff.
Tr. 15,915, at 10.
Gradually, then, the issue was joined between Licensee and the NRC Staff as to the long-term recommendations of section 2.1.3.b of NUREG-0578.11 There is, however, no 10 Dr. Ross's testimony otherwise is slightly misleading where he states that the purpose of his testimony is to justify the Staff position -- as if the position itself were already well known and communicated to and understood by all. See Ross, ff.
Tr. 15,915, at 2. An Appeal Board in another proceeding on the adequacy of post-TMI-2-accident modifications at another B&W designed reactor observed that the NRC Staff, in that case, at no time stated that reactor vessel level indication was "needed." Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station), ALAB-655, 14 N.R.C. 799, 815-816 (1981). The record in that proceeding was compiled in licensing board hearings held from February to May, 1980. See LBP-81-12, 13 N.R.C. 557, 566 (1981).
11 Following the hearing, the Commonwealth of Pennsylvania, which actively participated in the cross-examination of (continued next page)
dispute about, and this appeal does not challenge, the two critical findings by the Licensing Board which are important to the restart of TMI-1. First, the Licensing Board found that implementation of the short-term recommendations of NUREG-0578 section 2.1.3.b will be adequate to protect the health and safety of the public in the short term. I.D., 1 642. Second, the Licensing Board found that Licensee has demonstrated reasonable progress in meeting the long-term recommendations of NUREG-0578 section 2.1.3.b. I.D., T 672.
Questions Presented
- 1. Did the Licensing Board err in finding that additional instrumentation to detect inadequate core cooling is necessary at TMI-l?
- 2. If not, did the Board err in finding that reactor coolant level instrumentation or its equivalent is the addi-tional instrumentation to detect inadequate core cooling which is needed at TMI-l?
- 3. In any case, should the Licensing Board's decision be read to proscribe any particular design criteria for reactor coolant level or any other additional instrumentation at TMI-l?
(continued)
Licensee and Staff witnesses on this issue, took the position of recognizing that a coolant level meter would be desirable for the long term, but of urging that further generic studies and testing be undertaken by the Staff prior to a commitment by Licensee. See I.D., f 643.
Argument A. THE LICENSING BOARD ERRED IN FINDING THAT ADDITIONAL INSTRUMENTATION IS NECESSARY AT TMI-l TO DETECT INADEQUATE CORE COOLING The Licensing Board never found from the, language of section 2.1.3.b of NUREG-0578 alone that additional instru-mentation is necessary to detect inadequate core cooling.
Indeed, the Task Force appears to have been aiming for enhanced means of detection, but it did not limit its focus to instru-mentation alone. Procedures and training to recognize and respond to inadequate core cooling conditions received equal emphasis. See supra at 9, 10. There is ample room to conclude that the Task Force believed that improvements to procedures and training were more greatly needed than changes in instru-mentation:
With the hindsight of TMI-2, it appears that the as-designed and field-modified instrumentation at Three Mile Island Unit 2 provided sufficient information to indicate reduced reactor vessel coolant level, core voiding, and deteriorated core thermal conditions.
NUREG-0578 at A-ll. The Task Force concluded that the problem of unrecognized inadequate core cooling and low water level in the reactor vessel at TMI-2
. . . was the result of a combination of factors including an insufficient ranae of existing instrumentation, inadequate emergency procedures, inadequate operator training, unfavorable instrument locat.;on (scattered information), and perhaps insufficient instrumentation.
Id. (Emphasis added.]
Even the NRC Staff, following the conclusion in its safety evaluation report that the detection of reduced coolant level or the existence of core voiding at TMI-l can be readily determined with the saturation meter and other existing instrumentation, stated:
The operator must be made aware of the existing information and how to interpret it correctly. The burden of showing a marked improvement in the operator's ability to quickly recognize a condition of inadequate core cooling, and his ability to act upon this information, lies with improvement to the operator's training and instruction rather than the instrumentation.
Staff Ex. 1 at C8-16.
Licensee witness Keaten, who te'stified frequently before the Licensing Board on design issues and who obviously has studied the TMI-2 accident a good deal, testified repeatedly that the principal lesson learned from that accident was operator training and procedures, and not hardware changes.
See, e.g., Tr. 10,683-84 (Keaten).
Licensee has not questioned the conclusion that a lesson learned from the TMI-2 accident is the need to provide the reactor operator with instrumentation, procedures, and training necessary to readily recognize and implement actions to correct or avoid conditions of inadequate core cooling.
Neither does Licensee question the desirability of the long-term objective articulated by the Task Force in section 2.1.3.b of NUREG-0578 -- development of an easy-to-interpret, unambiguous indication of inadequate core cooling. Licensee's
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ - _ _ - )
position before the Licensing Board was that through improvements to the operating procedures and to operator training at TMI-l a marked enhancement of detection capability for inadequate core cooling has been achieved, and that no additional instrumentation has yet been found which would add to that capability.
We start, then, with the proposition that the Commission's Order and Notice of Hearing, including its reference to the NRR Director's recommendation for imple-mentation of NUREG-0578, created no presumption in favor of installation of unspecified additional instrumentation to detect inadequate core cooling.12 Consequently, the NRC Staff, as the proponent of the installation of additional instru-mentation, had a burden in this proceeding to prove the necessity of its version of this long-term recommendation in NUREG-0578. The Staff failed to sustain that burden, and the Licensing Board erred in finding that additional instru-mentation is necessary to provide reasonable assurance that TMI-l can be operated for the long-term without endangering the j health and safety of the public.
l While there is no controversy about the adequacy of the existing instrumentation, procedures and operator training l provided at TMI-l to avoid the onset of inadequate core cooling 12 Even if the Task Force had been more explicit, the i
necessity of its recommendations is open for adjudication in this proceeding.
and to respond to an inadequate core cooling situation if it occurs, it is important to review the evidence on these short-term actions in order to assess the necessity for any further actions. For the most part, the Licensing Board adequately summarized the record on the short-term. actions.
To achieve the goal of assuring adequate core cooling for power operation at TMI-1, the safety analyses which have been performed for the plant have defined the parameters which must be monitored. These important variables -- reactor power; reactor coolant pressure, temperature and flow; and containment pressure -- are measured directly and input to the Reactor Protection System and/or the Engineered Safety Features Actuation System. That is, for power operation the variables appropriate to assure adequate safety have been defined and these parameters are directly measured and input to the protection system. Water level in the core is not a part of the required instrumentation and no incremental protection system action has been identified based on such indication.
There is no known sequence of events which, from a power operation condition, could result in a low water level in the reactor vessel which would not be preceded by a reactor trip from the Reactor Protection System. Keaten et al., ff. Tr.
10,619, at 2, 3. In addition, reactor vessel water level is not an appropriate input to the Emergency Core Cooling System since the corrective action is initiated by a low pressure signal well in advance of core uncovery. Phillips-1, ff. Tr.
10,807, at 5.
Following reactor trip and engineered safeguards actuation, the goal of assuring adequate core cooling is achieved by maintaining subcooled conditions in the reactor coolant system or, in the absence of such conditions, by providing sufficient reactor coolant inventory. Keaten et al.,
ff. Tr. 10,619, at 4. Reactor coolant subcooling is assessed by monitoring system temperature ar.d pressure, and it is these parameters which are utilized most often in the TMI-l plant emergency procedures. Id.; Staff Ex. 1 at C8-14, 15.
The instrumentation available at TMI-1 which indi-cates inadequate core cooling consists of core exit thermocou-ples which indicate coolant superheat associated with excessive fuel cladding temperature, reactor coolant pressure sensors, cold leg and hot leg resistance temperature detectors (RTDs) which provide inputs to compute the margin to coolant satu-ration conditions, subcooling meters which will display the margin to saturation, and reactor coolant pump current which provides indication of increasing coolant quality while the pumps are running. I.D., 1 638. The subcooling meters represent one of the "short-term actions" recommended by NUREG-0578. In addition, instrumentation at TMI-l will be changed prior to restart to connect all 52 of the core exit thermocouples to read out in the control room, and to provide an expanded range (120 F-920 F) for the reactor coolant system hot leg temperature measurement so that the saturation meter can be used to detect the approach to inadequate core cooling outside the normal operating temperature range. Keaten et al.,
ff. Tr. 10,619, at 9; Lic. Ex. 1, S 2.1.1.6; Phillips-1, ff.
Tr. 10,807, at 7; Staff Ex. I at C8-15, 16. See also, I.D.,
1 639, as modified by the Licensing Board's Memorandum and Order Modifying Partial Initial Decision of December 14, 1981, dated January 26, 1982.
Plant procedures at TMI-l have been revised to emphasize the importance of maintaining an adequate saturation margin in the reactor coolant system and to provide guilaace for steps to be taken if the saturation margin is less than the required value. Keaten et al., ff. Tr. 10,619, at 7, 8. The revised procedures define the use of the information available from the core exit thermocouples, reactor coolant system temperatures and the new saturation meter in identifying when inadequate core cooling is approaching and to specify the operator action required to promptly enhance core cooling. Id.
at 9. See I.D., 11 651-653.
For non-mechanistic events at TMI-l beyond the design basis, B&W has developed guidelines for inadequate core cooling which define appropriate actions to prevent significant cladding damage and/or hydrogen generation. These guidelines, which employ instrumentation which will be available at TMI-l prior to restart, are based on recognition of core uncovery and provide guidance to aid in prevention of a situation deterior-ating to an inadequate core cooling condition. To develop these guidelines, a series of calculations were performed to
_ _ _ _ _ _ _ _ _ _ _ _ _ ____________u
l develop a correlation between core exit thermocouple tempera-tures, as a function of pressure, and peak cladding tempera-tures of 1400'F and 1800*F. Using this correlation, two levels of operator actions were identified. Keaten et al., ff. Tr.
10,619, at 10 and Fig. 1; Tr. 10,624-28 (Jones). The 1400*F limit is based on the potential for fuel pin swelling and rupture; the 1800'F limit is based on the metal-water reaction threshold. Staff Ex. 14 at 44.
The NRC Staff has concluded that these guidelines provide the operator with the correct sequence and actions to respond to an inadequate core cooling event, and that they are acceptable as the basis for developing TMI-l plant specific emergency procedures for inadequate core cooling. Staff Ex. 14 at 45. From these guidelines,. Licensee has developed imple-menting emergency procedures for TMI-1. Lic. Exs. 48 (Attachment 3) and 51. The Staff has reviewed these procedures and concluded that they adequately incorporate the B&W guidelines and are modified appropriately to incorporate plant specific information.13 Staff Ex. 14 at 28. In fact, the Staff testified that all of the methods available to terminate 13 Relying upon an NRC Staff proposed finding, the Licensing Board cites to Staff witness Phillips for the proposition that the Staff had not completed its review of the TMI-l procedures.
See I.D., 1 640. As Licensee pointed out in its reply to the Staff's proposed finding, a subsequent supplement to the Staff's safety evaluation reported that the Staff's review is complete and that the TMI-1 procedures are adequate. See Staff Ex. 14 at 28.
inadequate core cooling are included in the TMI-l procedures.
Tr. 16,001-003 (D. Ross).
There is a complete and unchallenged record on the training TMI-l operators receive which relates directly to their ability to recognize and respond appropriately to an approaching inadequate core cooling condition.14 The NRC Staff has found that adequate training has been provided on the causes of, recognition of, and response to inadequate core cooling. Staff Ex. 1 at C8-16, C8-49. The Licensing Board appears to have been satisfied with this training as well. See I.D., 1 641.
The NRC Staff and the Licensing Board appear to agree with Licensee that the existing instrumentation at TMI-l provides an unambiguous, easy-to-interpret indication that an inadequate core cooling condition is being approached, or that it exists. The Licensing Board found as follows, relying in l part upon Staff testimony (Phillips and Jensen):
TMI will have devices to measure the temperature of the coolant at the core exit i and in the hot legs above the core. If the measured exit temperature is below the saturation temperature, the core is covered with water. Jensen, et al., ff. Tr. 7548, at
- 10. Since saturated conditions must occur in the reactor coolant system hot legs before there is danger of inadequate core cooling, the instrumentation available to the opera-tors to detect a loss in the subcooling margin, including the new saturation meter which was not available at the time of the 14 See Licensee's proposed findings of fact 44-49, June 1, 1981.
1
I e
TMI-2 accident, provides information anticipatory to an inadequate core cooling 4
condition. Thus, the instrumentation provides the operator with knowledge that action should be taken to maintain or i reestablish the subcooling margin and that an inadequate core cooling condition is being Ja proached. See Keaten, et al., ff. Tr.
10~,619, at 8 (Keaten); Tr. 10,729-30
- (Keaten); Tr. 10,828-30 (Phillips).
If an accident occurs which nevertheless results in the uncovering of the core, superheated reactor coolant conditions would be indicated by core exit thermocouples and the expanded reactor coolant hot leg tempera-ture instrumentation. Keaten, et al., ff.
Tr. 10,619, at 5 (Jones); Phillips-1, ff. Tr.
10,807, at 4. The Staff's witness Jensen testified that the ranges of this instru-mentation used to monitor core cooling are adequate.for the operator to determine if the coolant in and above the core is subcooled, l saturated or superheated. Jensen, et al.,
ff. Tr. 7548, at 9.
I.D., 11 648, 649.
The Licensing Board then proceeds to address the Staff's sole complaint about the existing instrumentation. The Staff takes the position that while core exit thermocouples can provide an indication of the existence of inadequate core cooling, the measurement of superheated steam temperatures by the core exit thermocouples indicates inadequate core cooling imminent or already present. Staff Ex. 1 at C8-21. The Staff asserts that additional instrumentation is needed for the l period -- which the Staff states may be from 30 minutes to three hours -- between the onset of saturation and the indica-tion of superheated steam temperatures. The Staff asserts that a level meter is needed so that the operator does not " fly blind" for this period. See I.D., 11 649-650, 659.
i
First, it is inaccurate to characterize the operator as " flying blind." It is true that the operator does not know the precise level of the reactor coolant. It is equally true that the operator knows exactly what actions to take in this situation and that there is no additional action the Staff would have the operator take on the basis of level information.
The NRC Staff has not evaluated the possible actions the operator would take based on core level instrumentation.
Phillips-2, ff. Tr. 10,807, at 4. Analyses were performed by both B&W and GPU to determine if any incremental automatic or operator action could be identified on the basis of water level indication. Tr. 10,647-48, 10,658-59, 10,911-15 (Jones); Tr.
10,657, 60 (Keaten). Licensee's position is that no additional or earlier action, beyond that already provided for at TMI-1, can be identified to avoid or respond to an inadequate core cooling condition on the basis of water level indication.
Keaten et al., ff. Tr. 10,619, at 5, 12, 14 and 19; Tr. 10,661 (Jones). The Staff has not refuted Licensee's testimony on this point in any way.
Second, the Staff position raises the question, as the Licensing Board quickly discovered, of the definition of inadequate core cooling. Whether Licensee's instrumentation does not adequately anticipate inadequate core cooling, but merely indicates that such a condition is " imminent or already exists" depends upon how one defines the condition. For, although it is true that superheated steam temperatures indicate core uncovery imminent or already present,15 the I Staff's analysis only holds if core uncovery, by itself, means that the core is being inadequately cooled.
Surprisingly, the Staff seems not to have adopted a definition of inadequate core cooling for the purposes of directing the industry toward additional instrumentation to detect such a condtion. This should have been, it would seem, the first functional criterion established.
A Staff witness at the hearing provided the following definition, which the Licensing Board apparently adopted:
When the two-phase froth level begins to drop below the top of the core, the exposed fuel begins to heat up and will ultimately reach temperatures at which fuel damage occurs.
This is inadequate core cooling.
