ML13302B164
| ML13302B164 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/02/1981 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | Pagano F, Speis T Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8109110525 | |
| Download: ML13302B164 (49) | |
Text
SEP 2 1981
'0?
MEMORANDUM FOR: F. G. Pagano, Chief, Emergency Preparedness Licesiig Branch, DEP T. P. Speis, Chief, Reactor Systems Branch, DSI FROM:
F. J. Miraglia, Chief, Licensing Branch No. 3, DL
SUBJECT:
San Onofre-2"& 3, Low Power License Southern California Edison (SCE) has filed a motion with the ASLB for consid eration of issuance of a low power license for San Onofre 2 and 3.
As a result, it will be necessary for Emergency Preparedness staff and Reactor Systems staff to prepare testimony similar to that filed in the Diablo Canyon proceeding.
My understanding is that this testimony is to be filed by September 18, 1981.
The case attorneys have requested that draft testimony be provided to them by c.o.b. September 11, 1981.
I have enclosed a copy of the Diablo Canyon testimony for your use (Note:
Testimony on block valves is not relevant to the San Onofre proceeding).
It is imperative that the above schedule be met so as not to delay the San Onofre 2 & 3 proceeding. Please inform me (X29779) or Harry Rood (X28544) of the names of the individuals who will prepare the requested testimony, no later than c.o.b. September 4, 1981.
The project manager and case attorneys will meet with these individuals at an early date.
Drriginal signed bY Frank. Miraglia Frank J. Miraglia, Chief Licensing Branch No. 3 Division of Licensing cc:w/o encl DISTRIBUTION:
D. Eisenhut Docket FIle 50-361/362 R. Tedesco LPDR L. Chandler, OELD PDR R. Mattson NSIC B. Grimes TERA LB3Files_---
liRood 'w/errd.
8109110525 810902 DR ADOCK 05000361 RHoefling LPPDR OFFNSIC DLB# 1iles SURNAMRboeflingi DATE.........
QV.... dLIL NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981335-9\\
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 8, 1981 John F. Wolf, Esq., Chairman Dr. Jerry Kline Administrative Judoe Administrative Judge Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 Glenn 0. Bright, Esq.
Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 In the Matter of Pacific Gas and Electric Company Diablo Canyon Nuclear.Power Plant Unit Nos. 1 and 2 Docket Nos. 50-275 & 50- 323 0.L.
Dear Administrative Judge:
Enclosed find the NRC Staff Testimony of Bartholomew C. Buckley, John Sears, Frank C. Cherny, and Norman Lauben for the Diablo Canyon Low Power test proceeding. This testimony is filed pursuant to the schedule adopted by the Board in the February 13, 1931 Prehearing Conference Order.
Sincerely, rnBradley W. Jones ounsel for NRC Staff
Enclosure:
As stated above cc:
Service List
1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COM$ISSION 3
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4
In the Matter of
)
5
)
6 PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 O.L.
7
)50-323 O.L.
8 (Diablo Canyon Nuclear Power Plant )
9 Unit Nos. 1 and 2)
)
10 TESTIMONY OF BARTHOLOMEW C. BUCKLEY 11 Q.
Please state your full name.
12 A.
Bartholomew Christopher Buckley.
13 Q.
By whom are you employed?
14 A.
The Nuclear Regulatory Commission (NRC).
15 Q.
Describe the nature of your work with respect to the Diablo Canyon 16 operating license proceeding.
17 A.
Since March 11, 1979, as a Senior Licensing Project Manager, I am 18 responsible for the management of the NRC Staff's safety review and 19 the preparation of safety evaluation reports in support of the 20 issuance of operating licenses for Diablo Canyon, Units 1 & 2.
21 Q.
Have you, in the course of your professional experience, had 22 occasion to act as a project manager for plants other than Diablo 23 Canyon?
24 A.
Yes, the Rancho Seco and Crystal River Nuclear Power Plants.
N 0
S-2 25 Q.
Would you detail your professional qualifications?
26 A. Attached is a copy of my professional qualifications.
27 Q. Are you familiar with the SER and supplements thereto which have 28 been issued for the Diablo Canyon Nuclear Facility?
29 A.
Yes.
30 Q.
Did you participate in the preparation of those documents?
31 A.
Yes.
32 Q.
Please detail the extent of your participation in the preparation of 33 the SER and Supplements thereto for the Diablo Canyon Nuclear 34 Facility.
35 A. After receipt of SER inputs from the various technical branches, I 36 prepared Supplements Number 9, 10, 11, 12, 13 and 14. The Safety 37 Evaluation Report, dated October 16, 1974, and Supplements 1 through 38 8 were prepared by other project managers.
39 Q.
Would you please identify all the individuals who prepared the 40 sections of the SER and Supplements thereto which relate to the 41 contentions admitted in this low power proceeding?
42 A.
Emergency Planning:
Peter Tam (former NRC employee); Relief and 43 Safety Valves:
Edgar Hemminger 44 Q.
Have you reviewed the input to the SER of the above identified 45 individuals?
-3 45 A.
Yes.
47 Q.
Do you agree with the analysis and-conclusions reached by those 43 individuals within the limits of your professional expertise?
49 A.
Yes.
50 Q.
Considering that review do you, based on your professional 51 experience, have any concern that the issues raised by the admitted 52 contentions indicate any danger to the public safety and health by 53 operation of the Diablo Canyon Nuclear Facility at low power?
54 A.
No.
UNITED STATES OF AAERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 O.L.
)
50-323 O.L.
(Diablo Canyon Nuclear Power Plant )
Unit Nos. 1 and 2)
)
BARTHOLOMEW.C. BUCKLEY PROFESSIONAL QUALIFICATIONS DIVISION OF LICENSING I am a Senior Project Manager of the Staff of the U. S. Nuclear Regulatory Commission. I am responsible for the management of the NRC Staff's safety review and the preparation of safety evaluation reports in support of the issuance of operating licenses for Diablo Canyon, Units 1 and 2.
I received a Bachelor of Chemical Engineering Degree at the Catholic University of America in 1953 and I studied for a Master's Degree in Nuclear Engineering at the above cited University.
I started my employment with the Regulatory staff on January 3, 1972 and have worked as a Project Manager ir Operating Reactors Branch and as a Section Leader in the Division of Operating Reactors.
I was employed at the Naval Nuclear Power Unit, Fort Belvoir, Virginia from 1966 to December 1971.
I worked as a chemical engineer for approximately two years, and as Director of the Engineering Depart ment until December 1971.
From 1960 to 1966 I worked at the Atomic Energy Division of Allis Chalmers. I prepared primary system and auxiliary preoperational test
-2 procedures and conducted and evaluated these preoperational tests at the Elk River Reactor. I also prepared a water chemistry manual and performed water chemiistry analyses at the Elk River Reactor.
I prepared all fluid systems preoperational test procedures for the LaCrosse Reactor and taught plant operating personnel in the design and operating aspects of certain systems.
I worked as a process control engineer at the MELPAR Co. located in Virginia from 1958 to 1960.
