ML20150C330
| ML20150C330 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/15/1978 |
| From: | Schackleford B PACIFIC GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML16340A285 | List: |
| References | |
| NUDOCS 7811220182 | |
| Download: ML20150C330 (286) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
/
( In the Matter of )
) Docket No. 50-2 7 5-O L PACIFIC GAS AND ELECTRIC COMPANY ) Docket No. 50-323-OL
)
Units 1 and 2 - Diablo Canyon Site ) AMENDMENTS NOS. 72 AND 73
)
Pacific Gas and Electric Company hereby submits Amendments Nos. 72 (Hosgri) and 73 (FSAR) to its application for an operating j license for Units 1 and 2 at its Diablo Canyon Site. These amend-ments include material for Chapters 2, 8, 9, and 10 of the FSAR, and Chapters 7, 8, 10, 11, and new Chapters 12A and 12B of the Hosgri Seismic Evaluation. For a brief description of the changes made by these amendments, see the Summary of Amendment 72 and the Summary of Amendment 73 which precede the Removal and Insertion Instructions for the respective amendments.
O
\'- Subscribed in San Francisco, California, this 15th day of November, 1978.
Respectfully submitted, PACTFIC GAS AND ELECTRIC COMPANY By BARTON W. SHACKELFORD Barton W. Shackelford Executive Vice President JOHN C. MORRISSEY MALCOLM H. FURBUSH PHILIP A. CRANE, JR.
Attorneys for Pacific Gas and Electric Company By PHILIP A. CRANE, JR. 112to/ &
Philip A. Crane, Jr.
Subscribed and sworn to before me this 15th day of November, 1978 THEODORA COOKE (gggt)
('%)
(_f Theodora Cooke, Notary Public in and for the City and County of San Francisco, State of California My Commission expires January 28, 1981 i
MfENDMENT 72 INSTRUCTION SHEET (File this instruction sheet in the front of Volume 1 as a record of changes.)
The following instructions and check list are provided as a guide for the insertion of new pages for Amendment 72 to the Seismic Evaluation for Postulated 7.5M Ilosgri Earthquake, of the operating license application for Units 1 and 2 Diablo Canyon Site. The new pages are marked " Amendment 72" I and "(November 1978)" and contain both amended and supplementary material.
This material is indicated by a vertical bar with the figure "72" inscribed in the adjacent margin of the page. Where such marks appear adjacent to a blank portion of a page, a deletion is indicated. Where pages have been changed only to reposition material, with no change in content, only the amendment number and the date are given.
For a brief description of the changes made by Amendment 72, see the SLReiARY of AliENDMENT 72 which precedes the " REMOVAL-INSERTION" INSTRUCTIONS.
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SUtR1ARY OF AffEND? TENT 72 Location of Change Comment s
Table of Contents Changes reference to Chapter 12 to Chapter 12A (Buried Piping Systems; adds reference to Chapter 12B Buried Tanks).
7-1 Adds Section 7.6.2 Laboratory Testing 7-11 Adds Tabic 7-11, List of Mechanical Equipment Seismically Tested 7-9 Changes minor description of analysis of auxiliary feedwater pump turbine.
7-10 Updates discussion of testing of radiator housing.
7-11 Adds to description of boric acid tank model.
7-12 Minor grammatical correction.
7-13 Minor format correction.
7-14a Minor format correction.
7-17 Revises discussion of additional supply of auxiliary feedwater.
7-18 Adds an update of discussion of In-Plant Testing 7-19 Adds an update of discussion of Laboratory Testing.
7-20 through 7-22 Adds new description of Laboratory Testing.
Table 7-6, Sheet 2 Updates to reflect requalification Table 7-3, Sheet 3 Updates table, and 6 Table 7-7, Updates values.
Sheets 1 and 2 Table 7-7A Updates values.
Table 7-8, Updates values and notes.
Sheets 1 through 7 and Notes .
Table 7-11 Adds new table of List of Equipment Seismically Tested.
8-9 Updates discussion referring to Table 8-3, Stress Evaluation of Piping Systems.
8-13 Updates discussion of piping evaluation.
Table 8-3, Updates values to reflect latest evaluation.