Phillips-1, ff. Tr. 10,807, at 3. See I.D., 1 645.
This Staff definition should be rejected for two reasons. First, the testimony is unworkably vague as to when "this" occurs -- when the two-phase level begins to drop below the top of the core, when the fuel begins to heat up, or when fuel damage occurs. Second, it does not follow that when the two-phase level drops below the top of the core, fuel damage
, 15 Temperature measurements taken above the core which are at
) the boiling temperature (saturated) or below the boiling l temperature (subcooled) indicate that the core is covered and I
adequately cooled. Temperature measurements taken above the core which are above the boiling temperature (superheated)
I indicate that the core is not covered. Jensen et al., ff. Tr.
7548, at 8, 9.
temperatures necessarily will be reached. Tr. 10,621-22 (Jones).
If the reactor coolant system inventory is reduced and uncovery of the core begins, temperatures in the uncovered region will increase, causing superheating of the steam. In the past, the term " inadequate core cooling" has generally been applied whenever the core is not covered by either liquid coolant or a two-phase mixture, thus resulting in superheated conditions being indicated by the core exit thermocouples.
However, core uncovery by itself does not mean that the core is being inadequately cooled. For example, analyses of design basis, small-break LOCAs result in some core uncovery without any clad damage occurring. Keaten et al., ff. Tr. 10,619, at 6, 7. As a Staff witness testified, the most direct mea-surement of inadequate core cooling is the fuel temperature or surface temperature of the fuel cladding. Phillips-1, ff. Tr.
10,807, at 3, 4. The Commission has established the maximum fuel element cladding temperature by regulation. See 10 C.F.R.
S 50.46(b)(1). Consequently, Licensee proposed a definition which considers inadequate core cooling to exist when the fuel is uncovered to an extent and/or for a period of time such that the limits of 10 C.F.R. c 50.46 would be exceeded. See Keaten et al., ff. Tr. 10,619, at 7. Contrary to the. Licensing Board's inaccurate characterization of this definition, it does not state that inadequate core cooling may not occur prior to fuel damage. See I.D., T 646. That under Licensee's 1
definition inadequate core co(.ing may occur prior to the time of fuel damage is made clear by the benchmarks used by B&W to develop the operator guidelines. See supra at 22.
The Licensing Board erred in adopting an unworkably vague Staff definition of inadequate core cooling, and in ignoring uncha'lenged Licensee testimony to the effect that core uncovery may occur without any fuel clad damage.
i Utilizing the correct definition of inadequate core cooling, ,'
the core exit thermocouples do provide anticipatory indication of that condition. Keaten et al., ff. Tr. 10,619, at 14; Tr.
10,720-21 (Jones); Tr. 10,730 (Keaten). In short, there will already exist at TMI-l at the time of restart sufficient instrumentation, procedures and training to enable unambiguous, easy-to-interpret indication of inadequate core cooling.
B. THE LICENSING BOARD ERRED IN FINDING THAT REACTOR WATER LEVEL INSTRUMENTATION OR ITS EQUIVALENT IS NECESSARY FOR THE /
LONG-TERM OPERATION OF TMI-l l Licensee records, at the outset, that it does not challenge the Licensing Board's adoption of ". . . a standard l that 'necessary' modifications as stated in the Commission's l hearing order are modifications which would produce a substan-tial and additional protection to the public health and safety and which, based upon the record, are reasonable in view of tte technology, resources and risk involved." See I.D., 4 689.
l Rather, Licensee's quarrel is with the application of that standard to the evidence on this issue.
l
l .
It appears that this dispute between Licensee and the Staff' produced what may be described as a "close call" by the Licensing Board. Terming Licensee's litigative position "not l frivolous," the Licensing Board stated that " Licensee could have prevailed on the issue, although on grounds narrower than it argued." I.D., T 669.
Chairman Smith, in his separate statement on water level indication, was more candid. He stated that . . . if this had been a private litigation between t'.e Licensee and the Staff as adversaries without a strong public interest in the result, and without built-in adjudicators' expertise, the Staff might have lost on the issue of the long-term need for water-level indication in.this proceeding." I.D., Y 690.
In part, the Chairman's separate statement records the consistent reluctance of the NRC Staff to join issue on the matter of detecting inadequate core cooling. We have already observed the Staff's delay in filing testimony, and the failure of its initial testimony to take and defend a position. To a great extent, the Staff's testimony is conclusory and unexplained.16 In its proposed findings of fact filed with the i
l -16 For example, Staff witness Phillips testified that the
! Staff had reviewed Licensee's justification for no additional
' instrumentation (the B&W evaluation discussed above) and had found it unacceptable. Mr. Phillips then cites to a Staff lecter which was never offered into evidence. Phillips-1, ff.
Tr. 10,807, at 9. Consequently, the record contains Licensee's entire evaluation and no basis whatsoever for the Staff's rejection of it. For all we know, the Staff simply did not like the conclusion. Nevertheless, the Licensing Board recited this Staff position with apparent approval, even though the (continued next page)
J
i Licensing Board, the Staff largely ignored Licensee's testimony and Licensee's cross-examination of the Staff witnesses.
Moreover, even though Licensee submitted extensive proposed findings on this issue raising many contested sub-issues, including the deficiencies in the Staff's case, the Staff filed no reply findings whatever on this issue. I.D., 1 701 (Judge Smith). This situation caused Chairman Smith to observe that the Staff is largely in default. Id.
It is difficult to ascribe the Staff's approach to this litigation either to a lack of confidence or to arrogance.
Certainly an. undefended Staff position should not be sustained when it is competently refuted by opposing evidence. Issues as vital as this one should not be decided on the basis of a judgment whose rationale cannot be intelligibly articulated.
Vermont Yankee Nuclear Power Corporation (Vermont Yankee Nuclear Power Station), ALAB-229, 8 A.E.C. 425, 440 (1974). It should not matter whether such a judgment or position belongs to the NRC Staff or any other party. See Consumers Power Company (Midland Plant, Units 1 and 2), ALAB-379, 5 N.R.C. 565, 570-71, n.18 (1977).17 In finding for the Staff on the long-term need for reactor water level instrumentation, the Licensing Board (continued)
Licensing Board is totally unaware of the underlying bases, if any. See I.D., 1 655.
17 Cf. Public Service Company of New Hampshire, et al.
(Seabrook Station, Units 1 and 2), CLI-76-17, 4 N.R.C. 451, 462 (1976).
inexplicably ignored Licensee's testimony which refuted each and every basis advanced by the Staff in support of such instrumentation, and misconstrued one of the key reasons why Licensee opposes such instrumentation.
The Staff advanced several reasons why, i.n its view, operational safety at PWRs would be enhanced with reactor vessel water level indication. Mr. Phillips testified that the saturation meter, while providing a basis for initial actions, does not distinguish between anomalous transients which can drain the pressurizer and cause primary loop saturation due to cooling and shrinkage of primary coolant versus loss of coolant inventory which could lead to inadequate core cooling if it continues. Phillips-2, ff. 10,807, at 2. The Licensing Board appears to have adopted these reasons uncritically. See I.D.,
11 656-658.
Licensee testified, however, that the operator does not need to make an instant diagnosis of these alternative transients and that he could not do it with level information.
Whether it is an overcooling event or a LOCA, the operators' job is to restore primary system inventory and pressure with HPI. Diagnosis of an overcooling event is not required for the immediate action steps -- which are identical for a small-break LOCA and an overcooling event. Follow-up procedures and instrumentation are adequate to timely diagnose and respond i
safely to the particular transient.18 Tr. 10,632-36 (Jones, 18 TMI-l emergency procedures provide specific guidance on diagnosing the secondary side symptoms characteristic of an (continued next page)
M. Ross); Tr. 10,641 (Keaten); Tr. 10,677 (Jones). In any case, for a very small-break LOCA the primary system will stay solid for a period of five to ten minutes, depending on the size of the break; whereas, a severe overcooling event can result in a steam bubble within the reactor vessel. head region.
So that it is simplistic to imply that vessel level is a reliable diagnostic tool for distinguishing these events. Tr.
10,636-37 (Jones). Consequently, vessel level indication may in fact be more ambiguous information in the early rtages of a transient than the information operators at TMI-1 already have.
Tr. 10,664-66 (Jones).
Mr. Phillips also testified that water level indica-tion would provide indication of the effectiveness of HPI in recovering the system. Phillips-2, ff. Tr. 10,807, at 3.
However, because analyses show that the rate of recovery, including uncovery of the core for some breaks and locations, is transient-dependent, the operator will not be able to assess i definitively from water level whether HPI is effective. Tr.
l 10,687-88 (Jones). Chairman Smith, in his separate statement, observes that Mr. Jones has rebutted the Staff's testimony on l
this point and that Mr. Jones' testimony remains unrefuted.
I.D., 1 696. Yet, the Licensing Board adopted the Staff's position without further explanation. See I.D., T 660.
(continued) overcooling event, as distinguished from a small-break LOCA.
Tr. 10,643 (Keaten). See, e.g., Lic. Ex. 48 at 3.
The Staff testified that vessel level information is important and possibly essential to proper emergency procedures relating to use of the reactor vessel head vent, which is another NRC Staff requirement. Phillips-2, ff. Tr. 10,807, at 4, 5. Licensee reported, however, that the guidelines under development by B&W for vent use do not rely on water level indication. Tr. 10,692 (Jones). Again, Chairman Smith acknowledges the apparent lack of merit to this Staff testi-mony, I.D., 1 699, while the Licensing Board decision adopts the Staff position uncritically. I.D., 1 661.
The Licensing Board relied upon Staff testimony that level indication would also provide evidence that the core is covered during recovery from a TMI-2 type flow blockage condition, even though superheat may persist at the core exit thermocouples. I.D., 1 660. Again, Licensee rebutted this testimony.19 The TMI-2 accident showed that core exit thermo-couples did provide indication when the core had been recovered. Further, the operator's actions do not change on the basis of this information -- the operator must continue to refill the primary system until the subcooling margin is restored. Tr. 10,772-73 (Jones).
Staff witnesses Phillips and Ross each cited a l natural circulation cooldown event at St. Lucie-1 on June 11, 19 Chairman Smith correctly points out, however, that this point was overlooked in Licensee's proposed and reply findings.
I.D., 1 701 n.79.
l l
i l i
1 l 1980, as evidence that vessel level indication is desirable.
Phillips-2, ff. Tr. 10,807, at 4; Ross, ff. Tr. 15,915, at 3.
While the Staff provided virtually no description of the event, l Licensee witness Jones reported that this event at a Combustion l Engineering plant involved a loss of component cooling water to the reactor coolant pumps, which were then tripped. During the cooldown, a void was formed in the upper head of the vessel.
This was indicated to the operators because the level swing occurring in the pressurizer was rather large and could not be explained by the fact that the injection location was being changed from the cold legs to the pressurizer as a pressurizer spray. Tr. 10,688-89 (Jones). The Staff reports that the operators initially did not recognize the steam bubble and that unsafe operator action could have been taken; whereas vessel level information would have indicated the void formation in
- the upper head.20 Phillips-2, ff. Tr. 10,807, at 4; Ross, ff.
1 Tr. 15,915, at 3. It it not clear, however, what unsafe operator actions might have been taken. Tr. 10,690-91 (Jones).
It is undisputed, though, that the operators took the correct actions to control the plant during the St. Lucie event 20 Dr. Ross described this as "an extended period'of operator confusion"; whereas Mr. Imbro, the author of the Staff report on the St. Lucie event spoke in terms of " initial puzzlement."
Compare Ross, ff. Tr. 15,915, at 3, with Tr. 15,965-66 (D.
Ross). Dr. Ross had no more information on the event than was l available to Mr. Imbro. Tr. 15,966 (D. Ross). There is no basis, then, for the Licensing Board's conclusion that the operators were slow to diagnose this event. See I.D., 5 664.
1
t cited by the Staff. Tr. 15,966 (D. Ross). Indeed, Licensee cites the same event in support of its position that level information may mislead operators. If vessel level instru-mentation had been available at St. Lucie the operators may well have misdiagnosed the event as a small-break LOCA. Tr.
10,637 (Jones). We also note that the list of recommendations in the written Staff report on the event includes no mention of need for level indication. Tr. 10,691-92 (Jones).
Staf f witness Ross also cited a loss of coolant event of February 11, 1981, during cold shutdown at Sequoyah-1, a Westinghouse reactor. Ross, ff. Tr. 15,915, at 4; Tr. 15,960 (D. Ross). In this event, the operators lost pressurizer level in two minutes, and reestablished it in ten minutes. Tr.
15,962 (D. Ross); Ross, ff. Tr. 15,915, at 4. This restoration of pressurizer level, which from the standpoint of core cooling essentially means that the primary system is refilled, took place as a result of operator actions taken two or three minutes into the event. Tr. 15,962-64 (D. Ross). Conse-(
quently, it appears that the operators took very quick and appropriate corrective action without vessel level information.
Dr. Ross did not identify any additional action which the Sequoyah operators might have taken on the basis of level indication.
Again, while Chairman Smith at least acknowledges the existence of this responsive evidence by Licensee, I.D.,
11 697, 698, the Licensing Board decision simply recites the
- Staff's conclusory testimony. See I.D., 11 663, 664.
I l
Finally, the Staff throws in an argument, repeated by the Licensing Board, that while existing instrumentation may be sufficient to respond to TMI type accidents, it may not be sufficient to respond to other unidentifiable accidents. Tr.
10,892 (Phillips). Licensee does not understand this testimony and, while the Licensing Board discussed it, apparent confusion exists there as well:
Just what the Staff had in mind with reference to anomalous transients was not made clear except that they were outside the scope of proposed procedures and training.
Licensee's position is that the present procedures are adequate for any small-break LOCA. Keaten, et al., ff. Tr. 10,619, at 5, 12, 14 and 19. -- --
I.D., 1 662. Chairman Smith, on the other hand, characterized this as a need for level information for " anomalous, uniden-tified purposes" for " unidentified episodes . . . within a TMI-2 type transient." I.D., 11 704, 705. It appears bla-tantly inconsistent to observe that this undefined regime of events is outside the scope of present procedures and training, yet within a TMI-2 type transient.
This reasoning nearly condemns itself as a basis for making hardware modifications at a nuclear power plant. Beyond the solid evidence Licensee presented to rebut the bases for the Staff's position, Licensee advanced two arguments against l reactor water level instrumentation. One argument was under-stood by the Licensing Board and apparently rejected. The l second argument was not understood.
l The Staff recognized that it has not identified j differences in operator actions if water level were available j versus those actions now required by existing guidelines for l inadequate core cooling. I.D., 1 695 (separate statement of Judge Smith). This alone causes Licensee concern about a decision now that a reactor water level instrumentation system should be installed. The guidelines prepared by a distin-guished team of experts assembled by Licensee to perform a human factors review of the TMI-1 control room state, at the very outset of the operational guidelines:
The control room operators who man the main console should be provided with appropriate controls and displays to perform a set of defined functions. Controls and displays, including annunciators, which are not needed to perform those defined functions tend to divert the control room operators' attent'on and should not normally be provided to them.
It should be an objective to move out or keep out of the control room itself those person-nel, controls, and displays which are not related directly to the defined functions.
Lic. Ex. 23, Appendix A at 2. Consequently, Licensee's witnesses testified'that they are reluctant to install instru-mentation for which there is not an identifiable use. Tr.
10,644-45; 10,703 (Keaten); Tr. 10,706 (M. Ross). See I.D.,
l 1 654.
Licensee's second objection reflected its concern that reactor water level indication could mislead the operator into premature throttling of high pressure injection -- one of the key contributors to the TMI-2 accident. The Licensing Board apparently misunderstood this as a concern that the operators could not be trained to interpret the information properly. See I.D., 11 654, 670, 694. This was not Licensee's position at all. The concern is not with the operator's ability to comprehend training, but,with the engineer's ability to relate reactor water level information to appropriate guidance for operator action. This was explained in Licensee's proposed finding 65, June 1, 1981.
