ML19257A248
| ML19257A248 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/04/1979 |
| From: | Dieterich R SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | CALIFORNIA, STATE OF |
| Shared Package | |
| ML19257A247 | List: |
| References | |
| NUDOCS 8001030264 | |
| Download: ML19257A248 (15) | |
Text
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December 4, 1979 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312
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(Rancno Seco Nuclear Generating Station)
LICENSEE'S ANSWERS (SET NO. 1) TO THE FIRST SET OF INTERROGATORIES OF THE CALIFORNIA ENERGY COMMISSION DATED NOVEMBER 15, 1979 1.
INTERROGATORY:
Identify and provide summaries and conclusions of any documents prepared since March 28, 1979, with respect te ene facility or with respect to Babcock and Wilcox ("B&W") reactory systems which relate, in whole or in part, to any of the following:
a.
Small break loss-of-coolant accidents; b.
Conditions of inadequate core cooling; Sensitivity evaluations of delays in start-up c.
of the auxiliary feedwater system; d.
Sensitivity evaluations of steam generator design parameters such as volume and hydraulic characteristics; e.
Sensitivity evaluations of reactor trip setpoints, relief and safety valve setpoints, and ECCS setpoints; f.
Sensitivity evaluations of operating reactor-power level; g.
Sensitivity evaluations of pressurizer size and hydraulic characteristics; and h.
Sensitivity evaluations of reactor drain tank size and design pressure.
ANSWER:
Documents identified in response to this interrog-atory wil' be provided pursuant to Licensee's Response to California Energy Commission's First Request for Production of Documents.
Because the documents themselves will be produced, summaries and conclusions are not provided.
1666.238 8001030 2bN
2.
INTERROGATORY:
Describe the changes, if any, in racility design, equipment and/or operating procedures that have been proposed and/or instituted or are being studied or contemplated for the f acility as a result of the Three Mile Island ("TMI") incident.
Identify all documents related to any such changes.
ANSWER:
Tne following changes in design, equipment or operating procedures have been instituted at Rancho Seco as a result of the Three Mile Island incident:
In a letter of April 16, 1979, SMUD informed the Nuclear Regulatory Commission of several changes to Rancho Seco operating procedures relating to the oper-ation of the high pressure injection system, reactor coolant pumps, and operator utilization of pressure, temperature relationships.
In a letter of April 22, 1979, SMUD informed the Nuclear Regulatory Commission of a cnange which low-ered the high pressure reactor trip setpoint from 2355 psig to 2300 psig, and raised tne pilot operated relief valve setpoint from 2255 psig to 2450 psig.
In a letter of May 2, 1979, SMUD informed the Nuclear Regulatory Commission of procedure changes: for es-tablisning and maintaining natural circulation; for consideration of reactor vessel integrity in the determination of high pressure injection Listem ter-mination; to require prompt manual trip of the reactor on loss of feedwater to the steam generators, during a turbine trip, on loss of offsite power with loss of reactor coolant flow, on low steam genera *or level, or 1666 239 on low pressurizer level; and for reporting procedures for NRC notification any time the Rancho Seco reactor is not in a contro. led or expected condition of operation.
In a letter of May 14, 1979, SMUD informed the Nuclear Regulatory Commission of several operating procedure changes relating to the operation of the auxiliary feedwater system.
The NRC was also informed, in that letter, of a cnange in the control room annunciation for automatic start conditions of the auxiliary feedwater system, the addition of auxiliary feedwater flow indication in the control room, and the addition of the hard wired control grade reactor trip on loss of main feedwater or turbine trip.
The control grade trip of the reactor on turbine trip or loss of main feedwater was further described to tne NRC by a telecopy transmittal on May 30, 1979.
In a letter of August 27, 1979, SMUD informed the Nuclear Regulatory Commission of an operating procedure change requiring a trip of all operating reactor coolant pumps upon a reactor trip and initiation of high pressure injection caused by low reactor coolant system pressure.
1666 240 Tne f ollowing changes in design, equipment or operating procedures have been proposed for Rancho Seco as a result of t'.e Three Mile Island incident:
In a letter of September 17, 1979, SMUD informed the Nuclear Regulatory Commission of several design changes it is considering which would improve the auxiliary feedwater system reliability.
In a letter of May 21, 1979, SMUD informed the Nuclear Regulatory Commission of 2 proposed change for a safety grade automatic onticipatory reactor SCRAM on loss of feedwater or tarbine trip.
