ML19296A815
| ML19296A815 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/25/1980 |
| From: | Glasspiegel H, Keaton R METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | UNION OF CONCERNED SCIENTISTS |
| Shared Package | |
| ML19296A811 | List: |
| References | |
| NUDOCS 8002190078 | |
| Download: ML19296A815 (68) | |
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January 25, 1980 UNITED STATES OF AMERICA NUCLEAR '1EGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
LICENSEE'S ANSWERS TO A PORTION OF UNION OF CONCERNED SCIENTISTS' (UCS)
INTERROG, TORIES DATED JANUARY 4, 1980 Introductory Comments UCS's set of interrogatories dated Canuary 4, 1980, requests that with respect to each contention Licensee first answer five preliminary questions. The first four of these five questions relate to Licensee's position on the contention.
Licensee's position on the contentions will not be finally developed until testimony has been drafted, reviewed and finalized by Licensee and its contractors. Consequently, answers to these four preliminary questions are not available at this time. However, UCS will find that Licensee's basic disagreement with each contention is reflected in answers to other interrogatories pertaining to the contention.
UCS also requests with respect to each interrogatory the identity of and other infonmation concerning the expert (s) 8002190 o78<.:
sv * -
whom Licensee intends to have testify on the subject matter of the interrogatory. No such expert witnesses have been identi-fied by Licensee at this time. When expert witnesses have been identified, Licensee will so notify UCS.
Unless otherwise indicated below, further research or work engaged in or to be engaged in by or for the Licensee which may bear on the issues covered in these interrogatories is identified in the Licensee's Restart Report and the Staff's
" Status Report on the Evaluation of Licensee's Compliance with the NRC Order dated August 9, 1979", dated January 11, 1980.
Interrogatorv No.'l Explain the present Licensee position on UCS Contention No. 1.
Response
See introductory comments.
Interrogatory No. 2 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 3 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 1. Explain the reasons for which any such person dissents.
g . -
Response
See introductory comments.
Interrogatory No. 4 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention No. 1.
Response
See introductory comments.
Interrogatory No. 5 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Contention No. 1.
Response
FSAR and Restart Report references for UCS Contention No. 1 are given below:
FSAR Sections 14.1.2.6 Natural Circulation 9.5 Decay Heat Removal 6.1 ECCS Restart Report Sections 3.1.1 Emergency Procedures 3.1.4 Operating Procedures 6.0 Operator Training 8.3.5 Loss of Coolant Flow
- 8. 4 (g) Natural Circulation Test Supplement 1, Part 1, Responses to Questions 20, 24, 36, 37, 45 and 51 Supplement 1, Part 2, Responses to Questions 94 and 95 Supplement 1, Part 3, Responses to Questions 1, 2 and 3 Interrogatory No. 6 Explain whether or not natural circulation is an adequate means for removing decay heat from the reactor core in the event of a small loss-of-coolant accident ("LOCA").
s . .
Response
(A) Natural circulation occurs during most small break LOCA's when the reactor coolant pumps are not oper-ating and is an adequate means for removing decay heat. By natural circulation Licensee means the circulation of liquid water through the primary system. If the core is covered by a two-phase mix-ture of fluid, adequate core cooling is maintained even without natural circulation.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks fo'r B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
- 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC),
May 1, 1978.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
i . ,
Interrogatory No. 7 Explain in detail which of the short or long term measures recommended by the Director of Nuclear Reactor Regulation and to be implemented by the Licensee will prevent the formation of voids in the reactor cooling system as occurred in TMI-2.
Response
(A) The implementation of the short and long tern measures recommended by the Director of Nuclear Reactor Regulation will not prevent the forma-tion of voids in the reactor cooling system as occurred in TMI-2, although a number of meast.res will reduce the likelihood of void formation.
See references in Response to Interrogatory No. 5.
(B) None.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 8 Does the Licensee take the position that implementation of the short term and/or long term measures identified by the Staff will constitute conformance with the requirements of GDC 34? If yes, identify the specific measures and explain how their implementation will meet the requirements of GDC 34.
Response
(A) TMI-l was designed, constructed and licensed in accordance with the General Design Criteria listed in the 10 C.F.R. 50 Appendix A, as issued on July 11, 1967. The FSAR describes conpliance to these criteria. The short term and/or long term
i . ,
measures identified by the Staff to be implemented by the Licensee do not in themselves constitute conformance with the requirements of GDC 34.
(B) 1. Final Safety Analysis Report, Three Mile Island - Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 9 If the answer to Question # 8 above is "no", what additional measures will be taken by the Licensee to achieve conformance with GDC 34?
Response
(A) None.
(B) None.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 10 What is the schedule for implementation of the additional requirements, if any, described in question #9 above?
Response
See response to Interrogatory No. 9.
Interrogatory No. 11 Does the Licensee take the position that the short term and/or long term measures identified by the Staff to be imple-
.aented by the Licensee will constitute conformance with the
requirements of Gnc 35? If yes, identify the specific measuric and explain how t air implementation will meet the requirements of GDC 35.
Response
(A) TMI-l was designed, constructed and licensed in ac-cordance with General Design Criteria listed in 10 C.F.R. 50 Appendix A, as issued on July 11, 1967. The FSAR describes compliance with these criteria. The short term and/or long term measures identified by the Staff to be implemented by the Licensee do not in themselves constitute con-formance with the requirements of GDC 35.
(B) 1. Final Safety Analysis Report, Three Mile Island Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory c'amments.
Interrogatory No. 12 If the answer to question #11 above is "no", what addi-tional measures will be taken by the Licensee to achieve cen-formance with GDC 35?
Response
(A) None.
(B) None.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
a .
Interrogatory No. 13 What is the schedule for the implementation of the addi-tional requirements, if any, described in question #12 above?
Response
See response to Interrogatory No. 12.
Interrogatory No. 14 Explain the present Licensee position on UCS Contention No. 2.
Response
See introductory comments.
Intorrogatory No. 15 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing (s) , explain the prior position, and explain the basis for the change in position.
Response
See introdactory comments.
Interrogatory No. 16 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 2. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 17 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Conten-tion No. 2.
Response
See introductory comments.
Interrogatory No. 18 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 2.
Response
FSAR and Restart Report references for UCS Contention No.
2 are given below:
FSAR Sections 4.2.2.5 & 6 Reactor Coolant Pumps 6.1 ECCS 9.5 Decay Heat Removal 11.2.1 Radioactive Liquid Waste Disposal System Supplement 2, Part IX, Section 5.4 Restart Report Sections 2.1.1.8 Leak Reduction Program for Systems Outside Containment 2.1.2.3 Plant Shielding Review 7.3.1 Radwaste Capability 7.3.2 Plant Shielding Also see the references given for Interrogatories 5 and 31.
Interrogatory No. 19 Explain how the Licensee will provide forced cooling flow to the reactor during small LOCA's.