1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COMMISSION 3
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4
In the Matter of
)
6 PACIFIC GAS AND ELECTRIC COMPA'AY
)
Docket os. 50-275 O.L.
7
)50-323 O.L.
8 (Diablo Canyon Nuclear Power Plant )
9 Unit Nos. 1 and 2)
)
10 TESTIMONY OF JOHN R. SEARS 11 Q.
Please state your full name.
12 A.
John R. Sears.
13 Q.
By whom are you employed?
14 A.
I am employed by the U.S. Nuclear Regulatory Commission. I am a 15 Senior Reactor Safety Engineer in the Emergency Preparedness 16 Licensing Branch, Division of Emergency Preparedness of the Office 17 of Inspection & Enforcement.
18 Q.
Describe the nature of your work with respect to the Diablo Canyon 19 operating license proceeding?
20 A.
I am responsible for review and evaluation of the Emergency Plan for 21 the Diablo Canyon reactors.
22 Q.
Have you reviewed the emergency plans for plants other than Diablo 23 Canyon?
24 A.
Yes.
2 25 Q.
What previous review experience have you had with respect to 26 emergency plans?
27 A.
Prior to the summer of 1979, 1 was responsible for review of the 23 emergency plans for all operating reactors.
29 Q.
Would you detail your professional qualifications?
30 A. Attached is a copy of my professional qualifications.
31 Has PG&E revised the Diablo Canyon Power Plant Emergency Plan?
32 A.
PG&E filed a revision to the Diablo Canyon Power Plant Emergency 33 Plan in February of 1980. PG&E also provided additional information 34 in letters, PG&E to NRC, dated January 13, 1981 and February 27, 35 1981.
36 Q.
Did you participate in the NRC review of these documents?
37 A.
Yes.
38 Q.
How was the Emergency Plan evaluated?
39 A. The Emergency Plan was reviewed against the specific criteria of the 40 sixteen Planning Standards in Part 11 of the "Criteria for 41 Preparation and Evaluation of Radiological Emergency Response Plans 42 and Preparedness in Support of Nuclear Power Plants," NUREG-0654, 43 Rev. 1, November 1980. The same sixteen standards are listed as 44 requirements of 10 C.F.R. 50.47, Final Regulations on Emergency 45 Planning, August 1, 1980. The NRC Staff has prepared an Emergency 46 Preparedness Evaluation Report which lists each of the sixteen
-3 47 standards listed in 10 C.F.R. 50.47 in order, followed by a summary 48 of applicable portions of the Emergency Plan as they apply to the 49 Standard.
50 In addition, the Emergency Plan has been evaluated against the 51 requirements of Sections II.A.1.1 and III.A.1.2 of NUREG-0694 (now 52 superceded by the same sections of NUREG-0737).
53 Furthermore, California nuclear plant licensees and applicants 54 have been requested to provide analyses on the effects of ear'thquake 55 on their emergency plans, specifically in terms of the utilities' 56 capabilities to insure availability of personnel and equipment to 57 the sites.
58 Q.
Where are the results and conclusions of the NRC Staff's evaluation 59 reported?
60 A.
The NRC Staff's results and conclusions are reported in Appendix B 61 Emergency Preparedness Evaluation Report, to Supplement No. 14 to 62 the Safety Evaluation Report, April 1981.
63 Q.
What is the NRC staff conclusion?
64 A.
The NRC Staff conclusion is that the Diablo Canyon energency plan, 65 when revised in accordance with the commitments made, provides an 66 adequate planning basis for an acceptable state of emergency 67 preparedness and will meet the requirements of 10 C.F.R. 50 and 68 Appendix E.
- 4 69 Q.
Do you concur in the results and conclusions of the Emergency 70 Preparedness Evaluation Report?
71 A.
Yes.
72 Q.
Are there deficiencies in the Diablo Canyon Emergency Plan?
73 A.
Yes.
74 Q.
Has PG&E made commitments to correct these deficiencies?
75 A.
- Yes, PG&E has committed to correct these deficiencies before a 76 license for full power is granted.
77 Q.
What are these deficiencies?
78 A.
The principal deficiency is that a fast public alerting system is 79 not in place.
10 C.F.R. 50 Appendix E requires a demonstration of a 80 fast alerting system to the general public.
PG&E has described its 81 proposed system in Section 7.2.11 of the Emergency Plan.
PG&E has 82 contracted for the installation of 52 electro-mechnical sirens to 83 perform this function.
Installation will commence as soon as 84 necessary local permits are obtained.
85 Q.
What is the significance of this deficiency for low power operation?
86 A.
The requirement for a system to alert the public within 15 minutes 87 is based on an accident scenario which could resQut in an offsite 88 release of radiation in as little as thirty minutes from the time of 89 initiation of an accident.
The requirement is not necessary for 90 operation at 5% power because, even if a credible LOCA occured, the
-5 91 resultant heat up of the core to melting point would take at least 92 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and consequently a fast alerting system is not required.
93 Q. Is there a deficiency regarding a public information program?
94 A.
Yes, 10 C.F.R. 50 Appendix E requires the dissemination to the 95 public within the 10 mile Emergency Planning Zone of basic planning 96 information.
PG&E has a public information brochure in draft form.
97 The final version must be integrated with the emergency plans of 93 local governments which are presently being revised and upgraded.
99 Q.
What is the significance of this deficiency for operation to 5%?
100 A. The purpose of the requirement for a public information program is 101 to inform the public of how they will be alerted and what to do in 102 an emergency. Since the alerting system is not yet in place, there 103 is not yet a need for a public information program explaining the 104 system.
105 Q.
Are there other deficiencies in the Diablo Canyon Emergency Plan?
106 A. There are othe areas where the On-Site Plan is not in full 107 compliance with the elements of the guidance in NUREG-0654.
Many 108 have been addressed in the letter, PGE to NRC, dated January 13, 109 1931.
110 Q.
What is the significance of these deficiencies fs.
5% power?
111 A.
These deficiencies are not significant for operation at 5% power.
112 PG&E has committed in its letter to NRC, dated 2/27/81 that a full 113 power license NUREG-0654 will be substantially complied with.
-5 114 Q.
Will a supplement to the Emergency Preparedness Evaluation Report be 115 issued?
116 A.
Yes, a supplement to the Emergency Preparedness Evaluation Report 117 will be issued.
118 Q.
What will the supplement contain?
119 A.
The supplement will set forth the findings and determinations of the 120 Federal Emergency Management Agency (FEMA) as to whether State and 121 local emergency response plans are adequate and capable of being 122 implemented for full power operation.
123 Q.
Has FEMA made a finding for low power operation at Diablo Canyon 124 Power Plant?
125 A. Yes.
126 Q.
What was FEMA's finding for the adequacy of the present emergency 127 plans at Diablo Canyon?
128 A.
FEMA has made a finding that the present emergency plans at Diablo 129 Canyon adequately protect the public health and safety for the 130 purposes of low power testing.