Sheets 1 through 5 and 7 through 17 l
t _ _ _ _ _ _ _ _ _ _ - - _ - . _ _ _
- _ _ -. _ - . . . - _. . . _ __. _.. . ~ . _ _ _ _ _ . _ . _
SUtttARY-0F AMENDMENT 72 (Continued)
Location of Change Comment Table 10-1, Updates to reflect requalification.
Sheets 1 through 3 Table 10-11 Adds a new table on Vital Load Center Auxiliary Relay Panel 11-1 through Complete revision of Chapter 11.
Figure 11-14 12A-1 through Complete revision of Chapter 12 into Chapter 12A.
Table 12A-1 d
12B-1 through Completely new Chapter.
1 Figure 128-12 '
13 1 through Updates to reflect latest modifications.
13-6 NEW BINDER For casing the binders and to add the new sections, arrange the volumes so that they contain the pages indicated below.
1 Volume VII (New Binder) Appendix D, Tab 41 through Appendix L Page APP L-14.
(Hosgri Report)
Volume VI Appendix D, Tab 15 through Appendix D, Tab 40.
Volume V Appendix B through Appendix D, Tab 14.
Volume IV Chapter 10 through Appendix A, Page A-24.
Volume III Chapter 5 through Chapter 9, Table 9-1 Sheet 2 of 2.
Volume II Figure 4-1 through Figure 4A-38.
Volume I Chapter 1 through Chapter 4. Table 4-53.
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REMOVAL-INSERTION INSTRUCTIONS REMOVE INSERT AMENDMENT 72 tiATERIAL 1
Volume I Table of Contents Table of Contents Volume III 7-1 7-1 7-11 7-11 7-9, 7-10 7-9, 7-10 ,
7-11 7 il ,
7-12 7-12 7-13 7-13 7-14a 7-14a 7-17 7-17 7-18 7-18 7-19 7-19 None 7-20 None 7-21 None 7-22 Table 7-3, Sh. 3 & Sh. 6 Table 7-3, Sh. 3 & Sh. 6 Table 7-6, Sheet 2 Table 7-6, Sheet 2 Table 7-7 Table 7-7, Sheets 1 and 2 Sheets 1 and 2 Tabic 7-7A Table 7-7A Table 7-8, Table 7-8, Sheets 1 through 7 Sheets 1 through 7 ,
Notes for Table 7-5, 7-6, 7-7, Notes for Table 7-5, 7-6, 7-7, t
& 7-8 & 7-8 '
, None Table 7-11 8-9 8-9 8-13 8-13 Table 8-3, Table 8-3, Sheets 1 through 5 and Sheets 1 through 5 and 7 through 17 7 through 17 Table 10-1, Table 10-1, Sheets 1 through 3 Sheets 1 through 3 None Table 10-11 Complete Chapter 11 New Chapter 11 Complete Chapter 12 New Chapter 12A and 12B i
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REMOVAL-INSERTION INSTRUCTIONS .
(Continued)
REliOVE INSERT AMENDMENT 72 ffATERIAL l
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13-3 13-3
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SEISMIC EVALUATION FOR 7.5M HOSGRI EVENT fg Table of Contents Chapter 1 Executive Summary Chapter 2 Review of Original Seismic Design for Diablo Canyon Chapter 3 Characteristics of Hosgri Event Chapter 4 Structures Chapter 4A Cranes Chapter 5 Basis for llosgri Evaluation of Mechanical Equipment Chapter 6 Reactor Coolant System Chapter 7 Mechanical Equipment Chapter 8 Other Design Class i Piping Systems Chapter 9 Safeguards Ventilation Chapter 10 Electrical Equipment and Instrumentation Chapter 11 Outdoor Tanks Chapter 12A Buried Piping Systems Chapter 12B Buried Tanks Chapter 13 Modifications Appendix A Investigation of Effect on Seismic Design Using
\- Parkfield-5, 1966, N85E and Castaic, 1971, OfCE Records (scaled to 0.5G Peak Acceleration) As Input Appendix B Integrity of the Primary Piping Systems of the Diablo Canyon Nuclear Power Plants During Postulated Seismic Events Appendix C Seismic Analyais Methods Appendix D Studies in Support of the llosgri Seismic Evaluation Criteria and Responses to Comments and Recommendations by the ACRS and its Seismic Consultants Appendix E Responses to SEB questions on Hosgri Report: NRC Letter dated 8/24/77.