The HPI system provides an integrated makeup flow to the primary system such that the time delay to water level in the core may not intuitively reflect the fact that a normal recovery is in progress which will restore inventory. Tr.
10,649-50 (Jones). There is a very wide spectrum of events which the operator must be prepared to meet -- from a very small-break LOCA to a very large break, and at various loca-tions in the primary system. The behavior of the actual liquid level or the two-phase level varies enormously for these different transients, so that the coolant level and the rate-of-change in the coolant level cannot be categorized simply into a regime which is safe and one which is not. What may be nor2a1 and expected behavior for one break would be abnormal for another. Tr. 10,661-62 (Keaten). This is illustrated by a figure from Licensee's small-break LOCA cnalyses (ff. Tr. 10,663) which displays two-phase mixture height in the core over time for a variety of break sizes. For several breaks the two-phase mixture drops below the top of the l .
active core for a short period of time. Yet, the analysis shows that ECCS is working properly and that the operator should take no corrective action, but continue to rely upon the ECCS. Tr. 10,662-64; 10,674-76 (Jones); Tr. 10,682; 10,700-01 (Keaten). The concern is that because of inadequate analytical guidance the operator will prematurely throttle HPI -- at a time when the analyses predict water level should be high, but also predict that it will drop further into the transient -- so that inventory cannot be recovered; or that because of inadequate analytical guidance the operator inappropriately takes drastic action upon observing core uncovery, when level is predicted to recover with just normal HPI flow. Tr.
10,651-52 (Jones).
The Licensing Board ignored the fact that the Staff conceded this drawback. The Staff announced for the first time at the hearing, during the oral testimony of Mr. Phillips, that the Staff does not envision providing vessel level information directly to plant operators. The Staff apparently recognizes that level information can, for some periods during a tran-sient, provide misleading information. Consequently, it is proposed that the level information be fed into some sort of
! data processing equipment where it will be integrated somehow with other instrumentation to " weed out" false signals. Tr.
10,810-13, 10,818-23 (Phillips). Mr. Phillips testified that it would be unacceptable to base any operator action on level indication alone.21 Tr. 10,849-50 (Phillips). Licensee has 21 If the data processing system fails, however, the operator would have to rely upon the hard-wired backup instrumentation '
(continued next page) '
described what it considers to be the extreme difficulty of correlating primary coolant inventory versus time with the safety analyses performed for the plant. Tr. 10,684-85 l (Jones). The Licensing Board was not told by the Staff what information, whether on a CRT or other device, ultimately will be displayed to the operator. It is uncertain what impact this aspect of the Staff's testimony has on its other arguments supporting the utility of water level information.
Staff witness Phillips testified that: (a) there is still the possibility that the Staff ultimately will conclude that no system proposed to measure water level is acceptable, Tr. 10,833; (b) before the Staff determines whether any system is acceptable it will review the potential use of the informa-tion provided and weigh it against any detriments, Tr.
10,861-62; (c) in order to be found acceptable a proposed system will have to be found to provide an overall enhancement to safety, and the Staff will not make such a determination until the systems are installed, the operating methods have been identified, the calibration and test data is available, and the Staff is certain that these systems are indeed a plus to safety and will not lead to unsafe actions. Tr. 10,811, 10,864, 10,909. See I.D., 1 671.
One of the most important points Licensee asserts on this appeal is to challenge the Staff's decision-making process (continued) and to diagnose the plant condition with possibly anomalous vessel level information. Tr. 10,861-62 (Phillips).
I on this issue. The Staff appears to have reached a stubbornly held, but thinly supported position in advance of a deliberate engineering investigation. The Staff may ultimately be vindicated, and Licensee does not foreclose the possibility that in the future additional instrumentation may be developed which provides an enhancement to the safe operation of TMI-1.
Even if the Licensing Board disagreed with most of Licensee's evidence, it should have found that there is no basis now to decide that reactor water level instrumentation is necessary at TMI-1. Under the standard adopted by the Licensing Board for "necessary" actions, it has not yet been shown that additional instrumentation would produce "a substantial and additional protection to the public health and safety."
A reasonable application of the normal engineering method would sequence the selection of functional criteria, the identification of the alternatives and then the optimal choice for fulfilling the criteria, prior to the detailed engineering to apply the alternative, procurement and installation. Tr.
15,957 (D. Ross). The step of identifying the alternatives could include consideration of: reliability, ease of retrofit; in situ verification of calibration; probability of accident i
l survival; lifetime or long-term survival; accuracy; additional penetrations; simplicity; versatility; performance history; and cost. Tr. 15,958-59 (D. Ross).
Yet the Staff schedule presented at the hearing would have required installation, and obviously the detailed engineering and procurement, to precede Staff consideration of these factors. In fact, in order to have met the Staff's schedule for installation, licensees would have begun install-ing systems in their plants even before the issuance of a generic SER approving even the concept of the system. Tr.
10,838-39 (Phillips); Tr. 15,944 (D. Ross). There is nothing to prevent the Staff from ultimately disapproving such a system or failing to find a use for it.
Chairman Smith, in his separate statement, wrote that
. . . the facts do not seem to strongly support the Staff as to . . . (this] particular issue in . . . [the] proceeding."
I.D., 1 702. Licensee would go further and argue that the record weighs against the Staff. This was the wrong issue for the Licensing Board to have exercised its own expertise to save a drowning Staff case. The words of the Licensing Board's own decision make it plain that Licensee here is being directed to install an unknown system to provide unk,own information to the operators for an unknown purpose.
The test " laid down by the courts" and followed by the Atomic Safety and Licensing Appeal Board allows the Appeal l
Board to reject or modify the Licensing Board's findings "if, after giving its decision the probative force it intrinsically commands," the Appeal Board is convinced that the record warrants a different result. Niagara Mohawk Power Corporation (Nine Mile Point Nuclear Station, Unit 2), ALAB-264, 1 N.R.C.
l 347, 357 (1975); accord, Northern Indiana Public Service 1
l Company (Bailly Generating Station, Nuclear 1), A LAB-3 0 3, 2 N.R.C. 858, 867 (1975), and cases cited therein. Where the record will fairly sustain a result deemed " preferable" by the agency to that selected by the Licensing Board the agency may substitute its judgment for that of the lower board. Tennessee Valley Authority (Hartsville Nuclear Plant, Units lA, 2A, 1B and 2B), ALAB-367, 5 N.R.C. 92, 94 n.4 (1977), citing Duke Power Company (Catawba Nuclear Station, Units 1 and 2),
ALAB-355, 4 N.R.C. 397, 402-05 (1976).
The record here clearly calls for reversal by the Appeal Board.
C. THE LICENSING BOARD'S DECISION SHOULD NOT BE READ TO REQUIRE ANY PARTICULAR DESIGN CRITERIA FOR ANY REACTOR COOLANT LEVEL OR OTHER INSTRUMENTATION INSTALLED AT TMI-l The Licensing Board's decision is not free from doubt as to what requirement is being recommended for imposition on Licensee. Early in its decision on the detection of inadequate core cooling, the Licensing Board states:
In the discussion that follows we will use l the term " coolant level instrumentation" in a l broad sense to denote a system that measures I
coolant level, cr olant inventory, coolant density, or some parameter closely related to the foregoing.
l I.D., 1 637. Licensee would agree that the record does not l
l warrant a more finely focused definition of the potential types of instrumentation under consideration.
Later, however, the Licensing Board uses more i '
proscriptive language which, literally construed, might be
I l .
viewed to establish functional design criteria. The Licensing Board found that . . . a meter capable of measuring reactor coolant inventory from 100 percent to zero would be a useful and valuable operating adjunct and is needed in the long term,"
I.D., 1 665, and that "[a] meter to measure water 1,evel in the core or its equivalent is required in the long term." I.D.,
1 673.
Licensee doubts that the Licensing Board intended here to limit the type of additional instrumentation system the Staff could accept. Speaking to schedule, the Licensing Board stated that "[w]e leave it to the Staff and the Commission to require the installation at TMI-l consistent with the treatment of other similar reactors." I.D., 1 673. There is no basis in the Licensing Board's decision, or in the record, for the imposition of any particular design requirements for TMI-1.
Dr. Ross testified, for the Staff, that:
We do not wish to preclude the consideration of alternate systems which may be selected on the basis of superior functional performance, installation simplicity, safety concerns l associated with the installation, lower cost,
! or any other factors appropriate to a sound engineering approach to this problem.
Further, we recognize that the identified level measurement systems may not be physi-cally adaptable to some reactors without design modifications to the reactor vessel or internals which might create unacceptable safety conditions, though no such situation has been identified to the staff to date.
Ross, ff. Tr. 15,915, at 7, 8.
Consequently, if the Appeal Board rejects Licensee's first two arguments in support of its Exception No. 1, Licensee respectfully requests that the Appeal Board nevertheless supplecent the Licensing Board's decision by clarifying that the Staff has not been limited in terms of the design criteria to be employed for any additional instrumentation at TMI-1 to detect inadequate core cooling. .
EXCEPTION NO. 3 The decision by the Licensing Board that certain of the functions of the Emergency Support Director, which initially are assumed by the onsite Emergency Director, be trans-ferred within one hour after declaration of a site emergency to an individual located in the near-site Emergency Operations Facility is not supported by reliable, substantial and probative evidence, is based upon an erron-eous legal analysis of the regulatory requirements for plant staffing during an emergency, and inappropriately disregards internal management decisions properly vested with Licensee. See PID 11 1374-96, 2010(a).
Statement of the Case The issue addressed by Exception No. 3 raises
. significant and, in view of the manner in which it was resolved l
by the Licensing Board below, potentially troublesome questions over the division of responsibility between the NRC and its licensees as to the detailed manner in which licensee emergency response organizations will be staffed and organized. No one doubts the need to identify all components of a licensee's emergency response organization, to specify the responsi-l bilities and functions of each emergency response position, and
l -
to staff that organization with an adequate number of well-trained personnel. As part of its efforts to upgrade emergency planning around TMI, Licensee developed an emergency response organization which it rightfully believes is unparal-leled in the industry, and fully staffed that organization with well-trained personnel.22 Despite these efforts, the Licensing Board below, urged by the NRC Staff and the Commonwealth of Pennsylvania, directed Licensee to modify its emergency response organization so that certain functions, initially performed by the onsite Emergency Director, will be transferred offsite within one hour after declaration of a site emergency to an individual located in the near-site Emergency Operations Facility (" EOF"). As explained below, the Licensing Board's order that Licensee so modify its emergency response organization constitutes an unwarranted and inappropriate intrusion into Licensee's
[
well-intentioned internal management decisions, and is without 22 Licensee's emergency response organization is described in its prepared, written testimony relating to onsite emergency l preparedness (see Rogan et al., ff. Tr. 13,756) and in the l
TMI-l Emergency Plan (see Lic. Ex. 30). While Exception No. 3 deals with the organization and staffing of Licensee's offsite emergency support organization, to understand the concept of operations underlying that organization it is necessary to consider the organization and staffing of Licensee's onsite emergency response organization and the manner in which the offsite organization supplements and supports the onsite l organization. As a convenience to the Appeal Board, Licensee has included as an Appendix excerpts from its prepared written testimony that describe its onsite and offsite emergency organizations, the emergency response facilities to which they report, and the communications systems linking them together. .
(
legal or factual support. Accordingly, this Appeal Board should act to correct the Licensing Board's error.
No Commission regulation identifies the need for an Emergency Support Director, let alone directs that the Emergency Support Director be stationed at the EOF.within one hour after declaration of a site emergency. See Tr. 22,930 (Chesnut). As the Licensing Board correctly observed, three regulations do, however, bear indirectly on this question: (a) the requirement of 10 C.F.R. S 50.47(b)(8) that " adequate emergency facilities and equipment to support the emergency response are provided and maintained", (b) the requirement of 10 C.F.R. S 50.47(b)(2) that " timely augmentation of response capabilities is available, and the interface among various onsite response activities and offsite support and response activities are specified", and (c) the requirement of 10 C.F.R. Part 50, Appendix E, S IV.E.8 that "[a]dequate provisions shall be made and described for emergency facilities and equipment, including * *
- a licensee near-site emergency operations facility from which effective direction can be given and i
effective control can be exercised during an emergency." See I.D., 11 1377-79. '
i 23 With respect to the cited provision of 10 C.F.R. Part 50, Appendix E, the Licensing Board concludes that it is
" difficult" to interpret "during an emergency" as precluding the first four hours of an emergency. I.D., f 1390. If the Licensing Board meant to hold that this provision requires an Emergency Support Director at the EOF sooner than four hours after the declaration of a site emergency, it is wrong as a matter of law. Even the NRC Staff's guidance in NUREG-0696 interprets the phrase "during an emergency" to mean not during an unusualnext (continued event or an alert (the lowest two classes of page)
_47_
To find guidance relating to the Emergency Support Director one must turn to two NRC Staff documents: NUREG-0654 (Staff Ex. 7) and NUREG-0696 (Staff Ex. 8). Even there, the l guidance provided by the NRC Staff is less than crystal clear.
NUREG-0696 provides guidance on the functions; location, structure, and habitability; staffing and training; size; l radiological monitoring; communications; instrumentation, data system equipment, and power supplies; technical data and data system; and records availability and management for the EOF.
Staff Ex. 8 at 16-24. NUREG-0696 recommends that the EOF be activated for site and general emergencies (id. at 5), and further recommends that "[u]pon EOF activation, designated personnel shall report directly to the EOF to achieve full functional operation within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />" (id. at 19). With respect to the number and type of personnel to be stationed at the EOF, NUREG-0696 makes no specific recommendations. Rather, it provides (id.):
A senior management person designated by the licensee shall be in charge of all licensee activities in the EOF. The EOF staff will include personnel to manage the licensee onsite and offsite radiological monitoring, to perform radiological evaluations, and to interface with offsite officials. The EOF staff assignments shall be part of the licensee's emergency plan.
The specific number and type of personnel (continued) emergency) and not during the first hour after declaration of a site emergency. Whether "during an emergency" means one hour or four hours after declaration of a site emergency simply is not addressed by the regulations.
t 1
assigned to the EOF may vary accordir.q to i the emergency class. The staffing for each '
emergency class shall be fully described in the licensee's emergency plan.
l NUREG-0654 adds very little to this guidance. It recommends that an EOF be established in accordance with the guidance of NUREG-0696 (Staff Ex. 7 at 52), that the emergency response facilities, including the EOF, be activated and staffed in a timely manner (id.), and, as part of the so-called Table B-1 guidance, that a senior manager be stationed at the EOF within one hour for the purposes of " radiological accident assessment and support of operational accident assessment" (id.
at 37).
In response to this guidance, Licensee has committed to activating its EOF within one hour after declaration of a site emergency. This will be accomplished by making all EOF communications and data links operational within one hour and by staffing the EOF with at least six key personnel: represen-tatives from the Emergency Support Staff, Emergency Prepared-i
) ness Department, Environmental Command Center, Technical Functions Group, Communications Department, and a primary communicator. Licensee will station its Emergency Support Director at the EOF within four hours after declaration of a site emergency. During the three-hour span between activation 1
of the EOF and arrival of the Emergency Support Director, the Emergency Director in the control room will retain decision-making authority and will function as the senior corporate i
management spokesman for Licensee. See I.D., 11 1381-82; Lic. Ex. 58.