131 Q. Where is FEMA's finding documented?
132 A.
FEMA's finding is documented in a Memorandum for arold R. Denton 133 and John W. McConnel from FEMA/NRC Steering Committee and in the 134 FEMA/NRC Interim Agreement on Criteria for Low Power Testing at New 135 Commercial Nuclear Facilities -- SER Supplements Nos. 10 and 12.
-7 136 Q.
How is PG&E complying with the NRC request that all California 137 nuclear plant licensees' and applicants provide analyses on the 133 effects of earthquake on their emergency plans, specifically in 139 terms of the utilities' capabilities to insure availability of 140 personnel and equipment to the sites?
141 A.
I have been informed-by PG&E that-a contractor, has been employed-to' 142 study the impact of earthquakes on the emergency plans of the 143 utility and of offsite authorities.
144 Q..
When will the contractor issue their report?
145 A. The report is due in mid-May of 1981.
146 Q.
How has PG&E proposed to utilize the contractor's report?
147 A.
PG&E will revise the Emergency Plans to include the contractor's 148 recommendations. This will be done prior to full power operation.
149 Q.
Are there presently emergency plans for the Diablo Canyon facility?
150 A.
Yes.
151 Q.
Does the licensee's plan presently have a notification requirement 152 as regards accidents at the facility?
153 A.
Yes.
154 Q.
What does the licensee's present emergency plan require?
- 8 155 A.
The licensee's present emergency plan requires prompt notification 1 ~
of the county governments surrounding Diablo Canyon Nuclear 157 Facility. Governor Brown has admitted this.
- 15.
Q.
Did Governor Brown admit anything else as regards emergency 159 planning?
160 A.
Yes, both Governor Brown and Joint Intervenors have admitted that 161 any accident occuring during low power testing would release a 162 fraction of the existing fission product inventory at the Diablo 163 Canyon Nuclear Facility.
164 Q.
Are the present emergency plans in full force and effect?
165 A.
Yes.
166 Q.
What is your opinion about the adequacy of Diablo Canyon's combined 167 Applicant, State and local emergency response plans if an accident 158 were to occur during low power operation?
169 A.
My opinion is that, if, during low power operation, an accident were 170 to occur, the combined applicant, State and local emergency response 171 plans will insure that an adequate state of emergency preparedness 172 is in place to minimize the risks of hazard to the health and safety 173 of the public during fuel loading and operation to 5% power.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATO11IC SAFETY AND LICENSING BOARD In the Matter of
)
PACIFIC GAS AND ELECTRIC COMPAqY
)
Docket Nos. 50-275 O.L.
)
50-323 O.L.
(Diablo CanyonNuclear-Power Plant )........
Unit Nos. 1 and 2)
JOHN iR. SEARS PROFESSIONAL QUALIFICATIONS EMERGENCY PREPAREDNESS PROGRAM OFFICE Prior to 1952, 1 was employed in field jobs in various aspects of mechanical engineering. In 1952, I joined Brookhaven National Laboratory as a Reactor Shift Supervisor on the Brookhaven Graphite Reactor. While at Brookhaven, I completed a series of courses given by the Nuclear Engineering Department in nuclear engineering. These courses were patterned on the ORSORT programs. In 1956, I was appointed Project Engineer on the Brookhaven Medical Research Reactor. I-was a member of the design group, participated in critical design experiments, wrote specifications, coauthored the hazards report, was responsible for field inspection and contractor liaison, trained operators and loaded and started up the reactor. About three months after start-up, in 1959, following the successful completion of proof tests and demonstration of the reactor in its design operating mode for boron capture therapy of brain cancer, I accepted a position as reactor inspectok with the Division of Inspection, U. S. Atomic Energy Commission. In 1960, I transferred, as a reactor inspector, to the newly-formed Division of Compliance.
I was responsible for the inspection, for safety and
compliance with license requirements, of the licensed reactors and the fuel fabrication and fuel processing plants,-which use more than critical amounts of special nuclear material, in the Eastern United States.
In September 1968, 1 transferred to the Operational Safety Branch, Directorate of Licensing. My responsibility included development of appropriate guides for-evaluation-of.-operationsl aspect of..li.cense.
applications and staff assistance in review of power reactor applicants submittals in the areas of Organization and Management. Personnel Qualifications, Training Programs, Procedures and Administrative Control, Review and Audit, Start-up Testing Programs Industrial Security and Emergency Planning.
The Branch was reorganized as the Industrial Security and Emergency Planning Branch in April 1974 to place increased emphasis and attention upon areas of physical security and emergency planning.
In 1976, I transferred to the Division of Operating Reactors as the sole reviewer responsible for review of emergency planning for all the operating reactors in the United States.
New York City College, 1950 - Mechanical Engineering Argonne International School of Reactor Technology, 1961 -
Reactor Control Course GE BWR System Design Course, 1972 Popo-U.S. Army, 1974 - Course in Industrial Defense and Disaster Planning Instructor at DCPA, 1976, 1977 - Course in Emergency Planning
-3 Director, 1962 - Reactor Program, Atoms for Peace Exhibit, Bangkok, Thailand Director, 1966 - Atoms for Peace Exhibit, Utrecht, Holland
0 0
UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COMMISSION 3
BEFOr THE ATOMIC SAFETY AND LICENSING BOARD 4
In the Matter of 5
No.5)25OL 6
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 0.L.
8 (Diablo Canyon Nuclear Power Plant 9
Unit Nos. 1 and.2).
10 TESTIMONY OF FRANK C. CHERNY 11 ON RELIEF, SAFETY AND BLOCK VALVES 12 AT DIABLO CANYON NUCLEAR FACILITY 13 Q.
Please state your full name.
14 A.
Frank C. Cherny.
15 Q.
By whom are you employed?
16 A. I am employed by the U.S. Nuclear Regulatory Commission. I am a 17 Section Leader in the Mechanical Engineering Branch, Division of 18 Engineering, Office of Nuclear Reactor Regulation.
19 Q.
Describe the nature of your work with respect to the Diablo Canyon 20 operating license proceeding.
21 A. Task Coordinator for NUREG-0737 Item II.d.1 "Performance Testing of 22 Boiling Water Reactor and Pressurized Water Ractor Relief and Safety 23 Valve. The Diablo Canyon SER input for qualification of safety and 24 relief valves, and block valves, as required by TMI Item I.D.1 was 25 prepared under my supervision.
26 Q.
Would you detail your professional qualifications?
27 A.
Attached is a copy of my professional qualifications.
0
0 0
-2 28 Q.
Have you, in the course of your professional experience, had 29 occasion to review the design and qualification of relief, safety or 30 block valves at Nuclear Power Plants other than Diablo Canyon?
31 A.
Yes, I have reviewed such designs or assisted in their review.
I 32 have also participated in many meetings, discussions, design reviews 33 of a generic nature, for the last five years, related to the design 34 of overpressure protection systems for nuclear power plants.