PCandE Letters (Index items 5, 6, 7) 11/2/77; 12/1/77; 12/20/77; 2/2/78 (Amendment 59).
O d
(November 1978) 1 Amendment 72
Table of Contents (cont)'
-"s Appendix F Responses to MEB questions on Hosgri Report: NRC Letters dated 11/10/77, 11/15/77; phone call 1/30/78.
PCandE Letters (Index items 1, 2, 3, 4) 11/15/77; 12/30/77; 2/14/78; 2/23/78; 4/12/78; 5/11/78 (Amendments 59, 60, and 66).
Appendix G Responses to NRC Letter dated 11/10/77, about combined LOCA/ Asymmetric Load Analysis PCandE Letters (Index items 1, 15) 2/6/78; 2/22/78 (Amendments 60 and 66).
Appendix H Response to NRC Letter dated 11/22/77, about Load l Combinations Outside RCS.
1
- PGandE Letter (Index item 8) l l 1/18/78 (Amendment 60).
I Appendix I Response to NRC Letter dated 12/27/77, about Dumping.
1 PGandE Letter (Index item 9) 1/17/78 (Amendment 60).
Appendix J Response to NRC Letter dated 12/12/77, about Systems for Safe (Cold) Shutdown.
O PGandE Letters (Index item 12) 1/26/78; 4/17/78; 5/2/78 (Amendments 60 and 66).
Appendix K Response to informal staff request about operating basis carthquake.
PGandE Letter (Index item 14) 4/11/78 (Amendment 66).
Appendix L Incorporates into Hosgri Report letter dealing with DCNP Cranes.
PGandE Letter 5/26/78 (Amendment 70)
Note: A detailed table of contents appears at the beginning of each chapter.
O Amendment 72 11 (hcaember 1978)
CHAPTER 7 O MECHANICAL EQUIPMENT INDEX 7.1 Loading Combinations and Criteria 7.2 Pump and Valve Operability Assurance i
7.3 Equipment Evaluated 7.3.1 Minimum Equipment Needed following an Earthquake 7.3.2 Additional Items Needed in Case of Single Failure l
7.4 Discussion of Results 7.4.1 Auxiliary Feedwater 7.4.2 Emergency Diesel Engine Generator System 7.4.3 Charging and Boration O 7.4.4 Heat Removal V
7.4.5 Instrumentation 7.4.6 Qualification of Valves 7.5 Other Equipment Evaluated 7.6 Testing 7.6.1 In-Plant Testing 7.6.2 Laboratory Testing 72 (November 1978) 7-i Amendment 72
INDEX OF TABLES 7-1 Hosgri Seismic Evaluations Loading Combinations and Structural Criteria, Mechanical Equipment 7-2 Hosgri Seismic Evaluation Loading Combinations and Structural Criteria, Mechanical Equipment Supports 7-3A Minimum Required Active Valves for Hot Shutdown and/or Cold Shutdown 7-3B Valves Required in Case of Single Failure 7-3C Active Valves Required for Normal Hot Shutdown and/or Normal Cold Shutdown 7-4 Deleted 7-5 Summary-Seismic Qualification of Class 1 Equipment Required Following Hosgri Event 7-5A Summary-Seismic Qualification of Class 1 Equipment Needed in Case of Single Failure 7-6 Summary-Seismic Qualification of Additional Class 1 Equipment Not Required Following Hosgri Eunt 7-7 Summary-Seismic Qualification Minimum Required Active Valves for Hot q Shutdown and/or Cold Shutdown
,V 7-7A Summary-Seismic Qualification of Class 1 Valves Needed in Case of Single Failure 7-8 Summary-Seismic Qualification Additional Class 1 Valves Not Required Following Hosgri Seismic Event 7-9 Comparison of Natural Frequencies From Analysis and Testing 7-10 List of Valves Statically Tested at the Diablo Canyon Site 7-11 List of Mechanical Equipment Seismically Tested 72 O
V (November 1978) 7-ii Amendment 72
- g. 7.4 DISCUSSION OF RESULTS V
This section presents the seismic design qualification and the Hosgri evaluation of the mechanical conponents identified in Section 7.3.1 as the minimum required following an earthquake. Table 7-5 summarizes the results of the Hosgri evaluation for equipment needed following an earthquake.