Response
(A) For :ed cooling flow by the reactor coolant pumps to the reactor during small LOCA's is not necessary to maintain adequate core cooling and mitigate the consequences of the accident. As necessary, forced cooling flow is provided by ECCS. See ulso Response to Interrogatory No. 6.
i .
m (B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", J1ptember, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 253F-MWt Interncis Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
- 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC),
May 1, 1978.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 20 Explain how any and all proposed methods for providing "en-hanced" cooling to the core after a small LOCA will comply with all applicable NRC safety regulations and are sufficiently re-liable to safeguard the public health and safety.
Response
(A) See references identified in "B".
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evalu' tion Model.",
August, 1977.
- 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC) ,
May 1, 1978.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 21 If the Licensee plans to rely on the reactor coolant pumps, explain how the formation of voids will be prevented.
Response
(A) The reactor coolant pumps are not relied upon to mitigate the consequences of a small break loss-of-coolant accident.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evalue 77 Model",
August, 1977.
- 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC),
May 1, 1978.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 22 Explain how the reactor coolant pumps will comply with GDC 17 (on-site power supply).
Response
(.A ) The reactor coolant pumps do not have an onsite power supply.
(B) 1. Final Safety Analysis Report, Three Mile Island - Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 23 Explain how the reactor coolant pumps will comply with IEEE 279 (10 C.F.R. Part 50.55a(h) - controls).
Response
(A) The reactor coolant pumps meet the requirement of IEEE-279 for nonsafety-related equipment, but not the requirements for safety-related equipment.
(B) 1. Final Safety Analysis Report - Three Mile Island - Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 24 Explain how the reactor coolant pumps will comply with GDC's 2 and 4 (seismic and environmental qualifications).
Response
(A) The reactor coolant pumps comply with GDC's 2 and 4 to the extent required for protection of the reactor coolant pressure boundary.
(B) 1. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 2. .:.0 C. F. R. 50 Appendix A.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatorv No. 25 If the Licensee plans to rely on the Emergency Core Cool-ing System (ECCS) in a " bleed and feed" mode, explain how there will be sufficient capacity and radiation shielding for the storage of the radioactive water bled from the primary coolant system.
Response
The Licensee does not understand the term " bleed and feed" in the ccntext of this question. For the cooling methods used for loss-of-coolant accidents, see Response, including refer-ences, to Interrogatory No. 19.
Interrogatorv No. 26 Explain the present Licensee position on UCS Contention No.
3.
Response
See introductory comments.
Interrocatory No. 27 Does the current position differ from the position of the Licensee in ar.y prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response, See introductory comments.
Interrogatory No. 28 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 3. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 29 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are re-lied upon in formulating the Licensee position on UCS Contention No. 3.
Response
See introductory comments.
Interrogatory No. 30 4
Ident.fy all sections and page numbers of the FSAR and Re-start Repcut which contain subject matter pertaining.to UCS Con-tention No. 3.
Response
FSAR and Restart Report references for UCS Contention No. 3 are given below:
FSAR Sections 4.2.2.3 & Pressurizer and Heaters 4.2.4.4 8.2.3 On-site Emergency Power Restart Report Sections 2.1.1.3.1 Emergency Power Supply for Pressurizer Heaters Supplement 1, Part 1, Responses to Questions 11 and 14 Supplement 1, Part 2, Responses to Questions 18 and 30 Interrogatory No. 31 The staff has recognized that the " maintenance of natural circulation capability is important to safety (and) depends on the maintenance of pressure control . . . (which) is normally achieved through the use of pressurizer heaters." NUREG-0578,
- p. A-2.
(a) Do you agree?
(b) Explain why pressurizer heaters and their associated controls are not classified as " components important to safety," as discussed in GDC-17 and the Introduc-tion to Appendix A, to CFR Part 50.
Response
(A) (a) Licensee agrees only to the extent that mainten-
\
ance of natural circulation capability (as de-fined in Response to Interrogatory 6) is the preferred cooling mode during certain plant con-ditions and depends on the maintenance of pres-sure control which is normally achieved through the use of pressurizer heaters.
(b) The pressurizer heaters and their associated con-trols are not essential to mitigate.the conse-quences of accidents, and, therefore, are not classified as " components important to safety".
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 32 Explain in detail whether and in what manner the following design criteria would be met with respect to the pressurizer heater and its associated controls.
(a) GDC 22 (diversity)
(b) GDC 2 and 4 (seismic and environmental qualifications)
(c) GDC 20 (automatic initiation)
(d) GDC 3 and 22 (senaration and independence)
(e) GDC 1 (quality assurance)
(f) GDC 17 (adequate, reliable on-site power supply) and (g) the single failure criterion
b
Response
(A) The pressurizer heaters conform to the criteria for reactor coolant pressure boundary components and, as such, meet applicable requirements of GDC-1, 2 and 4. They also meet the requirements for non safety-related equipment imposed by the above GDC, but not the requirements for safety-related components.
(B) 1. Final Safety Analysis Report, Three Mile Island - Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 33 Explain the present Licensee position on UCS Contention No. 4.
Response
See introductory comments.
Interrogatory No. 34 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in' position.
Response
See introductory comments.
Interrogatory No. 35 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 4. Explain the reasons for which any such person dissents.
Resconse See introductory comments.
Interrogatorv No. 36 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are re-lied upon in formulating the Licensee position on UCS Contention No. 4.
Response
See introductory comments.
Interrogatory No. 37 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 4.
Response
See Response to Interrogatory No. 30.
Interrogatory No. 40 Explain the present Licensee position on UCS Contention No. 5.
Response
See introductory comments.
Interrogatory No. 41 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing (s) , explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 42 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 5. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 43 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are re-lied upon in ~ emulating the Licensee position on UCS Contention No. 5.
Response
See introductory comments.
Interrogatory No. 44 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 5.
Response
FSAR and Restart Report references for UCS Contention 5 are given below:
FSAR Sections 4.2.4.2 PORV (Electromatic) 7.3.2.2.3 Reactor Coolant Pressure Control 14.2.2.3 LOCA Restart Report Sections 2.1.1.2 Position indication for PORV and Safety Valves 2.1.1.3 Emergency Power Supply Requirements for PORV Supplement 1, Part 1, Response to Questions 13 and 16 Supplement 1, Part 2, Response to Questions 19, 20, 30, 36 and 37
Interrogatory No. 45 Does the Licensee agree that proper operation of power operated relief valves (PORV 's ) , associated block valves and the instruments and controls for these valves is essential to mitigate the consequences of accidents? Explain your response fully.
Response
(A) The proper operation of power operated relief valves (PORV's), associated block valves and the instruments and controls for these valves are not essential to mitigate the consequences of accidents. The accident analyses presented in the Final Safety Analysis Report do not rely on them to operate.
(B) Final Safety Analysis Report, Three Mile Island ~
- Unit 1.