On October 5, 1979, SMUD informed the Nuclear Regulatory Commission of furtner details for the safety grade anticipatory reactor trip.
In addition to the above described changes, SMUD is considering various other modifications as a result of tne NRC S taf f Report, "TMI-2 Lessons Learned Task Force Status Report and Short-term Recommendation", NUREG-0578.
These modifications are described in SMUD's letters to the NRC dated October 18, 1979, November 19, 1979, and November 26, 1979.
The letters identified above, along with other documents related to the changes, describe the changes and will be provided pursuant to Licensee's Response to California 1666 241 Energy Commission's First Request for Production of Documents.
3.
INTERROGATORY:
For each change described in response to Interrogatory 2, provide the following additional data:
a.
A description of the purpose of the change; b.
A senedule d'.cailing when the change was, shall or may be instituted; c.
A description of any constraints, including legal, regulatory, technological, or economic, whicn affect incorporation of the change; and d.
An estimate of the cost of the change.
ANSWER a, b:
Tne purpose (s) of each change and the schedule for implementation are included in the documents identified above in the answer to Interrogatory No.
2.
ANSWER c:
All proposed design and procedure changes must receive sufficient development, review and evalu-ation to determine their adequacy and acceptability prior to implementation.
Depending on the complexity of the items, this process may be protracted.
NRC approval may also be required, in which case further delay may be incurred and additional proposals may be required.
Following necessary reviews and approvals, hardware pro-curement may take several weeks to several years.
Quality assurance requirements must be met to assure that compon-ents meet the safety requirements and quality standards in-corporated in tne original safety analysis, and may cause delays in procurement or installation.
1666 242 Many changes, particularly those relating to the engi-neered safety features and reactor protection systems, necessitate a reactor shutdown to implement the cnange.
The costs of a shutdown are large, even for a minor change.
Therefore, it is economically desirable to make multiple changes at one time, usually at an annual refueling and maintenance outage wnen the plant is shut down for several weeks.
ANSWER d:
The dominant portion of the costs for the cnanges identified is for professional engineering time whicn is not accounted for in each specific change.
It is e s t ima ted, nowever, that total SMUD expenditures to date for these changes have been approximately $800,000.
4.
INTERROGATORY:
With respect to the potential small break analyses performed by SMUD in response to the NRC's Order of May 7, 1979, what operating instructions have been implemented by SMUD to define proper operator actions?
Describe any additional changes in operating instructions whicn have been implemented since March 28, 1979, together witn tneir rationale and dates of implementation.
Identify all documents relating to the small break analyses and/or operating instruction changes.
ANSWER:
The small break analysis information has been submitted to the Nuclear Regulatory Commission on May 14, July 2, August 27, and September 19, 1979.
The letters transmitting these analyses describe the operating proce-dure changes which were made as a result of the guidelines developed from the analyses.
These letters will be pro-vided pursuant to Licensee's Response to California Energy Commission's First Request for the Production of Documents.
Additional changes to operacing procedures not related to 1666 243 tnese small break analyses have been made as described in tne answers above to Interrogatories 2 and 3.
See the answer above to Interrogatory 1 for additional documents related to the small break analyses.
7.
INTERROGATORY:
Describe all documents or safety analydas prepared by SMUD for combinations of failures in the saf aty and relief valves of the primary system.
AN5WER:
The failure of a reactor coolant system safety and/or relief valve in effect is a small break loss of coolant accident.
The small break analysis discussed above in tne answer to Interrogatory No. 4 covers a spectrum of breaks which includes that represented by a stuck-open relief valve and any combination of stuck-open relief and/or safety valves.
See the answers to Interrogatories 1 and 4 for tne documents related to small break analyses.
8.
INTERROGATORY:
Describe the design bases of each component, including the integrated control system, which affects the reliability of the auxiliary feedwater system.
Describe the degree of seismic design and protection from other potential nazards such as on-site explosions, fires, or seismic failures of nearby non-safety related equipment and structures (including the natural draft cooling tower s).
The availability of backup water supplies for the auxiliary feedwater system should be included in response to this interrogatory.
ANSWER:
On September 17, 1979, SMUD provided the Nuclear Regulatory Commission with a draft version of an auxiliary feedwater system reliability analysis.
This analysis will be finalized in December, 1979, at which time it will be made available for inspection and copying.