Section 7.5 presents the qualification and evaluation of other Class I mechanical components.
7.4.1 AUXILIARY FEEDWATER The design qualification analysis of the turbine-driven auxiliary feedwater pump was done for a DDE horizontal acceleration approximately 85L of that for the Hosgri event. The Hosgri evabation was done by analysis for earth-quake motion, plus nozzle, pressure, thermal, and thrust loads. Nozzle and holddown bolt stresses were found to be within allowable values. Shaft deflection was calculated and it was shown that a clearance is maintained.
The auxiliary feedwater pump turbine was analyzed for DDE accelerations greater than those for the Hosgri event. This analysis included the governor control valve (FCV-15) and the turbine stop valve (FCV-152) which were found to be acceptable. Shaft bending stresses and deflections were found to be small. Calculated thrust bearing and journal bearing loads were less than one-third of allowable. Stresses in the holddown bolts also were low. The turbine was therefore not reanalyzed for the Hosgri.
7.4.2 EMERGENCY DIESEL ENGINE GENERATOR SYSTEM The emergency generators previously were analyzed and also were shock tested per MIL-S-901. The analysis at the DDE acceleration (about 74% of Hosgri) showed stresses in bolts, engine feet and skid beams to be within allowable values. Stresses in various component mountings (radiator and lube oil units) also were satisfactory. In the shock tests, the same type of engine was subjected to a series of five simulated underwater explosions, while mounted rigidly on a floating platform. During these tests, accelerations reached 55g (fifty-five g). During three of the tests the engine was in operation. No apparent damage or malfunction of the engine was noted as a result of these tests.
(November 1978) 7-9 Amendment 72
A more recent analysis showed that the units would be satisfactory for a horizontal acceleration of 0.50g and a vertical acceleration of 0.33g, if 72 the flexible supports under the diesels were removed and the diesels were rigidly fastened to the floor. This has been done.
As part of the Hosgri evaluation, in-place dynamic testing was done.
Analysis based on the characteristics determined in the test showed that 72 the radiator housing was rigid and that all members were within allowable .
l stresses.
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Miscellaneous diesel components, such as the diesel fuel filter, strainer, l transfer pump and priming tank, and the diesel starting air receivers also were analyzed previously, some at acceleration levels as high as Hosgri.
As a result, the priming tank was provided with a seismic brace. After this change, all these components showed such low stresses that additional analysis for the Hosgri was not necessary.
1 O 7.4.3 CHARGING AND BORATION V
The natural frequencies of the centrifugal charging pump assembly were determined by dynamic testing performed by the vendor. The results of the test agreed with the dynamic modal analyses previously performed on a comparable pump. In both cases, the first natural frequency of the pump assembly was shown to be greater than 33 Hz and the pump can be considered to be rigid. Qualification was acconplished by detailed comparison to the analyses performed on the comparable pump more recently analyzed. The pump was qualified for the simultaneous application of umbrella seismic accel-erations of 39 horizontal and 2g vertical which are significantly higher than those for the Hosgri. Stresses in pump and motor holddown bolts, anchor bolts, shaf ts, and nozzles were evaluated. Rotor and stator clear- l ances, pump and motor bearing loads, shaft alignment and impeller clearance were shown to be within vendor specified levels. The most critical areas I of evaluation were the pump discharge nozzlc and head flange which were both stressed to less than 85% of allowable.
O (November 1978) 7-10 Amendment 72
Summarily the centrifugal charging pumps have been shown to satisfy structural (Section 7.1) and operability qualification (Section 7.2)
O Q requirements, therefore demonstrating their functional capability for the Hosgri earthquake.