The essential difference between Licensee's emergency response organization and that sought by the NRC Staff is, as voiced by NRC Staff counsel, whether the Emergency, Support Director is physically stationed in the EOF or the control room during the first four hours after declaration of a site emergency.24 Tr. 22,984 (Tourtellotte); I.D., T 1384. With respect to this difference, all parties have identified the function of making protective action recommendations to the state as the crucial issue. I.D., Y 1385. Licensee correctly believes that the emergency response organization it has developed, trained and drilled is better able to make such' protective action recommendations then would be the case if this function was transferred out of the control room to the EOF during the early hours of an emergency.
Questions Presented
- 1. Did the Licensing Board err by overruling, on the basis of generic NRC Staff guidance documents, Licensee's j
24 NRC Staff counsel framed this difference as the location of the Emergency Support Director and the time of arrival of that person. Tr. 22,984 (Tourte11otte). In fact, the time of arrival is not really a separate issue. The record evidence indicates that if Licensee were to physically move its Emergency Director from the control room to the EOF, and designate its Operations Coordinator as the senior official in the control room, Licensee's emergency response organization would be acceptable to the NRC Staff. See Tr. 22,940 (Chesnut). Under this scenario no one has been added to the emergency response organization. The only change has been to 3 move one individual from the control room to the EOF. Thus, in Licensee's view, the entire dispute with the NRC Staff revolves around the location of this one individual.
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i i
internal management decision that the emergency response at TMI is enhanced by having protective action recommendations made from the control room during the early hours of an emergency?
2.
Did the Licensing Board err by ignoring Licensee's evidence that the emergency response at.TMI is enhanced by having protective action recommendations made from the control room during the early hours of an emergency?
- 3. Did the Licensing Board improperly intrude into the internal management responsibilities of Licensee by directing that protective action recommendations be made offsite from the EOF during the early hours of an emergency?
Argument A. THE LICENSING BOARD ERRED BY PLACING UNDUE RELIANCE ON GENERIC GUIDANCE DOCUMENTS THAT DO NOT CONSIDER OR EVALUATE THE ADEQUACY OF LICENSEE 'S EMERGENCY RESPONSE STAFFING AND ORGANIZATION AT TMI-l In resolving this issue, the Licensing Board cor-rectly observes that "[t]he burden of proof is clearly on Licensee to demonstrate on the basis of firm record evidence the workability and adequacy of its proposed compliance" with the applicable emergency planning regulations. I.D., 1 1391.
l Where the Licensing Board makes a fatal error is its conclusion l
that "NUREG-0654, NUREG-0696, Rev. 1, and the emergency l
planning rules, taken together, support a finding that the EOF should be fully staffed and operable within about one hour of l
declaration of a site emergency." I.D., T 1392. Relying on this " evidence" [i.e., the two NUREG's and the rule], the l
Licensing Board improperly holds: "In light of the Staff and l Commonwealth having come forward with this evidence, including the guidance of NUREG-0654, and the fact that Licensee has the burden of proof, the Board finds that the Licensee has not demonstrated an alternative regarding functions performed by the Emergency Support Director." Id.
As previously noted, the Commission's emergency planning regulations do not address the need for an Emergency Support Director and certainly do not require that he report to the EOF within one hour after declaration of a site emergency.
With respect to the two NUREG documents cited by the Licensing Board, the reliance placed upon them by the Licensing Board is totally unwarranted.
On their face, the NUREG's carefully indicate that the use made of them by the Licensing Board is wrong. Mr.
Eisenhut's cover letter to NUREG-0696 (identified as Generic l
l Letter No. 17) states in relevant part (Staff Ex. 8):
The Commission has approved NUREG-0696 and noted that this document provides general guidance only, is an acceptable way to meet the NRC rules and regulations, and that compliance with NUREG-0696 is not a requirement. [ Emphasis added.]
See also Tr. 15430 (Grimes). And, with respect to NUREG-0654, the Licensing Board itself, in its " Memorandum and Order on Effect of New Emergency Planning Regulations", dated March 23, 1981, has held (slip op. at 5, 6):
The board agrees that the guidance of this joint NRC-FEMA report [NUREG-0654), including the suggested implementation schedule, may be utilized as a standard by which to measure reasonable progress in this proceeding. How-
! ever, we are not prepared to cloak NUREG-0654 with the mantle of a regulation. Neither, apparently, is the staff, although we :: ave received inconsistent staff advice on this
- point. [ Citations omitted.]
l * * * * * * *
- l l
t At this point, it is our view that parties i are not precluded from attempting to show
, that compliance with NUREG-0654 is not necessary or sufficnent. We will make our
! findings based upon the factual record.
j [ Emphasis added.]
4 Despite the Licensing Board's promise to resolve
! disputes involving NUREG-0654 guidance on the basis of a factual record, it is obvious from the Initial Decision that the Licensing Board did not do this, but merely endorsed the guidance documents without any consideration of whether the documents present a reasoned basis for the recommendations set forth or whether the documents considered and addressed the site specific situation at TMI.
The record indicates that, while NUREG-0654, Table 1
B-1, recommends stationing an Emergency Support Director at the.
EOF within one hour, it does not describe any reasons support-
, ing that recommendation. Tr. 22,931 (Chesnut). Similarly, l
while NUREG-0696 has a section on EOF staffing (see Staff Ex. 8 at 19), it does not explain the basis for any position set forth in the document. These shortcomings are compounded by-I
I the fact that the authors of NUREG-0654 and NUREG-0696 had no l
knowledge of the TMI site specific emergency plan, including the staffing levels provided for by Licensee or the concept of operation governing Licensee's emergency response. Tr.
22,931-32 (Chesnut). Nor has the NRC Staff undertaken to publish any study evaluating where a licensee should station its official responsible for making protective action recom-mendations to the state. Tr. 22,933 (Chesnut).
In the face of this evidence, it was wrong for the Licensing Board to rely on NUREG-0654 and -0696 as some type _of
" evidence" supporting its conclusion that the function of making protective action recommendations should be transferred offsite during the early hours of an emergency. We next describe Licensee's evidence, which was either misunderstood or ignored by the Licensing Board, in support of Licensee's decision to have the onsite Emergency Director retain authority for making protective action recommendations to the state during the early hours of an emergency.
B. THE LICENSING BOARD EITHER MISUNDERSTOOD OR IGNORED LICENSEE'S REASONS FOR RETAINING DECISIONMAKING AUTHORITY ONSITE WITH THE EMERGENCY DIRECTOR DURING THE EARLY HOURS OF AN EMERGENCY The Licensing Board correctly summarizes Licensee's s.
observation that two conflicting lessons were learned from the TMI-2 accident: first, thought should be given to stationing the person making protective action recommendations outside the
l control room so as to minimize the number of people and functions performed within the control room, and second, thought should be given to stationing the person making protective action recommendations inside the control room, at least during the early hours of an emergency, so as to improve the accuracy of information and to minimize the likelihood of confusion about plant operations or radioactive releases.25 I.D., 1 1385. The Licensing Board also correctly summarizes the NRC Staff position as the need to transfer the radiological assessment and protective action recommendation function to a senior manager in the EOF at an early point during the emergency so as to relieve the in-plant Emergency Director of these responsibilities. I.D., 1 1387.
What the Licensing Board either did not understand 26 or chose to ignore was Licensee's retort that the NRC Staff had 25 During both the accident at TMI-2 and a subsequent incident at Crystal River there was confusion and l misunderstandings about important information transmitted l
offsite during the early hours following the emergency. Tr.
15,481 (Grimes).
26 Licensee believes the Licensing Board may have misunderstood its position since at a crucial point in its decision the Licensing Board dismisses Licensee's argument by mischaracterizing Licensee's proposed findings. In this regard, the Licensing Board concluded as follows (I.D.,
1 1392):
While Licensee would have us find that absence of the Director at the EOF is compensated for by Licensee's large offsite response organization (Licensee proposed finding 1 46), we are not reassured by this argument, and we agree with the Commonwealth that a large and complex off-site response argues in favor of the need for a single coordinator in the EOF.
[ Emphasis added.]
(continued next page) l l
~
l I
considered only the first of the two lossons locened from the !
l TMI-2 accident and that Licensee's emergency response organiza-tion had considered both lessons and had sought to balance the i conflicting lessons in a reasonable manner.
To ensure that the Emergency Director located in the control room is not overburdened with too many responsibil-ities, Licensee has provided him with three primary lieutenants (continued)
In fact, what Licensee urged the Licensing Board to find is quite different. Licensee Proposed Finding 46 states in relevant part:
Based on Licensee's large onsite emergency response organization (see 11 37-39, supra), the additional offsite staffing at the Alternate EOF, the ,
Environmental Assessment Command Center, and the Parsippany Technical Functions Center (see Rogan, et al., ff. Tr. 13756, at 33-36; Lic. Ex. 30, S 4.5.1.4, at pp. 5-16 to 5-21 and Figure 13), and the functions to be performed by the offsite emergency support organization (see Rogan, et al., ff. Tr.
13756, at 38-39), the Board finds that the staffing at the EOF (but for the question of the Emergency Support Director) is adequate. [ Emphasis added.]
See also Licensee Reply Finding 63.
As to the Commonwealth's concern that a large offsite organization argues in favor of the need for a single l :oordinator at the EOF, Licensee submits that this observation does not go to whether Licensee's senior management official, vested with authority to make protective action recommendations to the state, should be located at the EOF or in the control room. Management of the offsite organization has been considered and centralized by having each of the important segments of the offsite organization send its representative to the EOF within one hour after declaration of a site emergency.
See Lic. Ex. 58. This will assure proper coordination of Licensee's offsite support organization without the need to transfer the single corporate spokesperson from the control room to the EOF. .
. 1 in the areas of plant operations (Operations Coordinator),
technical and engineering support (Technical Support Center Coordinator), and radiological assessment (Radiological Assessment Coordinator). Reporting to the operations l Coordinator in the area of plant operations is the normal shift operating crew (responsible for actual plant control) and the operations Support Center Coordinator (responsible for in-plant maintenance and repair, in-plant radiological surveys and controls, and search and rescue missions). In this manner the Emargency Director; as the senior corporate manager, can exercise oversight in all important emergency response areas (including making protective action recommendations) without getting drawn into the minute-by-minute response in any single area. Tr. 23,077-78, 23,091-92 (Rogan); Lic. Ex. 30, at S 4.5.1.3.2, pp. 5-9 to 5-16 and Figure 12; see also Appendix.
There is no record evidence inconsistent or contrary to this conclusion. Indeed, the NRC Staff testimony tends to confirm Licensee's views as to the capabilities of its Emergency Director.
Licensee's Emergency Plan calls for 20 people on-shift at all times. By comparison, the NRC Staff guidance in NUREG-0654 specifies a minimum shift complement of only 10, and the ability to augment that shift staffing with 11 addi-tional people 30 minutes after declaration of an emergency.
l See Tr. 22,290 (Chesnut); compare Staff Ex. 7 at Table B-1.
I Thus, during plant operation Licensee has twice the minimum l
. i l
staffing acceptable to the NRC Staff; indeed, Licensee's 1
on-shift staffing almost complies with what is recommended 30 minutes after declaration of an emergency. This level of ,
on-shift staffing is one on the largest, if not the largest, encountered by the NRC Staff at any nuclear power plant. See '
I Tr. 15434 (Chesnut). As the NRC Staff's emergency plan reviewer observed, Licensee's on-shift' emergency organization is the "best I have seen. I have not seen any plan which has the level of expertise that the Licensee is planning on using."
Tr. 22,291-92 (Chesnut).27 As a result of this high level of staffing, .
Licensee's organization has special emergency response capabil-ities beyond those specified by the NRC Staff. This includes additional personnel to make the necessary notifications to offsite agencies, to monitor radiation releases and calculate offsite doses, and to conduct prompt offsite radiological surveys. Tr. 15,436 (Chesnut). In addition, since Licensee maintains a three-section duty roster for all emergency response organizatici positions, there is an increased likeli-hood -hat Licensee will have available at the time of any emergency a complete complement of fully trained personnel to fill all positions. Tr. 15,436-39 (Chesr d).
Moreover, in developing iri hser.,3ncy Plan, including that aspect of the plan which vests protuwelve action 27 Despite this relatively large staffing, it was the NRC Staff's view that Licensee's Emergency Plan does not station too many people in the control room. Tr. 15,472-73 (Chesnut).
recommendations with the Emergency Director during the early hours of an emergency, Licensee had the benefit of more than a dozen Emergency Plan drills run at TMI during 1980. The results of these drills were used to develop, and if necessary modify, the specific emergency organizations, communication links, and response procedures described in Licensee's Emergency Plan. Rogan et al., ff. Tr. 13,756, at 117. Neither of the two NRC Staff reviewers was familiar with these drills.
Tr. 15,440 (Grimes, Chesnut).
The NRC Staff draftsmen who developed the guidance in NUREG-0654 and -0696 had no knowledge of any of these facts.
While it may be that the NRC Staff's desire to divide the protective action decisionmaking function from the Emergency Director's other functions by moving that decisionmaker out of the control room and out of the plant to the EOF is an appro-priate concept of operations for the minimum staffing levels suggested in NUREG-0654 (see Tr. 22,976-81 (Tourtellotte)),
there is no reason to believe that it is the only appropriate way for organizing a licensee's emergency response. The NRC Staff explicitly recognized that alternative concepts of operation are possible. Tr. 15,485 (Grimes).
In the face of this evidence, the conclusion which Licensee urged on the Licensing Board, and which it either misunderstood or ignored, is that by providing the onsite Emergency Director with appropriate supporting staff the
! Emergency Director can properly discharge all of his t
I l responsibilities, including making protective action recommendations to the state. Moreover, Licensee's uncontradicted evidence shows not only is there no need to transfer the authority to make protective action recom-mendations offsite, but that any such transfer during the ear?.y hours of an accident is undesirable since it increases the likelihood that there will be a significant misunderstanding about important plant operating parameters or radioactive releases.28 C. THE LICENSING BOARD'S DIRECTION THAT DURING THE EARLY HOURS OF AN EMERGENCY DECISIONMAKING AUTHORITY BE TRANSFERRED OFFSITE TO THE EMERGENCY SUPPORT DIRECTOR CONSTITUTES AN IMPERMISSIBLE INTRUSION INTO THE LEGITIMATE INTERNAL MANAGEMENT RESPONSIBILITIES OF LICENSEE The broad authority of the NRC to regulate and control activities under the Atomic Energy Act at privately-owned utilization facilities is unquestioned. See, e.g.,
28 The Licensing Board also identifies a need for face-to-face communications between Licensee and the state as a basis for its decision. I.D., 1 1388. The Licensing Board dismisses Licensee's proposed solution to the communication problems by holding that " Licensee has not defined those solutions or sought to undertake them." Id.
In fact, the evidence of record indicates that these
- perceived communication deficiencies surfaced only during the i first one or two drills with the state (Tr. 23,088 (Rogan)),
l- and since then Licensee has offered to install an additional l dedicated phone for transfer of information to the state, but at the time of the hearing the state had not yet responded to that offer (Tr. 23,089-90 (Rogan)). Certainly this does not constitute an undefined solution that Licensee has not sought to pursue. .
l l
I Atomic Energy Act of 1954, as amended, FS 103(a), 161, 42 U.S.C. SS 2133(a), 2201. But even this broad grant of author-ity must have certain limits. In most cases, the NRC has sought to regulate the civilian nuclear power industry by promulgating generalized regulations and design criteria and requiring license applicants to demonstrate compliance. See, e.g., 10 C.F.R. Part 50, Appendix A. The NRC's emergency planning regulations are no different. See, e.g., 10 C.F.R.
S 50.47(b), and Part 50, Appendix E. The issue raised by Exception No. 3 is whether the Licensing Board at the urging of the NRC Staff has over-zealously construed the emergency planning regulations by making judgment decisions properly vested with Licensee.