Many 35 such activities were related to direct participation in industry 36 standards writing activities associated with overpressure protection
/37 of nuclear reactor plant components and inservice testing of nucal r 38 power plant pressure relief devices.
39 Q.
In the course of your work on Diablo Canyon did you review the 40 design and qualification of Reactor Coolant system relief valves?
41 A. Yes.
42 Q.
In the course of your work on Diablo Canyon did you review the 43 design and qualification of Reactor Coolant system safety valves?
44 A. Yes.
45 Q.
In the course of your work on Diablo Canyon did you review the 46 design and qualification of Reactor Coolant system block valves?
47 A.
Yes.
48 Q.
Are there General Design Criteria (GDC) which must be met which 49 apply to relief and safety valves?
-3 50 A. Yes, GDC 1, 14, 15 and 30 require the Applicant to assess their 51 RCPB, including safety and relief valves, to meet certain standards.
52 Q.
What standards of review does the Staff use when reviewing for 53 compliance with GDC 1, 14, 15 and 30?
54 A. In reviewing for compliance with GDC 1, 14, 15 and 30 the following 55 standards are used: (a) Standard Review Plan (SRP) 3.9.2,. "Dynamic 56 Testing and Analyses of Systems, Components, and Equipment."; (b) 57 SRP 3.9.3, "ASME Code Class 1, 2 and 3 Components, Component 58 Supports, and Core Support Structures."; (c) Regulatory Guide 1.48 59 "Design limits and loading combinations for seismic Category 1 fluid 60 systems components."; and (d) Regulatory Guide 1.68 "Pre-operational 61 and Initial Startup Test Programs for Water Cooled Power Reactors."
62 Q.
Would you briefly describe what "Standard Review Plan (SRP) 3.9.2.,
63 "Dynamic Testing and Analyses of Systems, Components, and Equipment" 64 requires as regards reactor coolant system safety and relief valves?
65 A.
Standard Review Plan 3.9.2 requires a (1) piping vibration preop.
66 test program (2) seismic qualification of safety related mechanical 67 equipment (3) dynamic system analysis to insure structural adequacy 68 of piping loops for LOCA & SSE.
69 Q.
Would you briefly describe what SRP 3.9.3 "ASME Cde Class 1, 2 and 70 3 Components, Component Supports, and Core Support Structures" 71 requires as regards reactor coolant system safety and relief valves?
-4 72 A. Standard Review Plan 3.9.3 (1) specifies what load combinations and 73 stress limits apply (2) operability assurance program for "active" 74 pumps and valves (3) design of pressure relief valve supports and 75 associated discharge piping supports.
76 Q.
Would you briefly describe what Regulatory Guide 1.48 "Design limits 77 and loading combinations for seismic Category 1 fluid systems 78 components" identifies as a standard as regards reactor coolant 79 system safety and relief valves?
'80 A. Regulatory Guide 1.48 delineates acceptable design limits and load 81 combinations associated with normal operation and accident 82 conditions.
83 Q.
Would you briefly describe what Regulatory Guide 1.68 84 "Pre-operational and Initial Startup Test Programs for Water Cooled 85 Power Reactors" identifies as a standard as regards reactor coolant 86 system safety and relief valves?
87 A.
Regulatory Guide 1.68 requires pre-operational testing of safety and 88 relief valves to demonstrate that they will operate as required.
89 Q.
What remains to be done by the Applicant to comply with GDC 1, 14, 90 15 and 30?
91 A.
The tests performed to date do not cover loadinjs which result from 92 transition flow from steam to water or solid fluid flow.
-5 93 Q.
How will the loadings which result from transition flow or solid 94 fluid flow be addressed with respect to reactor coolant system 95 safety and relief valves?
96 A.
A test program has been initiated by the Electric Power Research 97 Institute (EPRI) which will address safety and relief valve 98 operability including loadings resulting from transition flow from 99 steam to water and solid fluid flow. PG&E has committed to 100 participating in this program and has as one of its objectives to 101 satisfy the long-term requirements on SRV testing as set forth in 102 Section 2.1.2 of NUREG-0578, "TMI-2 Lessons Learned Task Force 103 Status Report and Short-Term Recommendations" included as of October 104 31, 1978 as Item II.D. in NUREG-0737. The Applicant has referenced 105 the ongoing EPRI/NSAC safety and relief valve testing program. A 106 description of the EPRI program was provided to NRC by EPRI in 1979 107 and an updated reVision in July 1980. As noted in Supplement 14, 108 the Staff is generally in agreement that the NUREG-0737 technical 109 requirements for safety and relief valves can be met subject to 110 receipt of additional information which was requested by letter of 111 November 26, 1980 to Russell C. Youngdahl.
By letter of December 112 15, 1980 EPRI responded to both the Staff's November 26, 1980 letter 113 and NUREG-0737. PG&E has referenced the EPRI December 15 response 114 in their February 26, 1981 letter.
115 The Staff has not completed its review of tte December 15, 116 1980 letter primarily as regards EPRI proposed documentation 117 submittal dates for safety and relief valves and comments made in 118 the letter regarding resolution of a block valve test program after
-6 119 July 1, 1981, the scheduled completion date for safety and relief 120 valve testing.
121 Based on the Staff review of the EPRI program and PG&E's 122 assurance that the EPRI program is applicable to the Diablo Canyon 123 safety and relief valve designs, I believe there is adequate 124 assurance that the NUREG-0737. requirement regarding performance 125 verification of the RCS relief and safety valves will be met 126 satisfactorily for Diablo Canyon.
127 Should this program demonstrate that these valves are not 128 qualified for the above-stated loadings the Staff will require the 129 licensee to take corrective actions. Present schedules indicate 130 that this testing will be completed by July 1, 1981.
131 With regard to the safety valves, there is presently no 132 evidence that these valves will not operate properly during the 133 anticipated transients which produce transition flow and solid fluid 134 flow.
135 Q.
Other than with respect transition flow and solid fluid flow, has 136 the Applicant demonstrated compliance with the standards you 137 identified as necessary to meet GDC 1, 14, 15 and 30?
138 A.
Yes, compliance with Standard Review Plan (SPR) 3.9.2, "Dynamic 139 Testing and Analyses of Systems, Components, and Equipment," at 140 Diable Canyon which includes relief and safety Yalves is 141 demonstrated in the Safety Evaluation Report on Diablo Canyon (SER) 142 Section 3.9.1. and SER Supplements 7, 8, and 9.
-7 143 Compliance with SRP 3.9.3, "ASME Code Class 1, 2, and 3 144 Components, Component Supports, and Core Support Structures" at 145 Diablo Canyon, which includes relief and safety valves, is 146 demonstrated in SER Sections 3.9.2 and 5.2.1.
(SER and SER 147 Supplement 14).
148 Compliance with Regulatory Guide 1.48 "Design limits and 149 loading combinations for seismic Category 1 fluid systems 150 components" at Diablo Canyon, which includes relief and safety 151 valves, is demonstrated in SER Section 5.2.1.
(SER and SER 152 Supplement 7 & 14).