In the mean-time, the basic conclusions of the draft report are valid and provide the information requested by this interrogatory.
Tne analysis identifies contributors to the Rancho Seco auxiliary feedwater system reliability, and employs a fault-tree analysis technique, using operating reliability data for contributing components.
It should be noted that the integrated control system does not affect the relia-bility of the auxiliary feedwater system since Rancho Seco operators are trained to control the auxiliary feedwater system independent from tne integrated control system.
In addition, SMUD has committed to install a control system separate from the integrated control system.
The auxiliary feedwater system is classified as a Seismic Class I system and is protected from on-site hazards such as explosions, fires, or seismic failures by location or barriers.
SMUD's letter of September 17, 1979, providing the draft reliability analysis and the commitment for a separate con-trol system, and the Final Safety Analysis Report, will be provided pursuant to Licensee's Response to the California Energy Commission's First Request for Production of Documents.
The backup water supplies for the auxiliary feedwater sys-tem consist of both the Folsom-South Canal and tne 2850 acre feet Rancho Seco Lake, which provide an essentially unlimited amount of water.
1666 245 9.
INTERROGATORY:
Describe tne design and/or safety analyses for containment building hydrogen control, if any, wnich have been prepared by SMUD or which SMUD has obtained from other sources.
Describe the containment building instrumentation and any anticipated changes.
ANSWER:
The design and analysis of tne containment building nydrogen control system is described in Appendix 14C to tne Rancho Seco Final Safety Analysis Report.
The containment building sampling system which would be used to measure hydrogen concentration is also described in this appendix to tne FSAR.
SMUD does not anticipate any changes to tne instrumentation system for measuring hydrogen concentration as a result of the Three Mile Island accident.
- 10. INTERROGATORY:
Describe each evaluation or study, if any, of controlled filtered venting which has been performed by SMUD or which SMUD has obtained from other sources.
ANSWER:
The Rancho Seco design incorporates a filtered ventilation system for the reactor building.
This system is described in Section 9.7 of the Final Safety Analysis Report.
This system is used to filter and clean the reactor building air during normal plant operation and is not intended for use following an accident.
The Nuclear Regulatory Commission's "TMI-2 Lessons Learned Task Force Final Report", NUREG-0585, recommends further study of design f eatures for core-damage and core-melt accidents.
SMUD has not performed any such evaluations or studies and is not aware of any such studies at this time.
- 30. INTERROGATORY:
Identify each person who SMUD expects to call to testify at the hearing in this proceeding.
For each person identified, provide: the subject (s) upon which the person may testify; a description of the substance of
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the testimony; and a description of the person's educa-tional background and professional qualifications.
ANSWER:
Robert A.
Dieterich, Senior Nuclear Engineer in SMUD's Generation Engineering Department, is expected to provide testimony on contentions, identified below, relating to the design of Rancho Seco.
Mr. Dieterich is a graduate of the University of Kansas with a Bachelor of Science degree in Engineering Physics, with a major area of study in nu: lear engineering.
He nas also taken graduate level courses in nuclear engineering from the University of Wasnington while employed by the General Electric Company in that state.
Mr. Dieterich worked for three years as a Process Physicist for tne General Electric Company at the Hanford operations in Richland, Washington.
In tnis position he performed all routine physics calculations for an assigned plutonium production reactor, providing fuel loading patterns and operating techniques consistent with established safety criteria.
Following this period, Mr. Dieterich accepted a position as a Nuclear Engineer with tne General Electric Company in their Nuclear Energy Division in San Jose, California.
In this position he performed analyses of design basis reactivity accidents for safety analysis reports.
He also participated in the licensing efforts for the Oyster Creek and Nine Mile Point nuclear power plants, and nad total responsibility for the preparation of the Nine Mile Point and Monticello technical specifications.
1666 247 Following his employment with the General Electric Company, Mr. Dieterich accepted a position witn tne Sacramento Municipal Utility District, where he has been employed for tne last nine years.
Mr. Dieterich is presently a Senior Nuclear Engineer in the Generation Engineering Department and has had responsioilities in the design, erection, startup and licensing of Rancho Seco.
Mr. Dieterich is a member of Sigma Pi Sigma (a physics nonorary society) and the American Nuclear Society.
Mr.
Dieterich is currently registered as a Professional Nuclear Engineer in the state of California, registration number N103.
The specific contentions to which Mr. Dieterich will tes-tify are listed below with a description of the substance of his testimony.