Counsel for the NRC Staff has candidly stated that the issue raised by Exception No. 3 "is a very, very close question and it really is one that is quite judgmental. There are advantages and disadvantages on either side." Tr. 23,081; see also Tr. 23,059-60, 23,062 (Tourte11otte). The NRC Staff also is of the view that reasonable assurance exists that appropriate protective measures can and will be taken in the event of a radiological emergency at TMI (see 10 C.F.R.
S 50.54(s)(2)(ii)) whether or not the person making protective l
action recommendations is located in the control room or at the EOF. Tr. 22,950 (Chesnut). In view of these concessions, Licensee is at a loss to explain why the NRC Staff has given no l weight to the fact that Licensee, in developing its Emergency _
I Plan, has decided that it would prefer to station the person making protective action recommendations in the control room rather than at the EOF during the early hours of an emergency.
Tr. 22,953 (Chesnut).
As explained by Licensee during the hearing, one factor influencing this decision is the background and experi-ence of particular personnel currently available to Licensee.
In assessing the onsite management talent available to respond to an emergency, Licensee has concluded that the optimum use of its personnel is to station the most senior corporate official, as Emergency Director, in the control room. The people designated for that post are capable of understanding and making use of information that is directly available in the control room. In order to assure that the Emergency Director is not overwhelmed with responsibilities, Licensee has provided the Emergency Director with sufficient staff to ensure that the Emergency Director is not drawn into the minute-by-minute response to the accident. Given both the management talent available and the location of the Emergency Director in the control room, Licensee believes that, until its primary Emergency Support Director can arrive at the EOF, the Emergency Director is best suited to provide protective action recom-mendations to the state. If Licensee were to take the person-nel it proposes to designate as Emergency Directors-out of the t
, control room, it would lose the sp9cial abilities of these people to understand and use information that is directly l
i available in the control room. See Tr. 23,045-50, 23,077-78,
}
l 23,085-86, 23,091-96 (Rogan).
It is uncontroverted on this record that the only factor entering Licensee's consideration in this matter is what it perceived to be the most effective means for protecting the public health and safety. There are no resource constraints, either in terms of finances or personnel availability, which influenced Licensee's decision. Tr. 23,097 (Rogan). Rather, it was Licensee's considered judgment that the best means of utilizing the technical and management talent available to it was by placing the senior corporate official in the control room during the first four hours of the accident. The contrary Licensing Board judgment is an unwarranted, and to Licensee's knowledge totally unique, invasion of Licensee's management prerogatives. It is, in Licensee's view, inconsistent with the Commission's own perspective of its regulatory responsibil-ities 29 and constitutes treatment different than that afforded to other licensees.30 The Licensing Board's judgment is wrong 29 In its decision in Federal Tort Claim of General Public Utilities Corp., CLI-81-10, 13 N.R.C. 773, 775 (1981), the j Commission described its role in regulating nuclear power as i
follows:
Within [the] framework (of the Atomic Energy Act] the I regulated industry (i.e., the licensee's and their suppliers and consultants) bears the primary responsibility for the proper construction and safe operation of licensed nuclear facilities.
30 There is testimony of record that the TVA has proposed an emergency response organization that does not place the decisionmaker at the EOF; the Commission nonetheless has found that organization acceptable. See Tr. 15,490-94, 15,524-25 (Grimes).
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as a matter of policy and is totally unsupported by the applicable facts or law.
CONCLUSION For all of the foregoing reasons, Licensee Exceptions Nos. 1 and 3 should be granted, and the Licensing Board's Partial Initial Decision modified accordingly.
Respectfully submitted, SHAW, PITTMAN, POTTS & TROWBRIDGE George F. Trowbridge, P.C.
Thomas A. Baxter, P.C.
Robert E. Zahler Delissa A. Ridgway Counsel for Licensee
. 1800 M Street, N.W.
Washington, D.C. 20036 (202) 822-1000 Dated: March 10, 1982 l
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APPENDIX l
9
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Lic 2/9/81 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289 dCN ) (Restart)
SI (Three Mile Island Nuclear )
Station, Unit No. 1) )
LICENSEE'S TESTIMONY OF ROBERT E. ROGAN, GEORGE J. GIANGI AND ALEXIS TSAGGARIS ON THE ADEQUACY OF ONSITE EMERGENCY PREPAREDNESS AT THREE MILE YELAND, UNIT 1 Volume 1 -- Testimony 1
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1 Table of Contents Page Outline
' Testimony I. Introduction (Questions 1-6) ....................................... 1 II. Development of the TMI-l Emergency Plan ,
(Questions 7-13) ......................................
6 III. Overview -- Concept of Operations (Questions 14-23) .................................... 13 IV. Organization and Coordination (Questions 24-53) .................................... 24 V. Initial Accident Assessment (Questions 54-75) .................................... 66
, VI. Initial Accident Notification (Questions 76-84) .................................... 86 c73 C- VII. Onsite Emergency Response (Questions 85-87) .................................... 93 VIII.Offsite Emergency Response (Questions88-111) ................................... 96 IX. Maintaining Emergency Preparedness (Questions 112-117)................................... 114 Figures 1-6 Tebles 1-3 Appendix A -- Professional Qualifications Appendix B -- Abbreviations Appendix C -- Onsite Emergency Planning Contentions
3 Outline This testimony deals with the adequacy of onsite emergency l'
preparedness at Three Mile Island, Unit 1. It addresses short-term action item 3 and long-term action item 4 from the Commission's August 9, 1979 " Order and Notice of Bearing,"
Board Question 4, and the onsite emergency planning contentions raised by intervening parties in this proceeding. In addition, this testimony demonstrates Licensee's compliance with the Commission's recently revised emergency planning regulations (45 Fed. Reg. 55402-13 (August 9, 1980)) and with the guidance set forth in NUREG-0654 (Rev. 1, November, 1980).
I. Introduction. The witnesses are identified, their involvement with emergency preparedness at TMI is described, dS. the purposes and organization of the testimony are explained, C
and the guidance used in developing the TMI-l Emergency Plan is set forth.
II. Development of the TMI-1 Emergency Plan. The historical development of the initial and three revisions to the Emergency Plan is described. The coordination between the Emergency Plan, on the one hand, and other TMI programs, the state emergency plan, the five county emergency plans, and local emergency preparedness, on the other hand, is explainud.
The status of NRC and FFMA reviews is set forth.
III. Overview -- Concept of Operations. The division of responsibility between onsite and offsite emergency planning is explained. Licensee's emergency preparedness program at TMI,
N
' including the distinction between the Emergency Plan and the
/ Implementing Document, is described. Major elements of the E Emergency Plan are summarized through a hypothetical appli-cation of the Emergency Plan to a small break loss-of-coolant
( ' accident.
IV. Organization and Coordination. There are three parts to this section. The first part describes the various emergency organizations, both,onsite and offsiter the 16tter of agreement between Licensee and certain offsite agencies are discussed in this lurt. The second part describes the onsite and offsite emergency response facilities. And, the third part describes the communication links between the various emergency response facilities.
- . v. Initial Accident Assessment. The information necessary to assess an emergency condition at TMI is described.
The classification of accidents is explained, incluhling a definition of protective action guides and an analysis of Licensee's emergency action levels. Th monitoring and assessment of radiation releases is described. This discussion
. includes an evaluation of ARAC, Licensee's REMP, and real-time offsite monitoring devices that can be remotely read.onsite.
VI. Initial Accident Notification. The initial calls to Dauphin County and PEMA are identified. The reason why the other four risk counties are not called, except in a General Emergency, is explained. The role of BRP in this communication l scheme is summarized. Public dissemination of information is described.
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t VII. Onsite Emergency Response. The mobilization of Licensee's emergency organizations and the onsite equipment l C !
available to assist in responding to an emergency is summa- '
rized.
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VIII. Offsite Emergency Response. This'sectlon demonstrates the coordination between Licensee's onsite emergency plan and the offsite emergency response plans. The plume exposure pathway EPZ and the ingestion exposure pathway EPZ for the TMI site are identified. The geographic extent of the plume exposure pathway EPZ is . justified in terms of the functions necessary for an adequate offsite response, including public education, early warning, notification to the public about the emergency, and ~ protective action options.
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- IX. Maintaining Emergency Preparedness. Licensee's 1
program to maintain an adequate state of emergency preparedness at TMI.is described. This program consists of training, drills and exercises, and ann'ual audits and reviews of the Emergency l Plan.
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III. Overview -- Concept of Operations
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Q.14 Describe the division of responsibility for emergency planning between Licensee and the Commonwealth of l Pennsylvania, f
s.
A.14 The assignment of planning responsibilities is clearly defined in state and federal regulations.
NUREG-0654 details the objectives and criteria necessary to develop complete and comprehensive emergency plans. Specific areas of responsibility are emphasized. In general, Licensee is responsible for all activities which occur onsite while the state and counties are responsible for offsite activities.
In order to fulfill its onsite responsibilities, Licensee relies on various offsite agencies, both governmental and private, to provide assistance beyond that available onsite. Similarly, the Commonwealth of Pennsylvania relies on Licensee to provide necessary information on plant status and
, radiation releases so that the state and county I
governments can carry out their offsite responsi-bilities.
EEP .- 15 (c.') Recognizing the joint nature of their responsi-EP- 15(E) bilities, Licensee and the relevant governmental agencies have taken steps to ensure a coordinated s
response. These steps include coordinated preplan-ning, redundant communication systems, and
1 17 Licensee-conducted training sessions for offsite agencies. Periodic drills' test communication links,
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offsite response of state and county agencies, and coordination among the various agencies.
k Q.15 With respect to the onsite responsibilities you referred to, describe the emergency preparedness program at TMI-1.
A.15 The Vice President Nuclear Assurance is responsible for nuclear safety assessment, quality assurance, training and education functions, system labora-tories, and emergency preparedness. This Vice President reports to the Executive Vice President, GPU Nuclear. There currently are nine personnel
- assigned to the Emergency Preparedness Department who O
EL' are located at TMI, including the Manager-Emergency Preparedness and a site Supervisor-Emergency Preparedness. The Emergency Preparedness Department is charged'with overall responsibility for emergency planning and for assuring the maintenance of an appropriate state of emergency preparedness at TMI.
In order to carry out these responsibilities, the TMI Emergency Preparedness Department has developed two separate, but coordinated, documents: the TMI-l Emergency Plan and the Implementing Document.
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! Q.16 Please explain further the distinction between the TMI-l Emergency Plan and the Implementing Document.
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0, A.16 The role of the Emergency Plan is as follows:
f Pursuant to 10 C.F.R. SS 50.54(q) and (u), an operator of a licensed nuclear power plant is
- s. required to submit a radiological emergency response plan which meets the standards of 10 C.F.R.
l S 50.47(b) and Part 50, Appendix E. This emergency l
l plan describes the facility's overall state o,f emergency preparedn'ess. It is a detailed document which includes, among other matters, organization and communication concepts, emergency action levels, assessment actions, emergency facil'ity details, emergency mobilization and response actions, p training, recovery, and letters of agreement with -
outside agencies. The emergency plan provides the
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basis for developing additional documents, such as l
l the implementing procedures, training program, and equipment inventories.
The role of the Implementing Document is as follows:
The Implementing Document provides a single source of pertinent and significant information related to emergency preparedness at TMI-1. It contains the procedures that would be required to: (a) ensure the l operational readiness of the Emergency Plan, and (b)
( direct the proper response by emergency personnel.
4 k
, While the Emergency Plan is a basic reference document, the Implementing Document is actually used by station personnel during an emergency.
{ The Implementing Document is distributed to those individuals, agencies, organizations, and facilities requiring the immediate availability of such information in an emergency. The detailed EPIP's included in the Implementing Document will, as necessary and appropriate, be used to assess conditions, classify the emergency, make required notifications, provide directions for requesting assistance, and provide step-by-step instructions for initiating protective and corrective actions.
9 E3- Q.17 What are the basic elements in responding to an emergency at TMI-1 that you considered in developing the Emergency Plan?
A.17 The basic elements in responding to an emergency are':
- 1. Assessment of plant conditions and clas-l sification of the emergency following an accident.
- 2. Notification of offsite agencies and support groups.
i 3. Mobilization of the applicable portion of the
) emergency organizations to cope with the l situation and continue accident assessment.
These elements were considered in establishing the TMI-1 emergency response organization, communication l
.w... - . . . -
to capabilities, need for response facilities and equipment.
Q.18 Assume that there was a small break loss-of-coolant accident ('LOCA") greater than make-up capacity at TMI-1. Briefly describe how the Emergency Plan would l
{-
l be implemented.
A.18 A small break LOCA of this magnitude initially would be indicated by makeup tank level decreasing and makeup flow increasing. , Reactor coolant presiure would decrease, the reactor and turbine would trip, t
l and the emergency core cooling system ("ECCS") would initiate. Containment pressure would increase such that the cause of ECCS initiation could be either high containment pressure (4.0 psig or greater) or
,h7 low reactor coolant pressure (1600 psig or lower).
9 The control room operators initially would be made aware of the situtation by alarms, instrument readings, or reports. The operators would ensure that the shift foreman and the shift supervisor were immediately informed.
I The shift supervisor, when informed of the emergency, is responsible for assessing the emergency (e.g.,
plant systems and cctor core status, and radiolog-l 1
ical conditions). Me would determine what immediate j
t actions must be taken and ensure that the procedure l
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for " Loss of RC/RC Pressure (Small Break LOCA)
Causing Auto BP Injection" (1202-6B) is implemented.
C The shift technical advisor would advise and assist
,- the shif t supervisor on matters pertaining to the safe and proper operation of the plant with regard to l nuclear safety. One step in the follow-up action section of procedure 1202-6B would refer the operator I '
to the EPIP on Site Emergency (1004.3), and direct him to inform the shift supervisor or shift foreman that a Site Emergency action level had been reached.
. In this case, the shift supervisor would classify and declare the emergency as a Site Emergency and would implement the' applicable EPIP. This would set in h2), motion corrective actions and offsite notifications.
9:'
We believe that the. emergency could be assessed and declared within 10 minutes.
Q.19 Af ter the i'nitial assessment function had been
, completed, what would happen next?
l A.19 The shift supervisor would assume the duties of the Emergency Director and announce to all station personnel over the public address system in Units 1 and 2 that a Site Emergency had been declared in Unit 1 and instruct the onsite emergency organization
- personnel to report to their stations. All non-essential personnel would be instructed to assemble
l 2-at the respective Unit 1 and Unit 2 warehouses.
( EEF)- 1 Initia1 notifications would de maae as follows: (13 EP LhfG) Dauphin County EOC; (2) PEMA EOC (staff duty EEP - 15(B) officer), (3) unaffected contro1 room; (4) NRC
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(Bethesda); (5) Institute of Nuclear P.ower Operations
("INPO"); (6) Babcock & Wilcox ("B&W"); and (7)
American Nuclear Insurers ("ANI").
PEMA would immediately notify BRP and all five counties within the ten Lile radius. BRP would confirm the existence of an emergency situation at TMI by activating the Radiological Line to the Unit 1 Emergency Control Center (control room). This line would be manned to maintain continuous communication throughout the emergency. Once BRP has verified that
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all five counties have been notified, it would advise the TMI Emergency Director accordingly.
Parallel to these notifications, the duty section
! superintendent would be called and informed of the emergency by the Emergency Director (shift super-visor). Callout of duty section personnel required to augment the onsite and offsite emergency j organizations would begin.