153 Compliance with Regulatory Guide 1.68 "Pre-operational and 154 Initial Startup Test Programs for Water Cooled Power Reactors" at 155 Diablo Canyon, includes testing of relief and safety valves, is 156 demonstrated in SER Section 14.
(SER).
157 Compliance with the appropriate sections of Appendix B to 10 158 C.F.R. Part 50 at Diablo Canyon, including relief and safety valves, 159 is demonstrated in SER Section 17.4.
(SER).
160 Q.
Is there anything, in your opinion, other than compliance with GDC 161 1, 14, 15 and 30, which demonstrates the reliability of the Reactor 162 coolant system relief and safety valves at Diablo Canyon?
163 A.
Yes.
In addition to complying with the requirements of GDC 1, 14, 164 15 and 30 as discussed above. The reactor coolant system safety 165 valves were originally designed and tested for operation on 166 saturated steam in accordance with the applicable edition and 167 addenda of Section III of ASME Boiler and Pressure Vessel Code.
- -8 168 Verification of this testing appears in the FSAR for Diablo Canyon, 169 Table 5.2-1.
As required by Article 9 of the ASME Boiler and 170 Pressure Vessel Code, the safety valve relieving capacity has been 171 provided so that the pressure limitation specified in the Code will 172 be maintained under all of the system transients and accidents 173 postulated to occu?.
174 Q.
Why do you believe that this further demonstrates the qualification 175 of the reactor coolant system and safety valves?
176 A.
Section III of the ASME Code provides specific valve functional 177 requirements and installation requirements for the reactor coolant 178 system safety valves. Additionally, it places restrictions on the 179 types of pressure relief valves that can be used for such 180 application and provides testing requirements for certifying the 181 relieving capacity of the safety valves.
182 Q.
Is there anything else you believe demonstrates the qualification of 183 the reactor coolant system and safety valves?
184 A.
Both safety valves and two of the three relief valves have been 185 seismically qualified to be functional after exposure to loads 186 resulting from the maximum hypothetical earthquake for Diablo Canyon 187 as documented in Amendment 50 to FSAR Table 7-7.
Also Safety and 188 Relief valves will be operationally tested during,he pre-op test 189 program performed in accordance with Regulatory Guide 1.68.
-9 190 Q.
Why do you believe that these facts further demonstrate the 191 qualification of the reactor coolant system relief and safety 192 valves?
193 A. The pre-op testing of safety and relief valves demonstrates the 194 operational readiness of the valve to lift within-the-prescribed-set 195 pressure range.
196 Seismic qualification assures that the valves will funGtion 197 should they be subjected to the maximum hypothetical earthquake 198 which was postulated for Diablo Canyon.
199 Q.
Are you familiar with I and E Bulletin 81-2?
200 A.
Yes.
201 Q.
Would you briefly summarize Bulletin 81-2?
202 A.
I and E Bulletin 81-2 is entitled "Failure of Gate Valves to Close 203 Against Differential Pressure".
The Bulletin discusses valve 204 closure tests recently performed by EPRI at the Marshall Test 205 Facility on seven gate valves of the type conmonly used as PWR PORV 206 Block Valves. The testing included closing the valve against full 207 flow steam differential pressure conditions selected as being 208 representative of those that a PORV might be expected to close 209 against. The Bulletin discusses the fact that three of the seven 210 tested gate valves failed to fully close when subjected to the test 211 conditions.
It notes that valves of the type that failed are also 212 supplied for utilization in a number of safety related applications
10 213 in addition to the PORV block valve application where closure with a 214 differential pressure across the valve is a requirement.
215 It requests that Licensees and Construction Permit Holders 216 determine whether any of the "failed" valves are installed in such 217 applications or planned to be installed.
If no valves in this 218 category are found this is to be reported to NRC. If one or more 219 valves of this type are identified in such an application or 220 intended for such an application, it imposes requirements for 221 specific actions to be taken by both Licensees and Construction 222 Permit Holders. Action to be taken include an evaluation of the 223 significance of the valve failure to close on system operability in 224 accordance with the plant technical specifications, modification of 225 valves so they are qualified for the intended service or obtaining 226 of qualified replacements.
227 Q.
What type of valves are present at Diablo Canyon?
228 A.
The Diablo Canyon plant has three Crosby HB-BP-86 (6M6) safety 229 valves, three Masoneilan 20,000 Series (2 NPS) Power Operated Relief 230 Valves and three Velan #B1O-3054B013M Motor Operated Block Valves.
231 Q.
Have any of these valves undergone the EPRI testing program?
232 A.
As of May 5, 1981 the following testing had been performed by EPRI 233 on valves of the type installed at Diablo Canyon:%
234 Power Operated Relief Valve - A Masoneilan 20,000 Series PORV 235 full flow steam test.
11 236
- B10-3054B013M Velan Motor Operated Block Valve - full flow 237 steam test. This valve is the same model as that used at 238 Diablo Canyon.
239 Q.
Were any of the block valves which failed the EPRI tests discussed k9 240 in I and E Bulletin 81-2-of the type to-be-used at.Diablo-Canyon?
241 A.
No.
242 Q.
What testjng remains to be completed on the valves at Diablo Canyon?
243 A.
Additional testing of the Masoneilan Power Operated Relief Valve to 244 include additional fluid effects that the valves could be exposed to 245 under design basis transient or accident events such as the effects 246 of loop seals and subcooled and saturated liquid is scheduled to be 247 completed by July 1, 1981.
248 Complete testing of the Crosby HB-BP-86 (6M6)
Safety Valve is 249 scheduled to be performed during June of 1981 with a scheduled 250 completion date of July 1, 1981.
251 The need for additional qualification testing of the Velan PORV 252 Block Valve, as of May 5, 1981, is under discussion between PWR 253 utilities and the NRC Staff. If additional testing is needed it 254 must be completed by July 1, 1982 as specified in NUREG-0737.
255 Q.
Will the remaining testing be completed prior to fuel load?
256 A.
The qualification testing of Safety Valves and PORV's of the type 257 installed on Diablo Canyon is scheduled to be completed by July 1, 258 1981, which will be well prior to fuel load.
-12 259 Q.
Have any Diablo Canyon valves failed during the testing program?
260 A. As of May 5, 1981, no valves of the type used in Diablo Canyon have 261 failed any of the EPRI tests.
262 Q.
What is done if a valve fails a test acceptance criterion?
263 A. EPRI has established a procedure so that all utilities participating 264 in the program, the NSSS vendors, the valve manufacturers, and the 265 NRC are all notified within a few days of any instances where a 266 valve fails a test acceptance criterion. The NSSS vendors, with 267 assistance from EPRI, assist the individual utilities with plant 268 specific evaluations of the safety significance of any such 269 failures. Depending on the results of these evaluations, actions 270 are taken by the utilities in accordance with the regulations as 271 regards reporting to NRC, possibly declaring equipment inoperable, 272 if installed on an operating plant, and modifications or 273 replacements of affected components for both operating plants and 274 plants like Diablo Canyon that have a Construction Permit. The 275 regulations also require the NSSS vendors and valve manufacturers to 276 report safety related equipment anomalies. Additionally, the Office 277 of Nuclear Reactor Regulation at the NRC independently reviews the 278 details of all reported failures on a case by case basis and a 279 decision is made as to what appropriate action should be taken.