(C) Mone.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 46 Does the Licensee agree that failure of these valves, instruments and controls can cause o,r aggravate a LOCA? Ex-plain your response fully.
Response
(A) The Licensee agrees that failure of these valves or the instruments and controls which operate these valves can cause a LOCA. The spectrum of small break loss-of-coolant accidents analyzed in the referenced material envelopes a break which
would include failure of these valves or their controls. Therefore, failure of these valves, instruments and controls cannot aggravate the consequences of a LOCA beyond that submitted in the referenced analysis.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2168-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
a 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC), May 1, 1978.
(C) None.
(D) Letter D. F. Ross to J. H. Taylor dated August 9, 1979.
(E) See introductory comments.
Interrogatory No. 47 Provide the justification for the failure to classify PORV's and associated block valves and their respective instruments and controls as " components important to safety", requiring compliance with safety-grade design criteria.
4
Response
See response to Interrogatories No. 45 and No. 46; see also response to Interrogatory No. 55.
Interrogatory No. 50 Explain the present Licensee position on UCS Contention No.
6.
Response
See introductory comments.
Interrogatory No. 51 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 52 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 6. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 53 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Conten-tion No. 6.
Response
See introductory comments.
Interrogatory No. 54 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 6.
Response
FSAR and Restart Report references for UCS Contention No.
6 are given below:
FSAR Sections 4.2.4.1 Pressurizer Code Safety Valves 4.2.2.3 Pressurizer 4.1.3.4 Codes and Classification - Relief Valves Supplement 2, Part VI 8.4 Quality Control Restart Report Sections 10.3.2' Safety and Relief Valve Testing Interrogatory No. 55 Describe in detail the methods by which the reactor coolant system relief and safety valves have been qualified to verify the capability of these valves to function during normal, transient and accident conditions. This description should in-clude specification of the environmental conditions assumed dur-ing normal, transient and accident situations and the means by which these environmental conditions were derived. Provide references to the Regulatory Guides applied in this analysis.
Response
(A) The pressurizer safety valves function to maintain the reactor coolant system pressure s.ithin the al-lowable design limits. The valves were qualified for the pressurizer application by a comSination of design, analysis and testing. Valve capacity was certified to existing code requirements for safety valves. This procedure utilized model test-ing to establish discharge factors to certify
the capacity of the full size valve. The ability of the valve to withstand the design and operational pressures was verified by a seismic report. Hydro-static testing, hot performance testing, and leak-age testing were conducted at the manufacturer's plant prior to equipment shipment. The hot per-formance testing consisted of verifying the valve set pressure to be within the required limits by successively lifting the valve with full system pressure. Prior to reactor operation, preopera-tional testing was performed to verify valve set-point. -
The pressurizer PORV was qualified for the pressurizer application by a combination of the design, analysis and testing. The ability of the valve to withstand the design and operational pressures was verified by a seismic report. Hydrostatic testing, hot perform-ance testing and leakage testing were conducted at the manufacturer's plant prior to shipment. Prior to reactor operation, preoperatior testing was per-formed with the system at operating pressure to en-sure proper valve function.
No external environmental conditions were specified in the procurement specifications and no Regulatory Guides were applied at the time of hardware procurement.
(B) 1. Safety Valves
- a. ASME Boiler & Pressure Vessel Code,Section III, Article 9, N-914.G.
- b. Dresser Industries, Seismic Design Analysis for 31739, dated 2/6/70. Pre-pared by C. K. Brewer.
- c. Q. A. Data Package - B&W I.D. No. 23-0593-00.
- d. Overpressure Protection Report BAW-10043, dated May, 1972.
- 2. Relief Valve
- a. Dresser Industries, Seismic Design Analysis for 31533VX, Date 1/17/69 prepared by D. C. Eaton.
. b. Q. A. Data Package - B&W I.D. No. 23-0594-00.
(C) Final Safety Analysis Report, Three Mile Island -
Unit 1.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 56 Did the Licensee fully apply the analysis of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2, to determine the expected valve operating conditions? If not, provide the justification for failing to do so.
Response
(A) Section 8.0 of the TMI-l Restart Report discusses the accidents and transients specified in Regulatory Guide
h 1.70, Revision 2. Regulatory Guide 1.70 was re-leased after issuance of the TMI-l Operating License.
(B) 1. Regulatory Guide 1.70, Revision 2, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", September, 1975.
- 2. Final Safety Analysis Report, Three Mile Island - Unit 1.
- 3. Restart Report 8.0 (C) None.
(D) See introductory comments.
,E) See introductory comments.
Interrogatory No. 57 Explain how the Licensee chose tne single failures applied to these analyses so as to maximize the dynamic forces on the safety and relief valves.
Response
(A) Failures that would iroduce maximum pressurizer safety valve flowrate were selected for the over-pressure protection analysis.
(B) 1. ASME Boiler and Pressure Vessel Code,Section III, Article 9.
1967 Summer Addenda 1968 Winter Addenda
- 2. BAW-1004.', " Overpressure Protection for Babcock and Wilcox Pressurized Water Reactors", May 1972.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 58 Explain how the test pressures utilized in these analyses were determined to be the highest pressures predicted by con-ventional safety analysis procedures.
Response
The licensee does not understand the term " test pressures utilized in these analyses". The overpressure protection analysis is identified in the Response to Interrogatory No. 57.
Interrogatory No. 59 How did the Licensee determine the test conditions for qualification of the control circuitry, piping and supports associated with the reactor coolant system relief and safety valves?
Response
(A) 1. Control Circuitry - Since operation of the PORV is not a safety function, safety-grade qualification of its control circuitry is not required. Test conditions were deter-mined in accordance with the functional re-quirements of the equipment. No control circuitry is associated with the pressurizer code safety valves.
- 2. Piping - The test conditions for qualifica-tion of the piping associated with the reactor coolant system relief and safety valves were determined in accordance with the requirements of USAS B31.7, Code for Pressure Piping, Nuclear Power Piping, dated February 1968, and as corrected for Errata under date of June 1968.
- 3. Supports - The pressurizer provides support for the reactor coolant system relief and safety valves. Qualification is by ar.alysis.
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1.
GAI Drawing E-304-653.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 60 Explain how the qualification testing of the reactor coolant system relief and safety valves and associated control circuitry, piping and supports complies with GDC 1, 14, 15 and 30.
Response
(A) The TMI-l plant was designed, constructed and licensed to the original design criteria ccn-tained in 10 C.F.R. 50 Appendix A, as issued on July 11, 1967. The compliance with these criteria is demonstrated in the TMI-l FSAR.
(B) 1. Safety Valves
- a. ASME Boiler & Pressure Vessel Code,Section III, Article 9, N-914.6.
- b. Dresser Industries, Seismic Design Analysis for 31739, dated 2/6/70.