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Q.20 What might Licensee's response be to this situation?
l A.20 Upon declaration of a Site Emergency, the entire onsite and offsite emergency organizations would l
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report to their respective emergency facilities. The
( onsite emergency facilities include the Emergency l
Control Center ("ECC"), Technical Support Center
("TSC") and Operations Support Center ("OSC"). The 7
s offsite emergency facilities include the Nearsite Emergency Operations Facility (" EOF"), Alternate Emergency Operations Facility ("AEOF"), Environmental Assessment Command Center ("EACC") and Parsippany l Technical Functions Center ("TFC").
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The ECC, located in the Unit 1 control room and adjacent shift supervisor's office, is the area where the command and control of all site-related emergency efforts and plant operations take place. Key 4'- personnel stationed in the ECC would be the Emergency b
Director, Radiological Assessment Coordinator
("RAC"), Operations Coordinator and the Communicator.
Major functions performed in this facility include onsite and offsite radiological assessment, offsite notifications, operational control of the plant and communication of technical data to BRP and NRC.
The TSC, located in proximity to the TMI-1 control room, contains the instrumentation needed to monitor plant status for a safe shutdown of the reactor when i the control room is uninhabitable. The key personnel l
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stationed in the TSC would be the TSC Coordinator and TSC engineers from the various disciplines. The TSC serves as an area outside the control room to l accommodate personnel acting in support of the command and control functions by furnishing more in-depth diagnostic and corrective engineering assistance.
l The OSC, located at the radiological controls access o
control point, provides an area in which shift.
personnel can gather for subsequent assignment to duties in support of emergency operations. Key personnel manning this center would be the OSC Coordinator, Chemistry Coordinator, Radiological j;3 controls Coordittator and Emergency Maintenance Coordinator. The major functions of these personnel are to initially dispatch radiological monitoring teams and to support operations in the areas of chemistry, radiological controls and maintenance.
The EOF, located at the TMI Observation Center, serves as the central point for: (a) providing overall corporate management and direction in responding to an emergency, (b) coordinating administrative and logistical support, (c) inter-facing with state and county representatives, and (d) 7
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II establishing the basis for long-term recovery
( efforts. Key personnel located at the EOF would be the Emergency Support Director, Emergency Support Staff, Assistant Environmental Assessment
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Coordinator, Public Affairs Representative, Emergency Planning Representative, Group Leader Chemistry Support, Technical Support Representative, and NRC and state representatives.
- The AEOF, located at the Crawford Station in Middletown, houses key positions of the offsite emergency organizations. Personnel at the AEOF would be the Group Leader Administrative Support, Group Leader Radiological Controla Support, Group Leader 45h Security Support and Maintenance and Construction t=r -
Manager. Major functions performed at this facility would be security and dosimetry processing of support personnel, maintenance support, call-out of additional support personnel and administrative support. The AEOF also serves as a back-up EOF,
, should the EOF become uninhabitable.
The EACC, located at Olmsted (Harrisburg International) Airport would be manned by the Environmental Assessment Coordinator ("EAC") and his staff of scientists. The major functions of these
lb personnel would be to perform and assess all offsite
(~ radiological and environmental monitoring.
The TFC, located in Parsippany, New Jersey, is where
( the Group Leader Technical Support and his staff report. The major functions of these personnel would be to provide technical leadership, guidance, analysis, evaluation and recommendations to the plant staff. .
Q.21 What would the offsite response be in this situation?
A.21 Based on the state and county emergency response plans, and our discussions with statt and county personnel, the following additional notifications CN EEP -- I wou1d eaxe piace. PExA wou1d noeify BaP and the five
~
- E P - 4 (G) risk counties. BaP would immediately call Tur-1 to make.an initial radiological assessment and to verify Licensee's ca11 to PEMA. Once the emergency has been assessed, BRP would call PExA, inform them of plant status, and advise them whether any protective l
l actions need be taken. BRP would then activate its I
emergency organization and establish an open line of communications with Licensee's RAC located in the ECC.
/ Q.22 How would the emergency be closed out?
IT A.22 In the specific case of the small break LOCA, which f was initially classified as a Site Emergency, the s
emergency would be closed out by shutting down and j
, cooling down the reactor and isolating the leak.
k The Emergency Director and Emergency Support Director the*, have joint responsibility for determining and declaring when the emergency situation is stable and has entered the recovery phase. They would evaluate the status of the emergency by monitoring instruments and reviewing all current and pertinent data
. available from emergency response and radiological monitoring teams. They would consider the emergency under control and in the recovery phase only when the h7g following general guidelines are met:
%f'
- l. Radiation levels in all in-plant areas are stable or are decreasing with time.
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- 2. Releases of radioactive materials to the environment from the plant are under control or have ceased.
- 3. Containment pressure is at normal levels.
- 4. Reactor plant is stable and in a long-term s%fe shutdown condition.
- 5. Any fire, flooding, or similar emergency conditions are controlled or have ceased.
Based on the sequence of events, one of the following would occur:
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- 1. A lower class of emergency might be declared by I the Emergency Director and the appropriate
( procedures would be implemented.
- 2. The Site Emergency might be closed out, with the concurrence of the Emergency Support Director, provided no recovery operations were required.
C 3. The Site Emergency might be shifted to a recovery mode by implementing the Recovery Operations Procedure (1004.24).
If the emergency is being reclassified, the NRC, Unit 2 control room, and other organizations as specified in the appropriate EPIP would be notified. BRP is in continuous contact with the TMI site and vould be updated as necessary. BRP, in turn, would notify PEMA, who would notify the five risk counties.
o If the Recovery Operations Procedure is being implemented, the appropriate organizations would be notified of the closeout of the emergency and that recovery operations are about to begin.
Q.23 Would you briefly describe what would happen if, i
instead of closing out the emergency, the situation continued to worsen?
A.23 Accident assessment would continue throughout the emergency, and if conditions warrant, the Emergency Director would escalate the emergency to a General t
Emergency. Notifications would be made to the five 1
risk counties -and to other organizations as specified in the EPIP for a General Emergency.
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- 1 The assessment actions for the General Emergency
(~ generally would be the same as for the Site Emergency, with some possible shift of emphasis to
- greater offsite monitoring and dose projection L efforts extending to distances farther from the plant. Additionally, since the projected doses are likely to be much closer to the U.S. Environmental Protection Agency's (" EPA") protective action guides
(" PAG's"), greater emphasis would be placed on the assessment of release duration for the purposs of making protective action recommendations.
IV. Organization and Coordination Q.24 Would you describe Figure 1, Licensee's Onsite 4Th Emergency Organization? -
O A.24 m ajor functional responsibilties within the The '
onsite emergency organization are vested in the Emergency Director, the Operations Coordinator, the OSC Coordinator, the RAC, the TSC Coordinator, and the Security Coordinator. In addition, the Communicator provides communications support for the ,
onsite emergency organization.
o c d ,o u 1 M a.b fe.w4 ca.. O'.,ec) n .
The Vice President TMI-1,.":..:;::.TMI-1, or'their
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designated a.1 ternate, performs the duties of the I Emergency Director. Until his~ arrival at the site, l
l i
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fo l l
4 the shift supervisor assumes the duties of the
- EP- 4(Jh) Emergency Director. If the shife supervisor is unavailable or becomes incapacitated for any reason, g the shift foreman assumes this position. The Emergency Director-has the authority and the responsibility to immediately and unilaterally initiate any emergency action, including providing protective action recommendations to authorities responsible for implement'ing offsite emergency measures. The Emergency Direct.or must classify and declare the emergency, and ensure that all required notifications are made, including those to offsite emergency response organizations. The Emergency Director implements the TMI Emergency Plan through the use of specific EPIP's, activates necessary portions of the emergency organization, and performs the other functions described in Section 4.5.1.3.1 of the Emergericy Plan. The Emergency Director would report to the ECC, and communicate with the Operations Coordinator, TSC Coordinator, RAC and Security Coordinator. He also would communicate with l
the offsite emergency organization through the EOF.
l The Operations Coordinator is responsible for directing operations and operations support activities through the shif t supervisor and the OSC k
gl
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. i Coordinator. The Operations coordinator reports to I
the Emergency Director and works closely with him in assessing plant conditions. He has no direct
(- communication links with onsite or offsite agencies.
t The OSC Coordinator is responsible for ' supporting operations in the areas of maintenance, radiological controls and chemistry. He reports to the Operations Coordinator and has the Emergency Maintenance Coordinator, Radiological Controls Coordinator, and Chemistry coordinator reporting directly to him.
The RAC is responsible for guiding the Radiological Controls Coordinator and the Radiological Analysis Support Engineers. In addition, he is responsible for coordinating the activities of various emergency response teams. As required, he would ditect the OSC Coordinator to dispatch onsite and offsite radiolog-ical monitoring teams that would report directly back l to him. He would coordinate initial radiological l
i assessment activities, review results, and report l -
I findings and make recommendations to the Emergency Director. He would interface with the EAC on radiological and environmental matters. The RAC maintains communications with BRP in order to update i
them on emergency status.
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22' The TSC Coordinator and his staff of engineers report
( to the TSC. They are responsible for analyzing current and projected plant status and, through close communication with the Emergency Director via the b Communicator, providing technical support, in-depth diagnostic and corrective engineering assistance, and recommendations regarding corrective actions. The specific duties of this group are described in Section 4.5.1.3.2.b of the Emergency Plau.
The TMI site security force operates in accordance with requirements established in the Security Plan and associated procedures. In emergency situations, the security force reports to the Security Coordinator, who, in turn, reports to the Emergency
[gg ,
Director. The security force is responsible for ,
personnel accountability, site access control, and plant security.
The Communicator functions as a communication liaison i between the Emergency Director and the onsite and i offsite emergency organizations. He reports to the ECC (shift supervisor's office) and controls the flow of information across the Operational Line and maintains communication between the TSC and the ECC with an intercom. Designated Communications i
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- e Assistants are responsible for maintaining communication with the NRC, making necessary
{ notifications to offsite agencies, and keeping a record (log) of all incoming and outgoing communica-C tions.
Additional information on the onsite emergency organization is included in Section 4.5.1.3 of the ,
Q.25 How does each member of the TMI-1 staff know what position he is to fill in the onsite emergency organization?
A.25 A duty roster has been developed to ensure that all positions in the onsite emergency organization are gag fully staffed. One section of the duty roster is CD' -
always on call. Each individual on the duty roster -
is preassigned a position in the onsite organization and is instructed as to what his functions are, where he is to report, and to whom he is to report. Duty roster personnel are responsible for maintaining a working knowledge of the current TMI Emergency Plan, l Implementing Document, and other related station programs, plans, and procedures. Individuals generally are assigned positions in the emergency organization which closely parallel their normal everyday duties. Particular assignments.are based on 1
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the selection criteria included in Table 8 of the Emergency Plan, training received, and driving distance from residence t'o the site.
' When an emergency initially is declared, are there O.26 s- sufficient personnel on-shift to staff the onsite emergency organization? .
A.26 Yes. Table 2 of this testimony shows the minimum EP -4(o) shif t complement of 20 onsite at all times, and the EP - 4C7)l2) onsite emergency organization positions that they EP- 4l.7)(3) would fill upon declaration of an .mergency. This is twice the on-shift complement required by Table B-1 of NUREG-0654 (Rev. 1). Moreover, this on-shift complement is more than adequate to promptly perform the initial accident assessment and notification
@ . functions of the emergency organization.
In particular, there are adequate personnel so that EP 4l7)h) the Emergency Director (shife supervisor) may assign
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two control room operators to monitor the plant (CRO
- 1 and Tagging & Switching CRO), a third :::t:01 ::::
operator to initiate calls to Dauphin County, PEMA, NRC and the unaffected control room, and additional personnel (chosen from the four auxiliary operators, two radiological controls technicians, and four maintenance personnel available) to conduct onsite and offsite r'adiological surveys.
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f aC Q.27 How many people with radiological controls (health g physics) training will be available to man the onsite emergency organization?
A.27 Immediately available would be one radiological
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IP-4(D) controis foreman and three radiological controis EP- 4(J)M technicians. The three technicians can be sp11e up to provide radiological monitoring and in-plant radiological controls. Within sixty minutes of the declaration of an emergency, a senior radiological controis engineer would be available to assume the position of RAC, two Radiological Analysis Support Engineers would be available to assist the RAC, and three additional radiological controls technicians would be available. In summary, four people trained P
in radiological controis would be available initially and ten (six additional) would be available within sixty minutes.
In addition, the EACC can be manned and operational within six hours after declaration of an emergency.
The EACC can supply four one-man teams and a two-man mobile monitoring laboratory. This can be augmented by three additional one-man teams, should it become necessary, i Q.28 Would you describe Figure 2, Licensee's Offsite
[ Emergency Support Organization?
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, A.28 The key elements of the offsite emergency support organization include the Emergency Support Director, Emergency Support Staff, Public Affairs Representa-( tive, Emergency Planning Representative, Group Leader l Administrative Support, EAC, Group Leader Radiological Controls Support, Group Leader Chemistry Support, Group Leader Technical Support, Maintenance and Construction Manager, and Emergency Support Communicator. The offsite emergency support organization provides technical and logistics support in the event of a serious or potentially serious emergency and is staffed by personnel from the normal station and technical support organizations, h
- The Emergency Support Director is the senior utility -
1 management representative at the TMI site. He reports to the EOF and is responsible for directing the offsite emergency support organization, and for providing advice and guidance to the Emergency Director on accident management responsibilities, i
The Emergency Support Director can monitor communica-tions on the Operational and Radiological Lines, and communicates directly with the Emergency Director on the Emergency Director's line.
I The Emergency Support Staff reports to the Emergency Support Director at the EOF and assists the Emergency
A7 Support Director by communicating with the offsite r
s emergency support organization Group Leaders and by providing status reports to the Emergency Support r~ Director.
( !
l The Public Affairs Representative reports to the !
Emergency Support Director from the EOF. He is responsible for implementing the Emergency Public Information Plan, preparing technically accurate news releases, and updating GPU Nuclear management on the status of the emergency.
The Emergency Planning Representative reports to the Emergency Support Director from the EOF. He is .
,, responsible for providing information relating to E;7 -
onsite, offsite, and state and local emergency -
facilities, and communication, personnel and resource capabilities. He also provides advice on the procedural requirements of Licensee's Emergency Plan.
The Group Leader Administrative Support reports to the AEOF. He is responsible for administrative and logistics functions required to support the onsite and offsite emergency organizations. In addition, he is responsible for security processing and badge issuance to personnel requiring site access.
l
l hY The EAC reports to the EACC and is responsible for the radiological environmental monitoring program
("REMP"). Once the EACC is activated, the EAC assumes control of offsite radiological and environmental monitoring and assessment from the RAC.
I Be communicates with the RAC in the ECC on the Environmental Assessment Line.
l The Group Leader Radiological Controls Support reports to the Emergency Support Director from the AEOF. He is responsible for all aspects of l
radiological controls support to the onsite emergency organization, including thermoluminescent dosimeter
("TLD") issuance, whole body counting, and obtaining additional equipment and personnel as necessary. The g ,
Radiological Controls Manpower Support and Personnel Monitoring Coordinators report directly to him.
The Group Leader Chemistry Support reports to the Emergency Support Director at the EOF. He is responsible for all aspects of chemistry support, including the establishment of a chemistry monitoring program and for obtaining additional equipment and personnel.as necessary.
The Group Leader Technical Support reports to the Emergency Support Director from the Parsippany TFC.
l
N Be is responsible for providing technical leadership, analysis, evaluation and recommendations to the onsite TSC Coordinator with respect to plant f
- conditions, reactor core status, and subsequent plant operations. He communicates with the o,nsite TSC Coordinator and the Technical Support Representative at the EOF on the Parsippany/TMI Line.
The Maintenance and Construction Manager reports to the Emergency Support Director from the AEOF and is responsible for maintenance support to the onsite and offsite organizations. He provides additional 1
maintenance personnel and equipment as required. The Group Leader Maintenance Support reports to him.