The boric acid tanks have been reanalyzed using a refined model which reproduces frequency, damping, mode shape data determined by field test, and y sloshing of the contents of the tanks. The model shows that stresses in the j anchor bolts and in the tank skirt, which are the most critical stresses for I these tanks, are within the allowable stresses, j The boric acid transfer pump assembly was qualified by comparison to an identical pump evaluated recently. Dynamic modal analyses were performed on the reference pump to determine natural frequencies. Computer analyses were performed with the ICES-STRUDL code. The results of these analyses showed the first natural frequency of the pump assembly to be greater than 33 Hz, and the pump was analyzed as a rigid component. The pump was qualified for the simultaneous application cf umbrella seismic accelerations of 39 horizontal and 29 vertical which ar2 significantly higher than those for the Hosgri. Stresses in pump and motor holddown bolts, anchor bolts, shafts, and nozzles were evaluated and shown to be below allowables. Rotor and stator clearance % pump and motor bearing loads, shaf t alignment, and impeller clearance were shown to be within vendor specified levels. The most critical area of evaluation was the suction nozzle where stresses were 97% of allowable. This evaluation was based on umbrella nozzle loads which were significantly higher than those for the Hosgri.
4 Summarily, the boric acit transfer pump was shown to satisfy structural (Section 7.1) and operability qualification (Section 7.2) recairements, therefore demonstrating its functional capability for the Hosgri earthquake. l The natural frequencies of the regenerative heat exchanger were determined by dynamic modal analysis with the WECAN code. The first two natural frequencies of the heat exchanger were determined to be 20,6 and 22.0 Hz; all other frequencies were above 33 Hz. The regenerative heat exchanger was evaluated for the simultaneous application of deadweight, pressure, O
Q seismic, and nozzle loads.
(November 1978) 7-11 Amendment 72
Included in the evaluations were nozzles support-to-shell attachments, supports, and support-to-foundation interfaces. These evaluations included bolts and welds.
Analysis for the simultaneous application of 0.979 horizontal and 0.47g vertical accelerations initially showed the U bolts in the support system 72 to be overstressed. Modificat'ons which are adequate for the Hosgri
- loads are discussed in Chapter 13.
1 The natural frequencies of the seal water injection filter were determined by dynamic modal analysis with the WECAN computer code. The first natural frequency of the heat exchanger was determined to be 26 Hz; all other frequencies were above 33 Hz. The seal water injection filter was evaluated for the simultaneous application of deadweight, pressure, seismic, and nozzle loads. Included in the evaluations were nozzles, support-to-shell attachments, supports, and support-to-foundation interfaces. These evaluations included bolts and welds. ,
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l Analysis for the simultaneous application of 1.15g horizontal and 0.60g vertical accelerations initially showed the baseplate to be overstressed.
Modifications which are adequate for the Hosgri loads are discussed in Chapter 13.
The natural frequencies of the boric acid filter were determined by dynamic modal analysis with the WECAN computer code. The natural frequency of the filter is 44.0 Hz so it was qualified by the equivalent static load method. The boric acid filter was evaluated for the simultaneous applica-tion of deadweight, pressure, seismic, and nozzle loads. Included in the evaluations were nozzles, support-to-shell attachments, supports, and support-to-foundation interfaces. These evaluations included bolts and wel ds.
Analysis for the simultaneous application of 1.0g horizontal and 0.679 vertical accelerations (which are higher than those for the Hosgri) showed stresses to be within allowables. The most highly stressed area was the anchor bolts which were stressed to less than 31% of allowable. The (November 1978) 7-12 Amendment 72
O boric acid filter has therefore been shown adequate for the Hosgri seismic event.
7.4.4 HEAT REMOVAL The gsidual heat removal pump was analyzed dynamically using an integrated 72 piping system /punp model for determination of seismic loads, deflections, etc., at critical areas of the nump. This analysis was performed using the WECAN computer code and included simultaneous application of conservative horizontal and vertical response spectra.
Stresses in the pump and motor holddown bolts, motor support stand, shaf t, and nozzles were evaluated and shown to be below allowable except for the motor holddown bolts. Rotor to stator clearance, pump and motor bearing loads, and impeller clearance were shown to be within vendor specified level s.
O Summarily, the residual heat removal pump has been shown to satisfy structural (Section 7.1) and operability qualification (Section 7.2) requirements except for the state of overstress in the motor holddown bolts.
Modifications which are adequate for Hosgri loads are discussed in Chapter 13.