Prepared by C. K. Brewer.
- c. Q. A. Data Package - B&W I.D. No.
23-0593-00.
- d. Overpressure Protection Report BAW-10043, dated May, 1972.
- 2. Relief Valve
- a. Dresser Industries, Seismic Design Analy-sis for 31533VX, Date 1/17/69 prepared by D. C. Eaton.
- b. Q. A. Data Package - B&W I.D. No.
23-0594-00.
(C) Final Safety Analysis Report, Three Mile Island -
Unit 1.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 61 Explain the present Licensee position on UCS Contention No. 7.
Response
See introductory comments. ,
Interrogatory No. 62 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing (s) , explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 63 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention No. 7. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 64 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention No. 7.
Response
See introductory comments.
Interrogatory No. 65 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 7.
Response
FSAR and Restart Report references for UCS Contention No. 7 are given below:
FSAR Sections 1.4.12 GDC 12 (Comparable to GDC 13 today) 7.3.2 Non-nuclear Process Inst.
2.1.1.3.4 Pressurizer Level Instrument Power Supply 2.1.1.6 Instrumentation to Detect Inadequate Core Cooling 3.0 Procedural Modifications 6.0 operator Training Supplement 1, Part 1, Responses to Questions 17, 18, 19 and 20 Supplement 1, Part 2, Responses to Questions 30, 92, 93 and 95
Interrogatory No. 66 Would a direct measurement of the reactor coolant level be of assistance to the reactor operator in determining the most appropriate remedial actions during a small break LOCA?
Response
(A) Plant instrumentation and operating procedures at TMI-l will permit prompt recognition of a LOCA and specify appropriate remedial actions. Cor-rectly specified level measurements may be of assistance to the operator it properly used in conjunction with other appropriate information.
(B) B&W Report, No. 69-1106001-00, "Small Break Operating Guidelines", November, 1979 (C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 67 Explain how present procedures and instrumentation permit prompt recognition of low reactor coolant level and inadequate core cooling.
Response
(A) The TMI-l procedures and instrumentation, using information presented in the reference identified in Part (B) below, will permit prompt recognition of low reactor coolant level and inadequate core cooling.
(B) B&W Document, 69-1106001-00, "Small Break Operating Guidelines", November, 1979.
h (C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 68 Would any of the short and/or long term measures recommended by the Staff to be implemented by the Licensee provide instrumen-tation to directly measure the water level in the fuel assemblies?
Explain your answer fully,
Response
(A) No. The short term and/or long term measures recommended by the Staff to be implemented by the Licensee do not provide for instrumentation to directly measure the water level in the fuel assemblies.
(B) 1. NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations."
- 2. Letter H. R. Denton to all Operating Nuclear Power Plants, " Discussion of Lessons Learned Short Term Requirements", October, 1979.
- 3. NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report".
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 69 If the answer to question #68 is "no", explain the basis for concluding that the provisions of IEEE 279, S4.8, as incorporated in 10 CFR 50.55a(h) have been complied with.
Response
(A) The safety analyses for TMI-l reported in the FSAR do not require a water level in the fuel assemblies as a signal input to the protection systems. Therefore, IEEE 279 does not apply.
(B) 1. IEEE Standards 279, " Criteria for Protec-tion Systems for Nuclear Power Generating Status."
- 2. 10 CFR 50.55a(h).
- 3. Final Safety Analysis Report, Three Mila Island - Unit 1.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 70 What short term modifications of existing procedures and/or instruments have been made at TMJ-l for monitoring water level in the fuel assemblies?
Response
(A) See Responses to Interrogatories 67 and 72.
(B) None.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 71 How do any such new procedures and/or instruments differ from those in p. ace prior to the accident at TMI-2?
Response
See responses to Interrogatories 67 and 72.
Interrogatory No. 72 Explain how any modifications proposed by the St2.ff or Licensee will provide more direct measurement of reactor cool-ing level and inadequate core cooling. What is the implementa-tion schedule for any proposed modification?
Response
(A) Modifications have been proposed by the Staff or Licensee which will provide more direct measure-ment of reactor cooling level and inadequate core cooling, These modifications are as follows:
sat /T sat
- 1. Implementation and use of a P meter to indicate Reactor Coolant System subcooling margin.
- 2. An expanded range of the hot leg RTD's.
- 3. The incore thermocouples will provide the operator an indication of inadequate core cooling.
These modifications are used in conjunction with the Inadequate Core Cooling Guidelines to allow the operator to recognize and recover from a situation of inadequate core cooling (see reference material).
The implementation schedule for these proposed modifications calls for installation prior to the restart of TMI-1.
(B) B&W Document, 69-1106001-00, "Small Break Operating Guidelines", November, 1979; Restart Report.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 73 Discuss how the reliability of information from any pro-posed instrumentation compares with the reliability of the direct measurement of the reactor coolant level.
Response
(A) No analysis has been performed of the reliability of the direct measurement of the reactor coolant level.
(B) None.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 74 Describe the training program to inform reactor operators of new procedures. Document the number of hours of instruction on these new procedures.
Response
Normally, the Supervisor of Operations places all new or revised procedures into the Revision Review Book in the Control Room. Each individual on shift is required to review the Revi-sion Review Book prior to assuming his duties on shift.
In accordance with TMI Unit i requalification program, all licensed personnel are trained in procedure changes on an annual cycle.
The Supervisor of Operator Training Schedules lectures on all new or revised procedures in the next Operator Training Period after promulgation of the procedure change.
As part of the current Operator Accelerated Retraining Program, which is being conducted prior to the restart of Unit 1, comprehensive procedures review program will be conducted.
In this program, each operator will have received the following training:
Operating and Emergency 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Procedure Review Control Room Review in which 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> procedures are used Procedures Update Review 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Interrogatory No. 75 Explain the present Licensee position on UCS Contention No. 8.
Response
See introductory comments.
Interrogatory No. 76 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing (s) , explain the prior position, and explain the basis for the change in pcsition.
Response
See introductory comments.
Interrogatory No. 77 Identify any officers or employces of, or consultants to, the Licensee who dissent from the present Licensee position on
37 -
UCS Contention No. 8. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 78 Identify the specific sof ions and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention No. 8.
Response
See introductory comments.
Interrogatory No. 79 Identify all sections and page numbers of the FSAR and Re-start Report which contain subject matter pertaining to UCS Con-tention No. 8.
Response
FSAR and Restart Report referelices for UCS Contention 8 are given below:
FSAR Sections 14.2.2.3 Supplement 2, Part X Restart Report Sections 8.3.15 Small Break LOCA Interrogatory No. 80 Discuss in detail what analysis has been performed for a spectrum of small break locations which demonstrates that the specific parameters of 10 CFR 50.46 will not be exceeded, with particular attention to peak cladding temperature (50. 46 (b) (1) )
and hydrogen formation (50. 46 (h) (3) ) .