I,$ $> -
The Emergency Support Communicator reports to the l Emergency Support Director at the EOF and is responsible for operation of the communication systems at the EOF and for the coordination of requests for outside assistance. He ensures that the primary and back-up communication systems are activated and operational, and maintains records of communications and status boards.
l Additional information on the offsite emergency L support organization is located in Section 4.5.1.4 of l the Emergency Plan.
l l
l !
h Q.29 How do personnel know their assignments in the l{ offsite emergency support organization?
A.29 A duty roster has been developed which assigns personnel to specified positions in the offsite
{
emergency support organization. Emergency responsi-bilities are assigned on the basis of the selection criteria set forth in Table 8 of the Emergency Plan, the individual's overall experience and training, and his current job position. The offsite personnel become familiar with duty stations and responsi-a attending periodic training sessions and bilities 'f participating in test exercises and drills. ,
Personnel assigned functional responsibilities in the offsite emergency support organization are expected to maintain a working knowledge of the current TMI Emergency Plan, Implementing Document, and other related station programs, plans, and procedures as may be required to perform their functions, l
l Q.30 Bow long would it take to staff the offsite emergency support organization?
l A.30 Depending on the emergency classification, all or part of.the offsite emergency support organization would be directed to report to predesignated l
locations. Upon arrival at the emergency response facility, personnel initially would activate
3l emergency communication systems and computer-based
( data links; inspect, inventory and place in operation as appropriate the emergency equipment present; and complete all tasks directed by the appropriate procedures. As personnel continue to ar. rive, the various functional areas would become fully operational and would support the onsite organiza-tion. The entire offsite emergency support organination can be fully manned within six hours.
Q.31 Would you describe the basic function of the offsite emergency support organization, noting particularly how those functions differ from the responsibilities of the onsite emergency organization?
l l A.31 The purpose of the offsite emergency support la organization is to provide overall corporate management and direction of emergency response, to provide technical advice and assistance, and to coordinate long-term logistical and administrative
! support for the onsite emergency response organiza-i tion and activities. In general, the offsite
, emergency support organization will:
- 1. Support the onsite emergency organization in engineering and technical matters with accident analysis, assessment, and technical advice on appropriate corrective actions to stabilize the plant.
L 2. Provide'for environmental monitoring and assessment in support of the onsite emergency organization.
l l
1
l 37 l
- 3. Provide liaison and communication with the NRC
.- and appropriate state and county agencies.
- 4. Provide for the dissemination of information to the public and the news media.
- 5. Provide security support.
- 6. Acquire materials, equipment, and services necessitated by the emergency.
- 7. Provide assist?.nce for reentry operations and post-accident planning.
- 8. Assign post-accident investigation and review renponsibiliti'es. -
These functions need not be accomplished immediately EEP- 40)) afeer dec1aration of an emergency. Rather, they are supplementary to, and in support of, the functions being performed by the onsite emergency organization.
This characteristic distinguishes the responsi-o 5 ~
bilities of the offsite emergency support organiza-tion from the onsite emergency organization.
Q.32 flould you describe Figure 3, Licensee's Long-Tera Recovery Organization?
A.32 A long-term recovery organization has been developed, i
- which would assume command of the emergency response l
from the onsite and offsite emergency organizations in cases where post-accident conditions either would be complicated or would be expected to extend over a L long period of time. The key elements in the GPU Nuclear recovery organization are: the Office of the
33 l .
President GPU Nuclear, Vice President Administration,
( Vice President Communications, Vice President Radiological and Environmental Controls, Vice President Maintenance and Construction, Vice President Technical Functions, Vice President Unit l"
Operations, Vice President Nuclear Assurance.
l l The Office of the President GPU Nuclear is responsi-l -
ble for overall recovery operations. This includes overseeing operations of the various functional groups and ensuring that all activities receive
. proper analysis and coordination.
The Vice President Administration is responsible for b providing the necessary administrative / logistics 3h requirements, such as communications, manpower, i
transportation, commissary arrangements, accommo -
dations, clerical support, and temporary office space and equipment.
l The Vice President Communications is responsible for coordinating the exchange of information with public and governmental agencies.
The Vice President Radiological and Environmental Controls is responsible for estab'ishing l policy, coordinating'and reviewing radiation and i
34 envirena:ntal controlo, including in-plent radiological controls management, and monitoring and quantifying the degree of contamination of buildings and personnel. -
(~ The Vice President Maintenance and Construction has I
the responsibility for directing the activities associated with major maintenance tasks and accomplishing field. work for major modifications.
l The Vice President Technical Functions is responsible l
for providing engineering support, technical planning and analysis, procedure support, control room technical support, data management, and support relating to licensing requirements.
o .
c Tmr-I C7
- The Vice President {L' nit ^;er s' f a-- '*"7 'S is responsible for performing all plant operations and maintenance activities, limiting and controlling personnel exposures, terminating or minimizing offsite releases, stabilizing plant conditions, restoring the plant's ability to function normally, l and responding to any further emergencies. He.is responsible for safely and effectively managing the grantities of radioactive gases, liquids, and solids that might exist during the initial phases of recovery.
l l
3fI The Vice President Nuclear Assurance is responsible
(. for implementing the Quality Assurance Plan, all necessary general employee, technical and recovery
- management training programs, and for review of the s
Emergency Plan and Implementing Document to ensure that a high degree of emergency preparedness is maintained for potentially hazardous recovery activities.
Additional information on the long-term recovery organization is located in Section 4.5.1.5 of the Emergency Plan.
G 33 Identify the mejer egencies et the etete level which would respond in the event of an emergency at TMI.and the primary functions they would perform.
c EF -
A.33 Al tate-level emergency response agencies have some common r .onsibilities. Briefly, they are: (a) develop and a tain plans for emergencies; (b) prepare and mainta rocedures for rapid dis-semination of informatio quick assembly of key
. personnel, and timely acquisi n of equipment and other resources; (c) maintain resou es inventories; and (d) identify critical functions and tivities necessary for adequate operational capability ring emergency situations.
l
36 1 Licen;ee end "ar; hey ."; dical Center is included in\ evision 3 to Licensee's Emergency Plan.
(
- 5. Specific ommitments from GPU related companies
(. -- Table 11 the Emergency Plan provides supplementary in raation on the manpower and equipment avails'_le om GPU related companies.
Moreover, with the reorg ization of GPU Nuclear, the executive autho ty that supervises operations at TMI also supervise nuclear related operations at the other GPU mpanies and therefore can assure emergency suppo from
- h ;;;;;nia .
( Q.39 Would you describe Figure 4, Emergency Response
~
Facilities?
)IU2
.A.39 The emergency response facilities are divided into four categories: onsite, offsite-near, offsite-general area, and offsite-out-of-state.
The onsite facilities are as follows:
- 1. Emergency Control Center ("ECC") is the Unit 3 control room and shift supervisor's office.
- 2. Technical Support Center ("TSC") is located in the remote shutdown room, in close neoximity te the Unit 1 control room.
- 3. Operations Support Center ("OSC") is located at the radiological controls access control point.
(
l 1
l ---
lli The offsite-near facilities are as follows:
Is
- 1. Nearsite Emergency Operations Facility (" EOF")
is located at the TMI observation Center, directly east of the site on Route 441.
- 2. Alternate Emergency Operations Facility ("AEOF")
is located at Crawford Station. .
- 3. GPU Nuclear Media Center is located at Crawford Station. j
- 4. Environmental Assessment Command Center ("EACC")
- is located at the Olmsted Airport.
- 5. Dauphin County EOC is located in the courthouse in Harrisburg.
The offsite-general area facilities are as follows:
- 1. Federal EOC is located at Capital City Airport.
, 2. BRP is located in the Fulton Bank Building in Barrisburg.
- 3. NRC Region 1 Office is in King of Prussia, Penn .
sylvania.
- 5. EOC's for the four risk counties other than Dauphin are located in the respective county courthouses.
The offsite out-of-state facilities are as follows:
- 1. NRC headquarters are in Bethesda, Maryland.
- 3. Parsippany Technical Functions Center ("TFC") is located in Parsippany, New Jersey.
l b5 Q.40 Describe the function of Licensee's three onsite emergency response facilities.
A.40 The ECC, located in the Unit I control room and adjacent shift supervisor's office, contains
{-
communications equipment, emergency radiological controls equipment, status boards, a dose projection microcomputer and offsite area maps. Command and control of all site-related emergency efforts originate from this center.
The TSC, located at the 322' elevation of the control building, below the control room, is an area where engineers can provide technical support and analysis to emergency response personnel in the ECC. The TSC
-- contains access to key plant parameters that may be "O
used in assessing accident conditions. Records, drawings, technical manuals, communication systems and other information sources also are located at the TSC. This technical information and communications equipment available in the TSC enable personnel at the center to provide a high level of technical assistance to those responsible for command and control of emergency efforts.
The OSC, located at the 306' elevation of the control
( building, is'the normal radiological controls access
k control point. The OSC contains communications equipment, emergency radiological controls equipment,
(~
offsite area maps and status boards. Shift personnel muster in this area for subsequent assignment to duties in support of emergency operations.
Q.41 Describe the function of Licensee's five offsite emergency response facilities.
A.41 The TMI Observation Center fronting on Highway 441, east of the TMI site, will be the EOF. This facility normally is manned as a public education center and is a well built permanent structure with adjacent parking areas. Sufficient area for helicopter landing is available. The EOF will house the key, o-- technical groups of the offsite emergency organiza-
~
Gi?
tion. In addition, BRP will send a liaison representative, and the NRC will locate its senior site emergency team at this location.
Crawford Station, located approximately three miles north of the TMI site, serves as the AEOF.
Radiological controls equipment, including decon-tamination supplies, will be located here. The AEOF also serves as a staging area for personnel preparing to go onsite. Offsite administrative and maintenance support activities will be conducted from this
(
location.
YO The EACC, located in offices at Olmsted Airport, will be made operational concurrent with the EOF. Once
({
operational, the assessment of all offsite radiolog-
- ical and environmental impacts will be done at the EACC. This includes offsite dose calculations, offsite monitoring of radiological releases via all major pathways, receipt and dissemination of all data
. . _ . received from offsite monitoring teams, and implementation of the REMP.
The Parsippany TFC will be located at GPU head-quarters in Parsippany, New Jersey. The Croup Leader Technical Support and his staff will report to this center. A representative of this group, designated the Technical Support Representative, will be dil dispatched to the EOF to make recommendations to the Emergency Support Director.
The Media Center, located at Crawford Station, con-tains equipment and facilities designed to support timely communications and dissemination of informa-tion on plant conditions and emergency operations.
Commercial facilities will be used to accommodate large press conferences beyond the capacity of the Media Center. Additional information on the Media Center is prcvided in the Emergency Public
dl
, Information Plan for TMI, which is Appendix B to the
~
(~ TMI-1 Emergency Plan.
(
Q.42 Are the emergency response facilities of the state and county governments depicted in Figure 4?
s A.42 Yes. The state EOC, located in the basement of the Transportation and Safety Building in Harrisburg, contains back-up power equipment, communication systems, and necessary supplies to accommodate the various state government agencies that would operate from this EOC. The risk counties also operate EOC's, located in the basements of the respective county courthouses. All have back-up power and the space and equ'ipment needed to ensure a coordinated response
, to an incident at TMI. BRP operates from its offices esp in the Fulton Bank Building in downtown Harrisburg.
Personnel from BRP also are located at the state EOC and at Licensee's EOF.
Q.43 Are the emergency response facilities of the various federal agencies also shown in Figure 4?
A.43' Yes. The Capital City Airport is the location of the federal EOC. The Airport, located about 10 miles WNW of the site, is owned and operated by the Commonwealth of Pennsylvania. The Department of Energy and EPA would be two of the key federal N
47-agencies to conduct operations from this facility,
( which was used for a similar purpose during the Unit 2 accident and proved satisfactory. NRC facilities from which assistance or advice would be requested in
( ~
the event of an accident are the NRC Region I Office in King of Prussia, Pennsylvania and the NRC Headquarters in Bethesda, Maryland.
Q.44 would you describe generally the communication systems linking the emergency response facilities you have just identified?
A.44 The communication systems to be utilized at the various locations consist of both two-way radios and land-line telephone systems. Reliability is provided through redundancy, alternate communication methods, dedicated systems, and routine use to ensure operational reliability. Information that would flow over these systems is divided into two major categories: operational data and radiological data.
This procedure ensures rapid transmission of information directly to key parties having closely related functions, thus eliminating errors associated with second-hand information. The significant networks are the operational Line, the Radiological Line, the Environmental Assessment Line, the s
Parsippany/TMI Line, the Parsippany/B&W Line, the NRC
43 Emergency Notification System (" ENS"), and the NRC
( Health Physics Network Li,ne ("HPN"). By providing well-defined and dedicated communication links, effective accident management from physically
(~
separate control and support centers is achieved.
Q.45 You referred to an " Operational Line". Please describe this ne'twork in more detail.
A.45 The Operational Line is a network of dedicated telephone lines with telephones located in the ECC (shift supervisor's office), OSC, TSC, EOF, AEOF and B&W in Lynchburg, Virginia. See Figure 5(a) of this testimony. The Operational Line permits an unimpeded discussion of plant parameters, system status, core conditions, and other pertinent technical data
{}g ,
necessary to resolve problems in accident mitigation and to keep all emergency response personnel apprised of current plant conditions. This capability enhances the accident management function and decision making process.
Q.46 You also identified a " Radiological Line". Would you describe this network in more detail?
A.46 The Radiological Line is a dedicated telaphone line with telephones located in the ECC (dose assessment area), OSC, EOF, AEOF, and two different areas at
. _ _ _ _ _ m
Q4 BRP. See Figure 5(b) of this testimony. This line permits the communication of plant radiological dose projections, offsite radiation monitoring results and
({ liquid effluent release data to BRP and other key emergency response personnel. -
Q.47 You also referred to the Environmental Assessment, Parsippany/TMI and Parsippany/B&W Lines. Describe these communication links in more detail.
A.47 Each of these dedicated telephone lines provides a capability for a particular type of communication that is anticipated to occur during an emergency.
The Environmental Assessment Line connects the RAC in the ECC (dose assessment area) with the EAC at the
]g . EACC (Olmsted Airport) and the Assistant EAC at the EOF. See Figure 5(c) of this testimony. Dose projection information and radiological assessments will be communicated over this line.
The Parsippany/TMI Line connects the TFC with the EOF and the TSC. See Figure 5(d) of this testimony.
This allows for a rapid exchange of information among the Group Leader Technical Support in Parsippany, the Technical' Support Representative at the EOF, and the onsite TSC Coordinator.
The Parsippany/B&W Line connects the TFC with the B&W technical functions group in Lynchburg, Virginia.
W This estab11shes See Figure 5(e) of this testimony.
a reliable channel of communciation for in-depth C. diagnostic and corrective engineering assistance between the facility operator and the nuclear steam supply system vendor.
Q.48 Please describe the communication links between TMI and PEMA.
A.48 Basically, there are two communication links.* The EP-l firse is the norma 1 teiephone 1and_11ne 11nx. The EP -l5(B) a1 ternate in the event of a telephone system fai1ure is the Nationa1 Warning System ("NAWAS"). NAWAS is a dedicated radio-telephone line designed to provide an immediate means of emergency information flow. The system is tested daily.
WP .
Q.49 Would you also describe the communication links between TMI-1 and Dauphin County?
A.49 Initial contact with the Dauphin County EOC is
~
norma 11y made by teiephone. Back-up communications E P -ISIS) are eurough a cross _ monitoring radio system, This Particular system is tested on a weekly basis.
Q.50 Is it anticipated that TriI would be in direct communication with the other four coun les?