280 For safety and relief valves of the type installed at Diablo 281 Canyon, as noted above, testing is scheduled to be completed by July 282 1, 1981.
If any failures of valves of this type occur in the EPRI 283 testing program, NRC will require the effects of the specific
9 S
13 284 failure on safe operation of the Diablo Canyon to be expeditiously 285 evaluated by PG&E on a schedule such that any necessary 286 modifications or replacements of safety or relief valves can be made 287 prior to initial fuel loading.
288 As noted above, it is not clear at this time whether additional, 289 testing is required to confirm the capability of the type of PORV 290 Block Valve installed at Diablo Canyon to open and close against all 291 fluid conditions that could result from design basis transients and 292 accidents. If more testing is required, as specified in NUREG-0737, 293 it must be completed by July 1, 1982.
294 If it is determined that additional testing of block valves is 295 required to confimn their performance capability, a procedure will 296 be established for expeditious handling of adverse test results.
297 Valve modifications or replacements, if any are warranted, will be 298 made to the Diablo Canyon PORV Block Valves on a schedule consistent 299 with the safety significance of any observed anomalies.
300 Q.
In view of the above testimony, do you have an opinion as to whether 301 fuel loading and low power testing can commence at Diablo Canyon 302 while PORV Block Valves remain to be tested?
303 A. As noted in this testimony, it has not been determined that 304 additional testing will be required to confirm the opening and 305 closing capability of the Diablo Canyon type block valves.
306 Based on the fact that the Diablo Canyon safety and relief 307 valves will be fully qualified prior to fuel loading for service 308 conditions far in excess of those conditions valves could be exposed
14 309 to during low power testing,- the other factors discussed aboe, and 310 on the testimony of Norman Lauben, it is my opinion that fuel 311 loading and low power testing can comnence at Diablo Canyon with no 312 adverse affect on the health and safety of the public.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMi1ISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 0.L.
)
50-323 O.L.
(Diablo Canyon Nuclear Power Plant )
Unit Nos. 1 and 2)
)
FRANK C. CHERNY PROFESSIONAL QUALIFICATIONS MECHANICAL ENGINEERING BRANCH DIVISION OF ENGINEERING I am a Mechanical Engineer in the Mechanical Engineering Branch responsible for the review and evaluation of design criteria of mechanical components, of methods of dynamic analysis and testing of safety related systems and components and of criteria for protection against dynamic effects associated with postulated failures of fluid system components for nuclear service.
I graduated from Marquette University with a B.S. degree in Mechanical Engineering in 1965.
From July 1965 to November 1968 I was employed by the Babcock &
Wilcox Co. at offices in both Barberton and Akron, Ohio. During the majority of this period I was engaged in materials engineering work, primarily writing technical ordering requirements for primary pressure boundary materials to be used for reactor vessels, steam generators, and presurrizers for both commercial and U. S. Navy nuclear systems.
In addition I had assignments of several months duration each in quality
-2 control engineering and nuclear stean supply system performance engineering.
From November 1968 to May 1974 1 was employed in the Pressurized Water Reactors Division of Westinghouse Nuclear Energy Systems. My work experience during this period includes the following:
From November 1968 to May 1970 and September 1970 to April 1971 as a Reactor Vessel Project Engineer based in Monroeville, Pa.:
(1) I had overall project engineer responsibility for design and construction of reactor vessels for several Westinghouse nuclear power plants in the U.S. My responsibilities included preparation of Design Specifications and review of vendor decumentation for compliance with Westinghouse, Utility, ASME, Architect Engineer and AEC requirements. I was personally responsible for coordination of the technical aspects of the transfer of two partially completed reactor vessels from a U.S.
manufacturer's shop to a European manufacturer for completion when schedular problems developed at the U.S. manufacturer.
(2) After the U.S.-Europe transfer of these components, I assumed responsibility for technical coordination between the primary Westinghouse nuclear engineering office in the U.S. and an overseas office established in Brussels, Belgium to do project engineering work for mechanical components used in Westinghouse nuclear plants both in the U.S.
and in Europe.b From June 1970 to September 1970 and from April 1971 to December 1972 I was employed by Westinghouse Nuclear Energy Systems in Europe based in Brussels, Belgium. My responsibilities included:
-3 (1)
During the June-September, 1970 period and from April 1971 to about April 1972 1 had project engineer responsibility for several reactor pressure vessels and a pressurizer. I also acted as Westinghouse engineering representative for U.S.
AEC Quality Assurance audits of European vessel manufacturers.
(2) From April 1972 to December 1972 I served as a lead engineer with a broader scope of responsibility. I was responsible for reactor pressure vessels, pressurizers and reactor vessel supports fabricated in Europe for Westinghouse Nuclear Plants.
Several engineers and a technician reported directly to me during this period. The work included preparation of Design Specifications, review and approval of vendor design and manufacturing documentation, and coordination with both U.S.
and European utility and regulatory representatives.
From December 1972 to May 1974 I was again based in Monroeville, Pa., this time as Senior Reactor Vessel Project Engineer. I was responsible for the technical adequacy of several reactor pressure vessels being manufactured in the U.S. for use in Westinghouse Nuclear Plants in Europe. I was also responsible, during the majority of this period, for the training of a Westinghouse Nuclear - Europe engineer temporarily based in the U.S.
In May of 1974 I started work for the Regulatory Division of the U.S. Atomic Energy Commission and have remained throug the transition into the U.S. Nuclear Regulatory Commission. In 1977 I was appointed as a Section Leader in the Mechanical Engineering Branch. As a branch technical reviewer and later as a Section Leader I have been
-4 participating in the review of constructionpermit and Operating License applications.
Since July of 1974 I have served as a member of the ASHE Section III Subyroup On Pressure Relief which is responsible for writing industry standards for the overpressure protection of light water reactor plant components. Additionally, since 1977 I have been a member of the ASME Working Group on Safety and Relief Valves. The Working Group ha.s recently completed work on a proposed industry standard entitled "Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." The proposed standard was issued for public comment by ASME early in 1931.
1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COM11SSION 3
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4
In the Matter of
)
6 PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos.
50-275 O.L.
7
)
50-323 O.L.
8 (Diablo Canyon 'Nucleaf Power Plant )
9 Unit Nos. 1 and 2)
)
10 TESTIMONY OF G. NORMAN LAUBEN.
11 Q.
Please state your full name.
12 A.
G. Norman Lauben.
13 Q.
By whom are you employed, and describe the work you perform?
14 A.
I am employed by the Reactor Systems Branch, Division of Systems 15 Integration, Office of Nuclear Reactor Regulation, U.S. Nuclear 16 Regulatory Commission. A copy of my professional qualifications is 17 attached to this testimony.