Response
(A) An analysis has been performed for a spectrum of small break locations which demonstrates that the specific parameters of 10 CFR 50.46 will not be exceeded. This analysis is documented in the reference material.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island - Utsit 1.
- 3. BAW-10064, "Multinode Analysis of Core Flood-ing Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
- 6. Letter, J. H. Taylor (B&W) to R. L. Baer (NRC),
May 1, 1978.
(C) None.
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 81 Discuss in detail how the analysis described in question
Response
(A) If the term "SER" described in the Interrogatory refers specifically to the Safety Evaluation Re-parc written for the TMI-l FSAR, References 1, 2, and 3 in Part B to this response were included.
References 4, 5, and 6 were submitted subsequent to the SER, and the analyses which differ are included therein.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island -
Unit I.
- 3. BAW-10064, "Multinode Analysis of Core Flooding Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
6 Letter, J.H. Taylor (B&W) to R.L. Baer (NRC),
May 1, 1978.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 82 What steps have the Licensee and the staff taken to prevent fuel cladding temperature from exceeding 2200* F.
and to prevent the reaction of more than 1% of the cladding with water or steam to produce hydrogen in the event of a small LOCA?
Response
(A) The analyses performed for the TMI-l system design demonstrate that with operation of ECCS the fuel cladding temperature will not exceed 2200* F. and that no more than 1% of the cladding will react with water or steam to produce hydrogen in the event of a small LOCA.
(B) 1. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568-MWt Nuclear Plants", September, 1972.
- 2. Final Safety Analysis Report, Three Mile Island -
Unit I.
- 3. BAW-10064, "Multinode Analysis of Core Flooding Line Break for B&W's 2568-MWt Internals Vent Valve Plants", April, 1973.
- 4. BAW-10104, Rev. 3, "B&W's Evaluation Model",
August, 1977.
- 6. Letter, J.H. Taylor (B&W) to R.L. Baer (NRC),
May 1, 1978.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatorv No. 83 Explain the present Licensee position on UCS Contention 9.
Response
See introductory comments.
Interrogatory No. 84 ,
Does the current positior differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
_ Interrogatory No. 85 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention 9. Explain the reasons for which any such person dissented.
Response .
See introductory comments.
Interrogatory No. 86 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention 9.
Response
See introductory comments.
Interrogatory No. 87 Identify all sections and page numbers of the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 9.
Response
FSAR and Restart Report references for UCS Contention 9 are given below:
FSAR Sections 7.1.2.2.4 Availability of Information 7.1.3.2.4 Status of Safeguards Equipment and Status of Reactor Building Isolation and Cooling 7.1.3.3.6 Bypassing Restart Report Sections 2.1.1.5 Containment Icolation Modification Supplement 1, Part 1, Response to Questions 26, 54, 58 and 60 Supplement 1, Part 2, Response to Questions 22 and 29 Interrogatory No. 88 Does TMI-l comply fully with the provisions of Regula-tory Guide 1.47? If not, explain how the design of TMI-l differs from the requirements of the Regulatory Guide.
Response
(A) The design of TMI-l differs from the provisions of Regulatory Guide 1.47, in that TMI-l does not have an automatic system to indicate bypass or inoperability at the system level as indicated in Regulatory Guide Position C1. Instead, indica-tion of bypass or inoperability is provided at selected component level.
(B) Regulatory Guide 1.47 (1973)
FSAR Section 7 Safety Evaluation, Docket No. 50-289 (July ll, 1973)
(C) None (D) See introductory comments.
(E) See introductory comments.
h Interrogatory No. 89 Is it the position of the Licensee that TMI-l can be operated with adequate protection for the public health and safety without complying fully with Regulatory Guide 1.477 Explain your answer fully.
Response
(A) It is the position of the Licensee that TMI-l can be operated with adequate protection for the pub-lic health and safety without complying fully with Regulatory Guide 1.47.
Regulatory Guide 1.47 (May 1973) provides an acceptable method for implementing the requirements of Section 4.13 of IEEE Std. 279-1971 and Criterion XIV of Appendix B to 10 C.F.R. Part 50.
The design of the TMI-l reactor protection and engineered safety features systems also meets the requirements of Section 4.13 of IEEE Std. 279-1971 and Criterion XIV of Appendix B to 10 C.F.R. 50.
Refer to the Safety Evaluation, Docket 50-289 dated July 11, 1973, page 7-2.
(B) Regulatory Guide 1.47 FSAR Section 7 Safety Evaluation Docket No. 50-289 (July 11, 1973)
(C) None (D) See introductory comments.
(E) See introductory comments.
h Interrogatorv Mo. 90 Explain how the reactor operator will be informed that a safety system has been deliberately disabled.
Response
Administrative procedures require that log entries and backup componenc testing be accomplished prior to taking any safety system out of service. These logs are reviewed by on-coming shift personnel before assuming shift responsi-bilities. In addition administrative procedures require that a Shift Foreman approve Switching and Tagging orders that re-move equipment from service and that tags or stickers be placed on the components taken out of service. Hence, operat-ing personnel would be alerted to equipment out of service by observation of these tags or stickers and by their log review prior to assuming the shift.
In addition to the above requirements, once each shift, control room indications that verify operational readiness of safety related equipment are required to be completed on a Safety Equipment Readiness Check list.
This check list is signed and reviewed by the on-coming Control Room Operators and licensed Supervisory personnel.
h Interrogatory No. 91 Explain the present Licensee position on UCS Contention 10.
Response
See introductory comments.
Interrogatory No. 92 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductor comments.
Interrogatory No. 93 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention 10. Explain the reasons for which any such per-son dissented.
Response
See introductory comments.
Interrogatory No. 94 Identify the specific sections and page numbers of the FSAR for TMI, Unit, 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Con-tention 10.
Response
See introductory comments.
Interrogatory No. 95 Identify all sections and page numbers of the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 10.
Response
FSAR and Restart Report references for UCS Contention 10 are given below:
FSAR Sections 7.1 Protection Systems 7.1.3 ES Actuation System 5.3 Isolation System (Containment)
Restart Report Sections 2.1.1.5 Containment Isolation Modifications 3.1.1 Emergency Procedures Supplement 1, Part 1, Response to Questions 26, 56 and 59 Supplement 1, Part 2, Response to Question 24 Interrogatory No. 96 Is it the position of the Licensee that " protection system,"
as referred to in Sec. 4.16 of IEEE 279 does not refer to high-pressure ECCS, low-pressure ECCS, containment isolation, emer-gency power or other prescribed safety functions? Explain your answer fully.