A.50 No, except in a General Emergency, in which event EP-1 r.icensee viii contace each county in para 11e1 wien EP- 4 (G)
16 the notification the counties would receive from
( PEMA. ,
Q.51 Previously you identified two communication links with the NRC. Please describe these systems in more C detail.
A.51 The two communication systems are the NRC Emergency Notification System (" ENS") and the NRC Health Physics Network Line ("EPN").
The ENS hotline is a dedicated telephone system that connects TMI and all other operating reactors with NRC headquarters in Bethesda, Maryland. It is used to report emergencies. The purpose of this line is to provide reliable notification and communication of operational plant data to the NRC. ENS hotline h :=)=
s phones are located in the ECC (control room and shift supervisor's office), OSC, TSC, and EOF. See Figure 5(f) of this testimony. Initial notification and communication with the NRC is made with the ENS phone in the ECC.. Once NRC representatives arrived in the ECC, they would take over communications on the ENS line. Senior NRC officials reporting to the site can speak with headquarters from the ENS phone at the EOF. The NRC can patch-in the Region I Office on this network.
(
Y7 In the event of a Site or General Emergency, the BPN line will be activated by the NRC operations center
(
in Bethesda, Maryland. This phone is part of a network that includes all nuclear power plants, the g-t NRC regional offices and the NRC operations center in Bethesda. The BPN is a restricted network and is not to be used by non-government employees except to report a significant event when both the ENS.and the k
commercial telephone lines are out of service. This system is dedicated to the transmission of radiolog-ical information by NRC personnel on site to NRC personnel in Bethesda and at the regional office.
HPN phones are located in the ECC (shift supervisor's
- office), the EOF, and the NRC resident site inspector's office. See Figure 5(g) of this
($h testimony.
Q.52 Are there additional means available for co'mmunica-tions among the various emergency response centers?
A.52 Other communication systems include: Emergency Director's auto-dialer phone, the Pennsylvania Bell system, GPU microwave system, TMI radio frequencies, the inter-control room hotline, the Emergency Director's hotline, the plant paging system, the maintenance and instrumentation phone system, and various plant alarms (i.e., radiation emergency, fire 4%
\
and reactor building evacuation). Each of these
( systems is described further in Sections 4.7.5.9 through 4.7.5.18 and Table 18 of the Emergency Plan.
Q.53 In addition to the flow of information across the communication links you have just described, will these communication links also be used'to support the decision making process?
A.53 Yes. There are two primary networks of emergency response decision making.
The first is the protective action network. The Emergency Director receives input and data from the RAC and EAC regarding offsite radiation levels and from the Operations Coordinator regarding plant status. Based on this information, the Emergency kh
- Director will make protective action recommendations to BRP. After receiving the protective action recommendation from the site and reviewing data from its own monitoring teams, BRP determines if protective action is warranted, and, if so, advises PEMA of the action to be taken. PEMA communicates with the Governor, or his designee, and with the Governor's consent, initiates the protective action.
The second network consists of decisions to be made regarding plant operations during an emergency.
Initially, the Emergency Director provides direction
Yi to plant operators responding to an accident. Once r~- the TSC is activated and B&W is contacted, the Emergency Director begins to receive technical recommendations over the Operational Line. When the
([
Parsippany TFC is manned, the Group Leader Technical Support and his staff assume responsibility for providing technical advice on plant operations.
'i . Initiel Accident A;;;;;;;nt Q.54 P1 se describe the basic components of accident ass sment.
A.54 The init al step in accident assessment is awareness of a pr'obl . This determination initiates an 0
investigative rocess intended to define the nature
~
hb of the problem w h sufficient specificity to permit -
an evaluation of pl t status and potential hazards.
Simultaneous with this investigative process, as information is developed, the shift supervisor will 1
implement appropriate respo e procedures. If conditions warrant, the shift pervisor will classify the emergency as an Unus 1 Event, Alert, Site Emergency or General Emergency d implement the j Emergency Plan in accordance with the a ropriate implementing procedure.
I L
~
Q.55 You identified awareness of a problem as the i tial step in accident assessment. Are there differen types of information that have to be monitored an analy wJ-Le p.eperly perform thi: :tep?
CM .) e) ir e C .
Lic 2/9/81 l i
i UNITED STATES OF AMERICA ,
NUCLEAR REGULATORY COMMISSION j BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
,N k
LICENSEE'S TESTIMONY OF ROBERT E. ROGAN, GEORGE J. GIANGI AND ALEXIS TSAGGARIS ON THE ADEQUACY OF ONSITE EMERGENCY PREPAREDNESS AT THREE MILE ISLAND, UNIT 1 Volume 2 -- Figures, Tables and Appendices l
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-.----- Coutur 3 ,*Illi I:seorgency
- 1=
Reginn I i&E --- --- y 'Illi Niiar-selta Control Center King ul Prunnelse. Capitol City g o ranorgency:Opura- ,
i l'A .Alrport p
- N ' tions Facility -
Unit 1 Control
- """"#8""'
C: 'isr osanstnin 9
" E *' " " # *
[' ( Center
'j i
l'ennsy l van in E* Officu linruieu of h
ih rgency Hanage h ment Agency Hnallation h v CPil.Nucleme (entA) Emergency Prutection g t Media
' g Opernt inna Ctr. Conter ilurrtuburg, PA llarrinburg, PA, l Crawford i
Station
~
o Other Cuientlen I.ncut Dauplein County Emernency HeinIc Ipal I t (co' Emergency h,TMI operatlonn Operattunn i*.mu ritiincy Operaations -
l Snpport Center Centern Operationts Center
- Rati f ologica l Cuntur _ _ - _ _ .
Controls llarrtuliierg, PA Acccca contral point U,-
il -
I t;'-
= ce no , wnwcorm roc i Torece . cmdmm i OtCMT l(n_N ,
ENVIRONMENTAL ASSESSMENT LINE o A Parsippany Alternate Till Teclinical
?)
E Emergency support Center uisclunc imbenck & Wileux Technical Corporation Functions '- : y Operations ,
iti nis t utory ,
romai nn lo'n
_ftatar_.,_ _. g _Facil tt _ _ __ _ _ ,_ _ _ __ _
- . - - - - - - ,, e
_ _ _ . _ _ _ _ _ _ . . * >4 ,
Hamitu Sinstalown gpgj rawford Statina koum Nitu .ne t tieniin ,Hin I.ynchbura VA Headquarters k -~
Parsipanny,NJ 8
\
m Environnental Assessment u Command Center Nucleur Fuilural - - . - - - - - - -
r'.aut r gu nc y Olmstead l iugulatory Airport Coinnil uu lon npurntlano g ,,
- - - . - - - - Uniitur q "' # 3'*" "" '
Heglon I i f. E - - - . - - - . - S THE Nour-alto a " '" "" "I
(*npitol City 8 thrgency Oparn-King of prunnin,
-W q Unilt l controt l'A , Airport [
h tienes Facility
""""/8h"'
,:sr os rrvstrans- h l
i
[1 4 k U "E"'" ""' "
v Center i office l'ennuy lvan in
- p.inergency llannge llurnau uf hg ,,
ment Agency unillation . ,,
" CPU-Hucient (plGtA) ihrgency Protection
%dia
Station
. Other Cennition I.ncal . 11auplain County Emergency HunicipalItleo' Emergency Ttt! Operatloan Operatlunn i*.mu rgenc y Operatioqu *
. Support Center CenLern Operatloon ,
Centue e ,g9 g Contur ,
_ , . _ , , , _ ' " ~
Controla Ilarrinburg PA Acccus control l'otnt I
PARSIPPANY/TMI LINE I
e S Alternate Till Teclinical Parsippany q l ,
Emergency suppurt center Hnelear linlicuck & Wileux Technical un corenration functions '- : Operations itenulatory h ~
Facilitt 8
ran==I nn lon _ feat.tr_ _. - gg s
_ . _ _ . _ _ _ _ itsmote Sluitdown I.ynchbura,VA Headquarters GPlj gf v .,,
' :rawford Station koom mtit .ne then.In ,Hn --
Parsipanny,NJ 8 l
- - - -- - Environisiental Assessment s Connand .
Center
. Hnclear Fudoral s'moritancy Olmstead l nenulatory Airport comuul au lon opurnt.inna g ..
- - - . - - - - - contar q
, g g. ,,,,
Hegion I i&E _ . - - - - _ - - y j, 'INI Noar-nito Contrnt Center Kleig of Prunnin, Capitol City p,g o Emergency Opurn-PA , Airport i g tions Facility g Unit'l Control ri s -
<a Hunm/Sist f t
- *1 N
Tsr 05mii?vacTini H Supervinur.s l'ennNlvanin h Office l'.mergency Hanage lluronn of -
ment Agency undlation h CPU-Nuclent (plGtA) Four ..mey Prestection >- Medla operat tone Ctr- Center llarr tuburg, l'A liari 1:iburg, PA, l
l Crawford Station
. Other Cmnitlen 1.ncut . Dauphin County '
Emergency Hunicipalition' Emergency THE Operatloan
- Operattunn i*.mu rgencY Operations - . Siippo r t Center ,
Centern operationn Center ' Rudiningical r.cntur . __ .
liarrtuburg, PA Acccca Cointrol Point m-
- 1 OQ
. -,e c..e nerairv racy Tvcc _. r.rNrDAl iOCATION _ __
PARSIPPANY/B & W LINE .
n Parsippany Alternate THI Teclinical
% Emergency Support Center p,,,. g , a r l'al'C"ek & Wi lcox Technical "
vi u..nu l a t *
- ry .
Corporation Functions '- y Operations Facility, _ _ _
comminalon _f tnt _RT- _. _ ,h
,_i
, Hemoto Sinstilnwn GPU Nitit ni ttsesita, tin I.ynelibu ra.VA lleadquart'ers Parsipanny,NJ b>3h., l 8
- rawford Station koom Environmental Assessment u Command Center Hitc l ea r Futtu ral -..___ _
i:enula tory s'.mo r r.unc y Olmstead l ,
Operatinno Airport.
c...missson g ,,.
Contur g "
Hegion i I&E b THI Huar-nito
"3*"""
'2"'.'. '1 _
i:Ing of Prunisin, Cup!Lul City 3 Emergenycy Opurn- bUnit" f."" 1 Control PA ,Alrport 3 tions Fac!!!Ly g
"""" 'I
- l U U 7tTr' 05milrva tTini~ di T (j Center N """'"""#,
v pi ottgcc Pennsylvanin "
einergency Hannne linrunu of hg ,,
ment Ay,ency Hutilation v CPU-Huclear (P10tA) i:mu rge*nc y l*rotection -
Media Operat lona Ctr. Center Hurrluhurg, PA :larrtulnerg, l'A.
l Crawford i
Station
. Ollier Canintjen I.nca l . Dauphin County Emergency Hintic ipal l t leo' Emergency THI Operationn Ope ra t lani: i;murgency Opernt tones i
. Support Center centern Operationn Center
- g,,gg,,,g gc,g C4nttdr . --~~~~
Cosit roin
. Ilarriuhurg, PA Accer.s control l'ain t 3
~. - mm nce mm _ __
~
A .
NRC EMERGENCY NOTIFICATION SYSTEM (ENS)
,e Parsippany y Alternate 'ml Tecluitcal Emergency -E support center unclear I"'I'C"ek & Wi le"* Technical vi .
Operations u .niilutiiry Corenration Functions '- e ,
Facilitt _ _ _
4 co in.lon fti.gT_t -.
,o .
_ g
- " * *" E " 8 3 "' E 'I"""
- GPU koom I.ynclihura,v4 lleadquarters 3>k
- rawford Station mut. net heniin ,Hn -
Parsipanny,NJ '\
Environmental Assessment u i
Command ,
Center Hui-l ea r Feelural sh runney Olmstead l ner.u t a to ry Airport co laulon opurat. tono g .
Conter *Dit l'.seatrr.csicy negion i I&E -------- p THI Hear-nito M r Cuntrol Center King of Prunnin, Unp!Lui City [,l !? F.mergency Opura-PA , Airport l g '
tions Facility ; Unit'l Control l
i I El y ,nTr ossarvitunr i j
""""/8h"',s Supervisor
( Cmu ,
i Officu Pennsyivanin 'q '
ihrgency Itaisage llaironti of ,
h ,,
ent Agency Hisillat ion j v CPU-Nuclear (l'iGtA) innergiape y l'rutoct ion g;
l Crawford Station
. Ottier Cennit jen I.ncal . Dauplain County Emergency Hanicipalltleo' Emergency THI Operatloan operattuun i*.mu rnanc y Operationii -
g Siipport Center
. (entern Upurationn Center * ,g,,,g,g,g ge ,g Centur _. _ '
controla liarrtulairg, PA Accer.s Contral Paint ,
c 1
o
. . _ . . . _.._____._.. ........, ,. . . . r n a . .neartnu _
NRC llEALTil PilYSICS NETWORK LINE (HPN) -
n Parsippany y Alternate Titt Technical Hoclear I"il' rock & Nilen* Technical Oi Emergency suppurt Center Corporation Functions *- : I Operations '
tienul u tor v C u== I nn i nn 0gatff_ _. _
,. h _FgihtL _ _ _ _______
g .
llaanto Sinitilown '
GPU Nitit .ite t tieniin ,Hin I.ynchburg,VA Headquarters vk ;rawford Stutlim hoom Parsipanny,NJ '\ --
- - Envice.wental Assessment a NRC Resident Command M Site Inspector's Lnoelear Center Office l'citu ral - _ . _ . _ _ .
- enulatory swirituncy Olmstead l opuratinno n Airport roinnil uu t na
- - - - - - - - Contur *}
I&E in THI Hoar-nite 3 lleginn i King of Prunnin, Cupitol City 8 Laurgency Opurn- *"",'."""U"""'
b $"N '*1 _
l'A , Airport g ]
tions Facility g Unit f Control Hnom/Sli!f t D
F b!
,gg os n,acu,n -
N H Supervisor's
( Center g gggg l'ennd i van in 'f'I "
usergency liannge nien t Agency thirunn of unillation $
}h v
cru Huctant (l'iGtA) Emergency Prntoction p Media i Center l
opernt Inna Ctr.
llarrtuburg, PA !!arrinburg, PA, ~
l Crawford Station o other Cnuntien 1.oca l . Dauphin County i*nie rnency Hnnicipalltleo' Emergency T!11 Operatloan Operatlunn i'.auirgons.y Operncions
- i
. Support Center-rentern Opurationn Centur
- g,,gog, ge,g Coutur -
g ,g, llarrtuliurg, PA Accces Contral Pn tiit o-m
. -i r..enreurv r a c'r e t Tt_rc crurom 9OrATinN
61 i
TABLE 2
( '
Number On-Shift Assignment Emergency Assignment 1 Shift Supervisor Emer'gency Director 1 shift Technical Advisor Performs normal functions l 1 Shift Foreman Performs n,ormal functions 1 Control Room Operator il Operates primary' plant 1 0 :t :1 T.::: 0 :::t:: f2 0- zuri :t :
1 Switching & Tagging Operates secondary plant Control Room Operator .
I HF Auxiliary Operators Radiological monitoring teams, fire brigade, emergency repair, plant, operations, c..wn.c.& m Radiological Controls Radiological Assessment
() 1 Foreman Coordinator 1 Senior Radiological Radiological Controls Controls Technician Coordinator 2 Radiological Controls In-plant radiological Technicians controls (assess control surveys, etc.), radio-logical monitoring teams 1 Senior Chemistry Chemistry Coordinator
. Technician 1 Shift Maintenance Foreman Operations Support Center Coordinator 4 Maintenance Personnel Emergency repair, search and rescue, radiological monitoring team drivers
_ _ _