18 Q.
Have you read "Memorandum and Order (Granting PG&E's and NRC Staff 19 Motions for Summary Disposition of Joint Intervenor's Contentions 5 20 and 13; Denying their Motions as to Contentions 4 and 24)"? In 21 particular, have you read Contentions 4 and 24 as stated therein?
22 A.
Yes.
23 Q.
Would you describe the scope of the subject matter addressed in your 24 testimony?
25 A.
I have been asked to address the safety significance of the low 26 power testing program as it would affect the necessity to have in
-2 27 place an emergency plan which meets all the requirements of 28 NUREG-0654.
I have also been asked to address the safety 29 significance of block valve failure during the low power testing 30 program.
31 Q.
What are the major factors affecting safe reactor operation during 32 low power testing?
33 A.
In supplement 10 to the Diablo Canyon SER (NUREG-0675), the risk 34 associated with low power testing was addressed.
As discussed in 35 that SER there are three major factors which contribute to a 36 substantial reduction in risk for low power testing as compared to 37 continuous full power operation.
First, there is additional time 38 available for the operators to correct the loss of important safety 39 systems needed to mitigate relatively high risk events, or to take 40 alternate courses of action. Secondly, supplement 10 also discussed 41 the significant events that contribute to that risk and the 42 reduction in risk associated with the significant postulated events 43 during the low power testing program. Third, there is a reduction 44 in required capacity for mitigating systems at low power.
45 Q.
What are the significant postulated events that could potentially 46 affect the public health and safety?
47 A.
The dominant events are (1) small break LOCAs with loss of the 48 emergency core cooling system (ECCS),
(2) transients with total loss 49 of feedwater, and (3) failure of double check valves between the 50 reactor coolant system (high pressure) and the residual heat removal
-3 51 system (low pressure) which results in a LOCA (inter-system LOCA) 52 outside containment, i.e. the interior of the reactor vessel 53 communicates directly with the environment.
Since the publication 54 of the Reactor Safety Study (WASH-1400), the NRC staff has continued 55 to study the risk to the public from potential severe accidents at 56 nuclear power plants. This effort has confirmed that the event 57 scenarios do-minating accident risks are generally the same for 58 different PWR designs, i.e., small break loss of coolant accidents 59 (LOCAs) and transients.
60 Q.
What was done specifically to address the question of risk due to 61 low power testing, and what was the conclusion of that effort?
62 A.
We have reexamined the dominant scenarios to estimate the reduction 63 in the probability of the event because of the additional time 64 available during low power operation for the reactor operators to 65 correct the loss of important safety systems needed to mitigate the 66 event or to take alternate courses of action. Similarly, we have 67 calculated the reduced fission product inventory for operation of an 68 initially unirradiated core at 5% power for 6 months and have 69 determined the reduction in potential public exposure via reduction 70 in potential release magnitudes. Risk is roughly proportional to 71 the probability of severe accidents (which lose the heat sink) and 72 to the fission product inventory in the core.
From these factors we 73 have estimated that the overall reduction in risk to the public
- 4 74 should be a factor of 400 to 1500 if a..plant is operated at 5% power 75 from initial startup for 6 months compared to continuous full power 76 operation.
77 Q.
In your review of fuel load and low power test programs what have 78 you concluded about the potential risk of low power testing?
79 A. Based on the actual power history of other reactors during their low 80 power testing program, the actual power history expected at Diablo 81 Canyon would result in even less available fission product 82 inventory. The peak power during this time period is only expected 83 to be 3 or 4% of rated capacity. Operation at this power level is 84 only expected for a maximum of 10 days.
This would result in a 85 further risk reduction by a factor of about 2. It is therefore 86 concluded that the public risk due to fuel loading and the proposed 87 low power test program is less than public risk due to full power 88 long-term operation by a factor of about 400 to 3000.
89 Q.
How could risk to the public be affected by small break or inter 90 system LOCA's?
91 A.
Risk to the public would occur only if there is release of 92 substantial amounts of radioactive fission products outside the 93 containment. This could occur only if there is a failure to cool 94 the core for an extended period of time. During this time the fuel 95 element cladding would have to fail by overheating. The reactor 96 coolant pressure boundary would have to be violated and the reactor 97 building containment would have to be violated. Of course the
-5 98 latter condition would occur for the intersystem LOCA. The 99 important factor is that the core would have to renain uncooled for 100 a significant length of time.
(Approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> during the 101 low power test.prograri).
102 However, NRC requires that all.facilities licensed to operate.
103 are provided with reliable and redundant emergency core cooling 104 systems (ECCS).
NRC regulations (10 C.F.R §50.46) require 105 applicants to analyze a spectrum of pipe breaks and locations with 106 various assumed equipment failures. These analyses are performed 107 with NRC-specified conservative assumptions and must demonstrate 108 coolability of the core and minimum generation of hydrogen.
109 Specifically, ECCS evaluations must demonstrate that ECCS 110 performance will result in a coolable geometry and less than 1%
111 core-wide metal-water reaction even with the reactor at 102% power 112 and worst-case linear heat rates. The Diablo Canyon ECCS is 113 required to conform to these requirements. Thus, for all power 114 levels the requirements of NRC regulations provide adequate 115 protection against severe core damage. Thus for small break LOCAs, 116 substantial risk to the health and safety of the public would occur 117 only if the ECCS failed to operate as designed. Therefore, when 118 performing risk assessment this condition must be evaluated.
119 Q.
With a reactor operating at a maximum of 5% of full power, could 120 these events lead to the significant amounts of core damage?
121 A.
No. I have looked at these events and have concluded, as discussed 122 herein, that at 5% power it is extremely unlikely that such events
6 123 would lead to significant amounts of core damage by "significant" I 124 mean 5% metal-water reaction.
125 Q.
Please provide the basis for your conclusions?
126 A. LOCA-analyses with-severly degraded.ECCS's were performed which 127 demonstrate the large amount of time that would be available at 5%
128 power for diagnosis and corrective action to prevent significant 129 core damage. For these analyses, it was conservatively assumed that 130 none of the pumped ECCS systems was functioning. I have had Sandia 131 Laboratories, our consultant, run a calculation for me using the 132 RELAP4 code to estimate the tine at which boil-off begins.
For a 133 small 4-inch cold-leg break LOCA, boil-off would not begin for about 134 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and uncovery would be delayed until about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Rapid 135 increase in fuel clad temperature, leading to severe core damage, 136 would not begin until about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
137 I have also obtained information from the NRC staff Project 138 Manager for the Sequoyah nuclear power plant concerning the actual 139 maximum power level and test duration for the low-power test 140 program. Sequoyah is a comparable facility and the test program is 141 similar to that planned for Diablo Canyon. This information 142 indicates that actual test power was about 4% full power or less and 143 lasted only some 8 days.
For similar conditions at Diablo Canyon 144 uncovery would not occur until about 4 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. More than 20 145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br /> would elapse before significant core damage would occur.