The natural frequencies of the residual heat removal heat exchanger were determined by dynamic modal analysis with the WECAN computer code. The first natural frequency of the heat exchanger was determined to be 30 Hz; all other frequencies were above 33 Hz. The residual heat excha'nger was evaluated for the simultaneous application of deadweight, pressure, seismic, and nozzle loads. Included in the evaluations were (November 1978) 7-13 Amendment 72
O For the Hosgri evaluation, the above analysis was repeated for higher acceleration values. Forces under both operation and shut off conditions were taken into account. Seismic restraints were checked for strength.
All values were acceptable and the pump can be considered qualified for the Hosgri event.
7.4.5 'INSTRUf1ENTATION The instrumentation listed in Paragraph 5 of Section 7.3.1 has been qualified seismically. These components are not discussed individually on this report but are described in Chapter 10 Local panels, pressure and differential pressure transmitters, temperature elements, cable trays, instrumentation containment penetrations, process racks, the main control board, and the hot shutdown panel are each given specific discussion in Section 10.3 and are listed in Table 10-1. Instrumentation wiring meets the requirements of Class 1E described in Chapter 8 of the FSAR, with the single exception that each boric acid tank has one, not mutually redundant, level transmitter.
O' (November 1978) 7 14a Amendment 72
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O The reservoir is divided into two sections which communicate but which can l be isolated. Both sections have piping connections to the plant. The pip-ing is asbestos cement and was not intended to be earthquake resistant. l These lines have valving at both ends and would be isolated if damaged in l an earthquake. They would be replaced by fire hose. A hose station will l be established near the reservoir. Redundant valved connections have been 72 l installed in the piping at the plant to receive the hose line(s) from the j reservoir.
l An additional supply of auxiliary feedwater may be obtained from the fire water tank. A normally closed pipe cross-connection has been installed so j 72 .
that water from this tank can flow by gravity to the suction of the auxili-
, ary feedwater pumps. The seismic evaluation of this tank is discussed in !
Chapter 11. i l
As an ultimate backup source to provide water for an indefinitely extended period of hot shutdown, provision will be made to allow seawater to be used as auxiliary feedwater. !
l The valves identified in Table 7-3B were evaluated for the Hosgri seismic Summary
\ event using the methods discussed in Sections 7.1 and 7.2.
results of the valve qualifications including methods employed for the respective valves are presented in Table 7-7A. Refer also to Section 7.4.6.
The seismic qualification of additional Class I equipment not needed fol-lowing an earthquake, but which has been analyzed as part of this evalua-tion, is summarized in Table 7-6. Additional Class I valves not needed i but analyzed are listed in Table 7-3C. Summary results of the valve quali-fications, including methods employed for the respective valves, are sum-marized in Table 7-8. The same basic methods and criteria used to qualify equipment and valves needed following an earthquake were used to evaluate this equipment.
O (November 1978) 7-17 Amendment 72
p)
L 7.6 TESTING Physical testing has been carried out as part of the Hosgri evaluation.
Some tests have been conducted on equipment as installed in the plant and some have been conducted at testing laboratories. The in-plant tests have confirmed frequencies and damping values used in the analysis but have not been to full acceleration levels. The laboratory tests have been full-scale qualification tests. The objectives of the tests were:
- 1. To check the mathematical models used for dynamic analysis.
- 2. To check the structural integrity and operability of components under conditions simulating the postulated Hosgri event. j
- 3. To obtain data and experience for further qualification of other l Diablo Canyon equipment.
The mathematical model validation was accomplished by experimental determination of the dynamic properties of the equipment tested, i.e.:
O a. the natural frequencies of vibration,
- b. the mode shapes (responses), and 72
- c. the critical damping associated with each mode.
The structural integrity and operability of components was demonstrated by shaker-table testing in the laboratory using test response spectra enveloping the required response spectra and by in-plant demonstration while components were statically deflected to the maximum calculated Hosgri deflections.
Table 7-9 shows a comparison of natural frequencies from analysis and from test for some items. Similar good agreement was found for mode shapes and critical damping. Table 7-11 lists the items of mechanical equipment which were tested.
7.6.1 IN-PLANT TESTING General Test Procedures The dynamic properties of the tested equipment were obtained using forced vibration techniques.
(November 1978) 7-18 Amendment 72
The methods of excitation used varied depending upon the particular item under test. As a rule one or more of the following methods were used:
- 1. Impact
- 2. Snapback j
- 3. Eccentric mass vibrator with variable speed control The natural frequencies (resonances) of the test specimen were determined.