Issue CEC 1-6 Will the modifications of subparagraphs a-e still leave the Rancno Seco emergency feedwater system in a condition of low reliability?
Hursh-Castro Contention 7 Rancno Seco, being a Babcock and Wilcox designed reactor, has insufficient timeliness and reliability of the emer-gency feedwater system, and therefore is unsafe and endan-gers tne nealth and safety of Petitioners, constituents of Petitioners and the public.
Description of Substance of Testimony This testimony will demonctrate that the Rancho Seco auxiliary feedwater system is a system of high reliability and sufficient timeliness, will discuss a reliability study 1666 248
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performed for that system, and will show that the modifica-tions described in subparagraph a to e nave improved that reliability. (See also Licensee's Answers (Set No. 2).]
Issue CEC 5-1 i
Whether those systems identified as contributing to releases of radioactivity during the TMI accident, which are outside containment, should be changed to vent into the containment ouilding?
Description of Substance of Testimony Tnis testimony will show that the containment isolation system at Rancho Seco is different from that at Three Mile Island, and that the possibility of radioactivity releases from outside the containment are reduced at Rancho Seco.
Tnerefore, radioactivity containing systems outside the containment building at Rancho Seco need not be changed to vent into tne containment building.
Issue CEC 5-2 Whether the containment building should be modified to provide overpressurization protection with a controlled filtered venting system to mitigate unavoidable releases of radionuclides?
Description of Substance of Testimony This testimony will show that the reactor containment building is designed to withstand accident pressures.
Hursh-Castro Contention 3 Rancno Seco, being a Babcock and Wilcox designed reactor, has a lack of direct initiation of reactor trip upon.the occurrence of of f-normal conditions in the feedwater 166,6 249
system, and therefore is unsafe and endangers the health and cafety of the Petitioners, constituents of Petitioners and tne public.
Hursh-Castro Contention 9 Rancho Seco, being a Babcock and Wilcox designed reactor, has not installed adequate nard-wire control grade reactor trip on loss of main feedwater and/or on turbine trip, and therefore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
Description of Substance of Testimony This testimony will indicate that a direct initiation of reactor trip upon turbine trip or loss of feedwater nas been installed at Rancho Seco.
Hursh-Castro Contention 5 Rancho Seco, being a Babcock and Wilcox designed reactor, has an actuation before reactor trip of a pilot operated relief valve on the primary system pressurizer which, if the valve sticks open, can aggravate an accident, and tnerefore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
Description of Substance of Testimony Tnis testimony will show that tne setpoints for the reactor trip on high primary system pressure and the pilot operated relief valve have been cnanged so that the reactor trip occurs before the pilot operated relief valve is actuated.
Hursh-Castro Contention 20 Rancho Seco, being a Babcock and Wilcox designed reactor, does not nave a hydrogen recombiner which may be necessary in the event of an accident caused by a loss of feedwater trans ient, and tnerefore is unsafe and endangers the health and safety of Petitioners, constituents of Petitioners and the public.
Description of Substance of Testimony This testimony will describe arrangements SMUD has made for the provision of a hydrogen recombiner on short notice at Rancho Seco. 1666 250
s FOE Contention III(c)
The NRC orders in issue do not reasonably assure adequate safety because there is no reasonable time for implementa-tion of the long-term modifications established in the Commission orders.
Description of Substance of Testimony This testimony will provide the schedules for the imple-mentation of the long-term modifications established in the Commission Order.
It will be shown that Rancho Seco is being operated safely at the present time and that the long-term modifications are sufficient to provide continued reasonable assurance that the facility will respond safely to feedwater transients.
1666 251 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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)
SACRAMENTO MUNICIPAL UTILITY DISTRICT
)
Docket No. 50-312
)
(Rancho Seco Nuclear Generating Station)
)
AFFIDAVIT OF ROBERT A.
DIETERICH County of Sacramento )
SS State of California
)
Robert A.
Dieterich, being duly sworn according to law, deposes and says that he is a Senior Nuclear Engineer in the Generation Engineering Department of the Sacramento Municipal Utility District; and that the answers contained in " Licensee's Answers (Set No. 1) to the First Set of Interrogatories of the California Energy Commission dated November 15, 1979" are true and correct to the best of his knowledge and belief.
Robert A.
Dieterich Sworn to and subscribed before me this 4th day of December, 1979.
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