-7 143 Q.
Based upon this assessment, what is the likelihood of significant 147 core damage at low power due to a LOCA?
148 A.
As I have indicated above, with the NRC's requirements for reliable 149 ECCS performance, the potential for severe core damage and 150 associated significant hydrogen generation is very small even at 151 full power. The time available at low power for the operator to 152 take corrective action ranges from 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> to more than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> in 153 the event of a small LOCA. In addition, the coolant flow required 154 to dissipate decay heat at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> following a LOCA would be only 155 about 8 gpm which is within the capacity of the centrifugal charging 156 pump used for the normal make-up systems. Because of the time 157 available for the operators to correct malfunctions in the ECCS or 158 to initiate cooling with the normal charging system, we believe that 159 the probability of a small LOCA resulting in excessive fuel damage 150 and significant radiological release is reduced by at least a factor 161 of 400 to 1600 for low power operation as compared to operation at 162 full power.
163 Q.
What about the potential effects of other significant events, e.g.,
164 transients with total loss of feedwater, on the public risk?
165 A.
At 5% power the number of events that can result in failure to 166 adequately cool the core is greatly reduced.
All transients 157 initiated by turbine trip are eliminated since the turbine is not on 168 line. Total loss of feedwater caused by any other transient becomes 169 negligible with respect to core damage and public risk.
In such a 170 case, core heat is transferred through the steam generators from the
171 primary to secondary systems.
After a scram from 5/ power I 172 calculate that it would take about 2 days to boil the steam 173 generators dry, conservatively assuming no feedwater makeup. During 174 that period of time, diagnosis, corrective action or alternate heat 175 removal methods could easily be accomplished. Moreover, by that 176 time fission product heat will have decayed sufficiently so that 177 passive steam heat losses (radiant heat transfer) would be e.nough to 178 keep the reactor cool, even if no corrective action were taken. As 179 a consequence we believe that the total risk reduction associated 180 with these events is about 1000 to 20,000 lower compared to full 181 power operation.
182 Q.
If, as you say, feedwater transients are of almost no concern at low 183 power, what about other scenarios? Could they not now become 184 dominant at low power?
185 A.
Other transients (steam line break, steam generator tube rupture, 186 rod ejection and ATWS) were also examined. A similar reduction in 187 risk was evident. Therefore, these transients did not become 188 dominant.
189 Q.
But an ATWS event could result in a significant primary to secondary 190 heat imbalance. Why isn't this of concern?
191 A.
If one assumes that the highly unlikely scenario f total loss of 192 feedwater is followed by a failure.of the reactor system to scram 193 (i.e., the worst ATWS event), complete boil-off of the water in the 194 steam generators would occur in 45 minutes.. During this period of
-9 195 time there are a number of things the operator could do to bring the 196 reactor to safe shutdown, including initiation of the boron 197 injection system and diagnosis and correction of the failure to 193 scram. These would terminate the event before boil-off of 199 significant reactor.vessel-inventory and thus, well before the onset 200 of severe core damage. Moreover, at low power, significant 201 overpressurization of the primary system does not occur because of 202 the low integrated reactor power. It should be emphasized that the 203 sequence which involves a transient, total loss of feedwater, and 204 failure to scram as independent events has such a low probability 205 that it cannot be considered a credible event (less than 10-7 per 206 reactor year).
207 Q.
How does the above discussion relate to the need for a qualified 208 emergency plan during low power operation?
209 A. The above discussion shows that abundant time (at least 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) is 210 available to take corrective action to mitigate or terminate the 211 most likely scenarios which could affect public risk during low 212 power testing.
For some sequences of concern at full power, no 213 action would be required during low power operation to prevent 214 public risk.
Under these conditions the risk is so small that there 215 is virtually no need for a qualified emergency plan.
216 Q.
What is the safety significant of a failure of a block valve during 217 low power testing?
10 218 A. None. Each block valve is located upstream of one PORV. There are 219 three sets of these valve pairs in parallel. In theory a block 220 valve could fail open or closed. During operation the block valves 221 are normally open and the PORVs are closed. If the PORVs are not 222 cha.11.enged the position of the block valves has no significance.
223 PORVs can be challenged only if a heat imbalance occurs. As 224 discussed earlier the transient events that could cause this 225 condition are those for which secondary heat sink is lost when 226 feedwater is lost. At low power these events have virtually no 227 safety significance and PORV challenge is extremely unlikely.
228 Q.
But what if a block valve and a PORV should both fail open?
229 A. The likelihood of this double failure is extremely remote. As noted 230 in the testimony of Mr. Frank Cherny, the Diablo Canyon type PORVs 231 will be tested in the EPRI program prior to fuel loading. If this 232 should occur it would simply be a small LOCA which was discussed 233 previously. The size of this LOCA would be even smaller than that 234 previously discussed. Thus even more time would be available to 235 correct, mitigate or avert fuel damage and public risk.
- 11 236 STATEMENT OF PROFESSIONAL QUALIFICATIONS 237 NORMAN LAUBEN 238 My name is George Norman Lauben.
I am employed as a Nuclear 239 Engineer in the Reactor Systems Branch, Division of Systems Integration, 240 U.S.
Nuclear Regulatory Commission.
I have worked in the field of 241 nuclear reactor for 19 years, and in nuclear activities for 23 years.
I 242 have worked for the Commission and its predecessor, the Atomic Energy 243 Commission, since 1968. During this time I have worked directly on 244 reactor safety matters, including Emergency Core Cooling System (ECCS) 245 performance review and Loss-of-Coolant Accident (LOCA) analysis.
246 I was a member of the 1971 AEC ECCS task force and the AEC Staff 247 Panel for the ECCS Rulemaking Hearing. I an the author of the TOODEE2 248 computer program used by the NRC and the nuclear industry for transient 249 fuel pin thermal analysis during a LOCA. I was a member of the technical 250 team that accompanied Mr.
Harold Denton to the Three Mile Island Reactor 251 on March 30, 1979.
252 I have a B.S. and M.S. in Chemical Engineering from Case Institute 253 of Technology (now Case Western Reserve University).
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 O.L.
)
50-323 O.L.
(Diablo Canyon Nuclear Power Plant
)
Unit Nos.
1 and 2
)
CERTIFICATE OF SERVICE' I hereby certify that copies of TESTIMONY OF BARTHOLOMEW C. BUCKLEY, JOHN SEARS, FRANK C. CHERNY, AND NORMAN LAUBEn in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 8th day of May, 1981.
John F. Wolf, Esq., Chairman Richard E. Blankenburg Atomic Safety and Licensing Board Panel Co-publisher U.S. Nuclear Regulatory Commission Wayne A. Soroyan, News Reporter Washington, D.C.
20555
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20555
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, U.S.. Nuclear Regulatory Commission Dr. Jerry Kline Washington, D.C. 20555
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-2 John R. Phillips, Esq.
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20555.
State Capitol Sacramento, California 95814 Mark Gottlieb California Energy Commission David S. Fleischaker, Esq.
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