The range between 2 and 35 Hz was searched, and any natural frequencies were detected as resonance peaks in the response.
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.The model damping was determined either by the band-width method (Reference IEEE Standard 344-1975) or by determining the time decay of motion (log decrement method).
After the natural frequencies had been identified, the actual motion of the test specimen was determined for each frequency. The phase and 72 relative amplitude data were used as inputs to draw a mode shape for each resonance.
Operability Tests of Values In-Place An insitu testing program employing static load methods was conducted by Westinghouse at the Diablo Canyon Site, demonstrating operability of required active valves while subjected to normal operational and equivalent seismic loads. The valves included in this program are identified in Table 7-10 and include sizes three to fourteen inches as well as motor and air operated valves. This selection is sufficient to cover all valves identified in Tables 7-7 and 7-7A. While subjected to equivalent seismic leads, the valves were operated and paremeters such as stroke time, operator deflection, and motor current and voltage were measured and evaluated. Effects of system pressurization on valve operability were also evaluated. All valves operated satisfactorily and well within design guidelines, clearly demonstrating operability of the valves identified in Tables 7-7 and 7-7A. ( A comprehensive report of this program,'"0PERABILITY DEMONSTRATION OF SELECTED DIABLO CANYON VALVES BY THE STATIC LOAD METHOD" has been submitted to and discussed with the NRC staff.)
.(November 1978) 7-19 Amendment 72
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I 7.6.2 LABORATORY TESTING l
Two main steam safety valves, one 14-inch Limitorque-operated valve and J two pneumatically-operated valves were qualified seismically by full scale j testing on a laboratory shaker table. Test procedures and results have l been reported to the NRC.
The main steam safety valves installed in the plant have both cast and j forged bodies. Two valves were tested, one of each type. Each weighs about )
1200 lbs. They were attached to the test machine in a manner that simulated :
l normal in-service attachment. They were pressurized with nitrogen to 1060 l psig which is about 95% of the maximum valve set pressure. Each valve was 1
instrumented with six accelerometers and four strain gauges located on points ;
where high stresses were expected. A low level sine sweep was conducted for resonances in each of the three orthogonal axes of the valve. At each ;
resonance frequency a resonance dwell was performed at 4.5g for 30 seconds.
The valves were subjected also to biaxial random motion. The random motion 72 for each horizontal axis was excited simultaneously with the vertical axis.
Independent signal sources were used for the horizontal and vertical axes so that input phasing was random. Five OBE and one SSE tests were performed in each orientation. The duration of each test was 30 seconds.
The test response spectra enveloped the required response spectra for the postulated Hosgri event corresponding to the location of the valves, with a wide margin. At the completion of the tests, the set pressure was the same as prior to the testing. The maximum stresses recorded were below yield. Visual examination of each valve upon completion of each test showed no strucutural damage. The tests proved that the main steam safety valves are qualified to stand the postulated Hosgri event loading conditions.
The objectives of the 14-inch Limitorque-operated valve test were:
- a. To test the structural integrity and prove the operability and performance of a 14 inch double disc motor-operated gate valve and its accessories (November 1978) 7-20 Amendment 72
- b. To qualify seismically this valve and similar valve-actuator
/~T . assemblies at Diablo Canyon for seismic inputs associated with V
the postulated 7.5M Hosgri event.
The tested valve is a 14 inch weld-end, outside screw and yoke, double
, disc stainless steel valve of the 300 lbs series. Sections of pipe with )
flanges were welded to both ends of the valve. Two stem-mounted switches l were tested seismically with the valve / actuator assembly. The valve was
! filled with water and pressurized to 350 psig. The total weight of the valve / actuator assembly including the brackets and water was 6300 lbs.