Response
(A) It is the position of the Licensee that " protection systems" as referred to in Sec. 4.16 of IEEE 279 refers only to the actuation portions of the high-pressure ECCS, low-pressure ECCS and containment iso-lation. The scope of the " protection system" as re-ferred to in Sec. 4.16 of IEEE 279-1968 is defined in Sec. 1 of that document to encompass "all electric and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generat-ing those signals associated with the protection func-ion". The " protection system"sreferred to includes
the Engineered Safeguards Actuation System (ESAS) and those devices and circuitry up to and including the actuation device input terminals in the motor control center for the high pressure ECCS, low pressure ECCS and containment isolation system valves and pumps.
(B) 1. Final Safety Analysis Report, Three Mile Island -
Uni; 1
- 2. IEEE 279
- 3. 10 C.F.R. 50.55 (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 97 Does the Licensee take the position that the design of TMI-l complies with Sec. 4.16 of IEEE 279, as incorporated in 10 C.F.R. 50.55 (a) (h) , in light of the TMI-2 accident? Explain your answer fully, particularly with reference to the operator's shut-off of the ECCS.
(A) ' The Licensee takes the position that the design of TMI-l complies with Sec. 4.16 of IEEE 279 as identi-fied in FSAR section 7, page 7-la and Safety Evalua-tion Docket No. 50-289, page 7-2.
During the TMI-2 accident, the protection system actuation of the ECCS went to completion once initi-ated. Subsequent deliberate operator action was used to modify the operation of the ECCS.
Restart Report Section 2 (C) None
(D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 98 Does the Licensee take the position that the TMI-2 opera-tor prevented a protection system action from going to completion?
Response
(A) Section 1, " Scope", of IEEE-279 defines " nuclear power generating station protection systems" as, . . . all electrical and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating those signals associated with the protec-tive function. These signals include those that actuate reactor trip and that, in the event of a serious reac-tor accident actuate engineered safeguards such as con-tainment isolation, core spray, safety injection, pressure reduction, and air cleaning."
From the above definition, the actual pumps and valves that carry out emergency core cooling are ex-cluded. Accordingly, per paragraph 4.16 of IEEE-279 the requirement that " . . . a protection system action shall go to completion" is satisfied when a signal has been transmitted from sensors to actuation device in-put(s) as was the case during the TMI-2 accident.
(B) IEEE-279 (C) None (D) See introductory comments.
s
(E) See introductory comments.
Interrogatory No. 99 If the answer to question No. 98 above is "yes," what specific design features have been recommended, planned or implemented to prevent this from recurring?
Response
See response to Interrogatory No. 98.
Interrogatory No. 100 Explain the present Licensee position on UCS Contention 11.
Response
See introductory comments.
Interrogatory No. 101 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 102 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention 11. Explain the reasons for which any such per-son dissents.
Response
See introductory comments.
Interrogatory No. 103 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee 's sRestart Report, which are relied upon in formulating the Licensee position on UCS Contention 11.
Response
See introductory comments.
Interrogatory No. 104 Identify all sections and page numbers oJ the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 11.
Response
FSAR and Restart Report references for UCS Contention 11 are given below:
FSAR Sections 5.6 Ventilation and Purge System 14.2.2.3 LOCA Supplement 1, Part II, Evaluation of Containment Purge as a Means of Post Accident Hydrogen Control Restart Report Sections 2.1.1.4 Post LOCA Hydrogen Recombiner 2.1.2.1.1 Containment Hydrogen Indication Supplement 1, Part 2, Response to Question 91 Interrogatory No. 105 Does the Licensee plan to install a hydrogen recombiner in TMI-l? If so, when? If not, what method of hydrogen control will be employed?
Response
(A) Metropolitan Edison plans to install a hydrogen re-combiner in TMI-1. One recombiner will be installed prior to TMI-1 restart and provisions for a second recombiner will be made in accordance with NRC Regu-latory Guide 1.7.
(B) TMI-l Restart Report (C) None s (D) See introductory comments.
(E) See introductory comments.
Inter;ogatory No. 106 If a recombiner is to be used, what will its capacity be?
How does this capacity compare with the amount of hydrogen gen-erated at TMI-2 and with the total amount of hydrogen which could be produced theoretically in the event that 100% of the fuel cladding reacted with water or steam?
Response
'A) The nominal capacity of the recombiner will be 57 SCFM.
The required capacity of a recombiner is dependent upon the hydrogen generation rate at the time than hydrogen concentration in the reactor building reaches the lower. flammability limit. No detailed analysis has been made of the hydrogen generation rate from the TMI accident.
(B) TMI-l Restart Report (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 107 Provide the detailed technical justification for the Licen-see's conclusion that the proposed design capacity for the TMI-l hydrogen recombiner is suitably conservative.
Response
(A) The Restart Report has included the appropriate sizing calculation of the recombiner for TMI-l on the basis of Regulatory Guide 1.7, 10 C.F.R. 50.44 and 10 C.F.R. 50.46 (b) (3) . Based on existing acceptance criteria our proposed recombiner provides a'mple capacity.
(B) TMI-l Restart Report. See Response to Interrogatory No. 104.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 108 Explain the present Licensee position on UCS Contention 12.
Response
See introductory comments .
Interrogatory No. 109 Does the current position differ from the position of the Licen-see in any prior proceedings? If so, identify the proceeding (s),
explain the prior position, and explain the basis for the change in position.
Response
See introductory comments .
Interrogatory No. 110 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Con-tention 12. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. lll Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention 12.
Response
See introductory comments.
Interrogatory No. 112 s Identify all sections and page numbers of the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 12.
Response
FSAR and Restart Report references for UCS Contention 12 are given below:
FSAR Sections 6.1.2.12 ECCS Environmental Considerations 6.2.2.10 Reactor Building Spray System Environmental Considerations 7.1.1.7 Protection Systems - Environment 8.2.2.10.g Selection of Control, Instrumentation and Power Cables 14.1.2.9 Steamline Failure 14.2.2.3 LOCA Supplement 2, Part IV and Part IX Restart Report Sections 8.4 (G) Upgrading of Instruments Inside Containment 2.1.1.2.3 Environmental Qualification of PORV and Safety Valve Position Instrumentation 2.1.1.5.2 Environmental Qualification of Containment Isolation Modifications 2.1.1.6.4.1 Incore Themocouple Leads 2.1.1.7.3 Steam Generator Level Instruments 2.1.2.1.1 Post Accident Monitoring Instruments Supplement 1, Part 1, Response to Question 10j Supplement 1, Part 2, Rocponse to Question 14 Interrogatory No. 126 Explain the present Licensee position on UCS Contention 13.
Response
See introductory comments.
Interrogatory No. l', 7 Does the currett position differ from the position of the Licensee in any prior proceeding? If so, identify the proceed-ing(s), explain the e ior position, and explain the basis for the change in position.
Response
See introductory comments. s
Interrogatorv No. 128 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention 13. Explain the reasons for which any such person dissents.
Response
See introductory comments.
Interrogatory No. 129 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on UCS Contention 13.
Response
See introductory comments.
Interrogatory No. 130 Identify all sections and page numbers of the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 13.
Response
FSAR and Restart Report references for UCS Contention 13 are given below:
FSAR Sections 14.0 Safety Analysis Restart Report Sections 8.0 Safety Analysis Appendix 8A Supplement 1, Part 2, Response to Questions 39 and 92 Interrogatory No. 131 Does the Licensee take the position that all credible accidents have been included within the design basis for -
TMI-l? Provide all relevant documentatiqn supporting your conclusion.
h
Response
(A) Yes. Safety systems have been provided to prevent or mitigate the consequences of all credible acci-dents, and the consequences of all credible acci-dents are bounded by the accident analyses for design basis accidents.
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1; Restart Report (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 132 What is the probability that an accident beyond the design basis for TMI-l will occur? Provide all relevant documentation supporting your conclusion.
Response
(A) The Licensee has not performed probability assessment studies of any accident beyond the design basis of TMI-1.
(B) None (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 133 What is the probability that the following accident scenari,os described in WASH-1400 will occur at TMI-l?
h
- a. PWR 5
- b. PWR 4
- c. PWR 2 Provide all relevant documentation supporting your conclusion.
Response
(A) The Licensee has not undertaken probability assess-ment studies for hypothetical accidents of PWR release categories #2, #4, or #5 for TMI-1.
(B) None (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatorv No. 134 State what the health safety and environmental conse-quences would be (short and long term) of each of the accident scenarios identified in question #135 above, including property damage.
Response
It is assumed that the wording in the Interrogatory should have referenced question #133 above instead of question #135 above.
See response to Interrogatory No. 133.
Interrogatory No. 135 Define " Class 9" accidents.
' Response A viass 9 accident has been defined by NRC in a proposed
. annex to 10 C.F.R. 50 Appendix D. Licensee has no other definition
m but notes the importance of distinguishing between accident events and accident consequences.
Interrogatory No. 136 Prior to the accident at TMI-2, did the Licensee have an opinion as to the probability of the TMI-2 accident? What was that opinion?
3esponse (A) No.
(B) None (C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 137 Provide all draft and final analyses, memoranda, reports recommendations or other documents produced by, relied upon or consulted by the licensee relating to the probability or con-sequences of accidents beyond the current design basis and/or to measures designed to mitigate the consequences of such acci-dents at TMI-1.
Response
See Responses to Interrogatories 132, 133, 134 and 136.
Interrogatory No. 138 Does the Licensee agree that " . . . the accident at Three Mile Island exceeded many of the present design bases by a wide margin and was evidently a significant precursor of a core-melt accident . . ." (NUREG-0585, p. 3-5).
Response
(A) The licensee agrees that some of the design bases as defined in Reference 1 were exceeded during the s
s .
Three Mile Island-2 accident as described in Refer-ences 2 and 3. The licensee does not agree that the Three Mile Island-2 accident was a "significant precursor of a core-melt accident," based on the conclusions reached in References 3 and 4.
(B) References
- 1. TMI-2 Final Safety Analysis Report, Docket No.
S0-320.
- 2. NSAC-1, Analysis of Three Mile Island Unit 2 Accident, July 1979.
- 3. Report of the President's Commission on the Accident at Three Mile Island.
- 4. Letter Report, W.H. Layman (NSAC) to NSAC Coordinators and Owners Group Chairman, "TMI-2 Meltdown Studies".
- 5. NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report", October 1979.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 139 Does the Licensee agree that " [blecause the accident at Three Mile Island exceeded many of the present design bases by a wide margin and was evidently a significant precursor of a core-melt accident, . . . the NRC should begin to formulate re-quirements for design features that could mitigate the conse-quences of core-melt accidents." (NUREG-0585, p. 3-5).
. Re sponse (A) As stated in the Response to Interrogatory 138, the Licensee does not feel that the TMI-2 accident was "a significant precursor of a cose-melt accident".
Nevertheless, we recognize that the NRC Task Force recommended rulemaking relating to the considera-tion of design features to mitigate core-melt and degraded core accidents. However, any future incor-poration of these mitigation features should be given very careful consideration and NRC research in this area should continue in order to determine the necessity of such requirements. Requirements for design modifications should be considered in the context of total risk and other actions appropriate to minimize risk.
(B) References
- 1. NUREG-0 5 85, "TMI-2 Lessons Learned Task Force Final Report," October 1979.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 140 Given that TMI-2 has been identified by the Staff as a Class 9 accident, and Class 9 accidents pose the gravest threat tc the public safety of all possible nuclear reactor accidents, ex1' lain how the Licensee can " provide reasonable assurance tha t appropriate measures can and will be taken in the event of an emergency to protect public health and safety . . . " with-ouc consideration of Class 9 accidents and their consequences.
Response
Absent a description of the accident scenario (s) intended to be covered by this Interrogatory, Licensee is unable to answer the Interrogatory. '
m Interrogatory No. 141 Explain the present Licensee position on UCS Contention 14.
Response
See introductory comments.
Interrogatory No. 142 Does the current position differ from the position of the Licensee in any prior proceedings? If so, identify the proceed-ing(s), explain the prior position, and explain the basis for the change in position.
Response
See introductory comments.
Interrogatory No. 143 Identify any officers or employees of, or consultants to, the Licensee who dissent from the present Licensee position on UCS Contention 14. Explain the reasons for which any such person dissents.
Response
See introductory comments .
Interrogatory No. 144 Identify the specific sections and page numbers of the FSAR for TMI, Unit 1, and the Licensee's Restart Report, which are relied upon in formulating the Licensee position on Contention 14.
Response
See introductory comments.
Interrogatory No. 145 Identify all sections and page numbers of the FSAR and Restart Report which contain subject matter pertaining to UCS Contention 14.
s
4 %
Response
FSAR and Restart Report references for UCS Contention 14 are given below:
FSAR Sections See References for Interrogatory 112 Restart Report Sections See References for Interrogatory 112 Interrogatory No. 147 Does the licensee agree that some systems and components presently classified as non-safety-related can have an adverse effect on the integrity of the core because they can directly or indirectly affect temperature, pressure, flow and/or reac-tivity? Identify all such systems and components related to the core cooling systems.
Response
(A) There are no systems and components presently classi-fied as non safety-related that can have an adverse effect on the integrity of the core. There are some systems and components classified as non safety-related which can initiate or produce a transient condition, but safety systems are adequately designed to protect the integrity of the core in the event of such transient condition.
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1 (C) None (D) See introductory comments.
(E) See introductory comments.
s
i .
Interrogatory No. 148 Which of the short and/or long term measures recommended by the staff, or proposed and/or implemented by the Licensee, are directed toward preventing adverse effects on the integrity of the core caused by non-safety-related systems and components?
How will these measures correct the deficiencies identified in NUREG-0 5 7 8, Section 3.27 What is their schedule for implementa-tion?
Response
(A) Recommendation 9, page A-14 of NUREG-0585, "TMI-2 Lessons Learned Task Force Final Report" states "The owners of operating plants and all plants under construction should be required to evaluate the interdction of non-safety and safety grade sys-tems . . . to assure that any interaction will not result in exceeding the acceptance criteria for any design basis accident." This recommendation is directed at determining if there are any previously unidentified interactions which could result in ex-ceeding acceptance criteria including core integrity limits. The approach to and schedule for implementa-tion of this recommendation are currently being developed. No other short and/or long term measures recommended by the staff or proposed and/or imple-mented by the Licensee are considered to be directed toward preventing adverse effects on the integrity of the core caused by non safety-related systems and com-ponents.
s
(B) 1. NUREG-0 5 7 8, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations. "
- 2. NUREG-0 5 8 5, "TMI-2 Lessons Learned Task Force Final Report".
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 149 Does the Licensee propose to classify the reactor coolant pumps as safety-related? If not, explain your answer fully.
Response
No. S3e response to Interrogatory No. 19.
Interrogatcry No. 150 Does the licensee propose to classify the steam genera-tors as saf cy-related? If not, explain your answer fully.
Response
(A) The steam generators are safety-related in that they are part of the reactor coolant pressure boundary.
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 151 With which of the General Design Criteria applying to safety-related systems, structures and components does the Licensee propose to comply, with respect to the reactor coolant pumps? If the Licensee proposes to comply with less than all of them, explain why those not applied have been excluded.
m
Response
For the reasons stated in Responses to Interrogatories 6 and 19, the reactor coolant pumps meet the requirements of GDC 2 and 4 to the extent noted in the Response to Interrogatory 24.
No changes are proposed in the reactor coolant pumps.
Interrogatory No. 152 Answer the same question as #151 above with regard to the steam generators.
Response
(A) Section 1.4.1 of the TMI-l FSAR lists and discusses those gensral design criteria that were considered during the design and construction of the unit. No changes are proposed in the steam generators.
(B) Final Safety Analysis Report, Three Mile Island -
Unit 1.
(C) None (D) See introductory comments.
(E) See introductory comments.
Interrogatory No. 153 Explain how the Licensee can assure the adequate protec-tion of the public health and safety when systems and components, which are classified as non-safety-related but already have been demonstrated to contribute to the aggravation or mitigation of the TMI-2 accident, do not meet all safety grade design criteria.
Response
See response to Interrogatory No. 147.
Dated: January 25, 1980 s
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
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
AFFIDAVIT OF ROBERT W. KEATEN District of Columbia, ss:
ROBERT W. KEATEN, being duly sworn according to law, deposes and says that he is the Manager of Systems Engineering of General Public Utilities Service Corporation; that the information contained in Licensee's Responses to the following Interrogatories are true and correct to the best of his knowledge and belief: Union of Concerned Scientists' ("UCS")
Interrogatories Nos. 1-37, 40-47, 50-112, 126-145, and 147-153, filed by the UCS on January 4, 1980.
/
fU. tw Robert W. Keaten Manager of Systems Engineering Sworn to and subscribed before me this 451 day of January, 1980.
buy 't fj{!N':r
- w. Mo,t,try Public
,My Commission Expires '.
s e
January 29, 1980 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensina Board In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
CERTIFICATE OF SERVICE I hereby certif that copies of " Licensee's Answers to a Portion of Union of Concerned Scientists' (UCS) Interrogatories Dated January 4, 1980" together with Licensee's accompanying affidavit and transmittal letter, were served on those persons listed on the attached Service List by U.S. mail, first class, postage prepaid, this 29th day of January, 1980.
/ YQ , . .dj$
Harry H Glasspieg'el '
De.ted: January 29, 1980
..s s en U'iITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
) (Restart)
(Three Mile Island Nuclear )
Station, Unit No. 1) )
SERVICE LIST Ivan W. Srcith, Esquire John A. I.evin, Esquire Chairan Assistant Counsel At=ic Safety and Licensing Penrsfl vania Public U*dlity Cam'n Board Panel Pest Office Ecx 3265 U.S. Nuclear Pagulatory C=:rissien Harrisburg, Pennsylvarla 17120 Washingten7 D.C. 20555 Karin W. Carter, Esquire Dr. Walter H. Jcrdan Assistant Attorney General Atric Safety and Licensing 505 Executive House Board Panel Pcst Office Scx 2357 881 West Cuter Drive Harrd eurg, Pennsylvania 17120 Cak Ridge, Tennessee 37830 Jchn E. Minnich Dr. Linda W. Little Chai=an, Dauphin Counry Scard Atc:ric Safety and Licensing of Ccnmissioners Board Panel Dauphin County Ccurthouse 5000 P.e =itage Drive Front and Market Streets Pale.@, North Carolina 27612 Harrisburg, Pennsylvania 17101 Janes R. Tourtellotte, Esquire Walter W. Cchen, Esquire Office of the Executive Lecal Director Censumer Advccate U. S. Nuclear Pegulatory C5=rission Cffice of Consumer Mvocate Washington, D.C. 20555 14th Floor, Strrberry Square Ha risburg, Pe=.sylvania 17127 Docketing and Service Section Office of the Secretary U. S. Nuclear Pegulatory Ccmissicn Washingten, D.C. 20555
Jc h D. Cunninghan, Esquire Karin P. Sheldon, Esge. ire Attccey for Ueaterry Tcar. ship Attc =ey for Pecple Against Nuclear T.M.I. Steering Ccmittee Energy Shelden, Haren & Weiss 2320 Scrth Seccnd Street 1725 Eye Street, N.W., Suite 506 Harrisburg, Pennsylvaria 17110 Washingten, D.C. 20006 Theodore A. Adler, Esquire Widoff Peager Selkcwit: & Mler Pcbert Q. Pollard Chesapeake Energy Alliance Post Office Ecx 1547 609 Montpelier Street Harrisburg, Pen:tsylvania 17105 Balthere, Maryland 21218 Ellyn R. Weiss, Esqdre Chauncey Kepferd Atta=ey for the Crien of Ccnce=ed Scientists J"#4th H. Jernsrud Shelden, Haren & Weiss Environmental Coalition en Nuclear 1725 Eye Street, N.W. , Suite 506 Power Washingten, D.C. 20006 433 Criand Avenue State Ccliege, Pennsylvania 16801 Steven C. Shclly 304 South Market Street Marvin I. Isais Machanicsburg, Pennsylvania 17055 6504 Bradford Ter ace Philadelphia, Pennsylvania 19149 Gail Bradford Marjorie M. 4tTodt Eclly S. Keck Iagislaticn Chaiman R. D. 5 Anti-Nuclear Grouc Papresenting York Ccatesville, Pennsylvarla 19320 245 West Philadelphia Street York, Pennsylvania 174C4 s