The test was instrumented with 8 triaxial accelerometers and 27 strain gauges. The test imputs represented more conservative responses than the l
postulated Ho;gri event inputs. l l
l Before any dynamic testing was done, the valve / actuator assembly was tested I manually and electrically. Time to open and to close the valve, volts, amperes, watts of the actuator motor, limit switch function, etc., were N observed and/or recorded. Resonance searches were conducted on each orthogonal axis with the valve in the fully closed position and in the {
f ully open position. The response (mode shape) and the damping were 1 determined for each datected resonance. The valve / actuator assembly was (
subjected to-a series of single frequency sine beat tests at 2, 4, 6, 8, 10, 12, 16, 20, 24 and 32 Hz per axis and at the resonance frequencies detected in each axis of excitation. Each excitation lasted for at least 20 seconds. l During each sine beat test, the valve operator was energized and the valve was stroked. The valve / actuator assembly was subjected also to a series of random biaxial multi-frequency test notions for each horizontal axis combined with the vertical axis. Independent (non-phase coherent) multi-frequency motions were used for the horizontal and vertical axes.
All test response spectra enveloped the required response spectra. There were five OBE tests before each postulated 7.5M Hosgri seismic event test.
Each test lasted for 20 seconds with simultaneous opening / closing of the valve. The valve / actuator assembly met the stringent test requirements and proved structural integrity and operability. At the end of all tests, the valve was found tight. The valve was inspected visually and no evidence of damage was found.
(November 1978) 7-21 Amendment 72
During the course of the shaker table testing, it was noticed that under
() conditions of prolonged vibration parts of the actuator which were not properly tightened and fastened might loosen. All safety-related valves with similar actuators at the Diablo Canyon site have been inspected and 72 modified (if necessary) to insure that they are torqued to sp'.cification and that they have the proper type of fasteners and locking devices.
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n (November 1978) 7-22 Amendment 72
Sheet 3 of 6 FABLE 7-3 VALVES B. VALVES REQUIRED IN CASE OF SINGLE FAILURE i
I. Auxiliary Feedwater System LCV-110 2" 900# E-H Op. Stm. Gen. 1-1 Aux. Fw. Supply i LCV-lli 2" 900# E-H Op. Stm. Gen. 1-2 Aux. Fw. Supply LCV-ll3 2" 900# E-H Op. Stm. Gen. 1-4 Aux. Fw. Supply LCV-115 2" 900# E-H Op. Stm. Gen.1-5 Aux. Fw. Supoly .
FCV-37 4" 600# Mtr. Op. Loop 1 Steam to AFW Turbine FCV-38 4" 600# Mtr. Op. Loop 2 Steam to AFW Turbine FCV-436 8" 150# B' fly Mtr. Op. Aux. Fw. Pump 1-1 Raw Water Supply FCV-437 8" 150# B' fly Mtr. Op. Aux. Fw. Pump 1-2 Raw Water Supply Unnumbered Valves at Raw Water Reservcir l II. Chemical and Volume Control System I p FCV-110A 2-RA4200 Air Op. Boric Acid Blender Inlet Q HCV-142 8107 3-RA58RGP 3-GM58FN Air Op.
Mtr. Op.
Chg. Pumps Discharge to Regen. Hx Chg. Pumps Discharge to Regen. Hx Mtr. Op. Chg. Pumps Discharge to Regen. Hx 8108 3-GM58FN 8146 3-IA58DGP Air Op. Chg. to Loop 4 Cold Leg 8147 3-IA580GP Air Op. Chg. to Loop 3 Cold Leg (Alt.)
8387B 3-T58 Manual FCV-128 Manual Bypass Chg. Pump 1 8387C 3-T58 Manual FCV-128 Manual Bypass Chg. Pump 2 8403 3-T58 Manual HCV-142 Manual Bypass 8471 2-X420 Manual Boric Acid Blender Bypass Unnumbered 2"-T58 8145 Manual Bypass III. Residual Heat Removal System 8809 A 8-GM58FM RHR Heat Exchanger to Cold Leg 1 & 2 8809 B 8-GM58FM RHR Heat Exchanger to Cold Leg 3 & 4 IV. Component Cooling Water System FCV-430 30" 150# B' fly Mtr. Op. CCW out of CCWHx No. 1 FCV-431 30" 150#- B' fly Mtr. Op. CCW out of CCWHx No. 2 O
v (November 1978)- Amendment 72
Sheet 6 of 6 l l
l III. Residual Heat Removal System l
} HCV-113 2-RA56RE Air Op. RHR Hx. Letdown Recirculate HCV-670 8-BAS 4D Air Op. RHR Hx. Bypass 8702 14-GM48SEZ Mtr. Op. Hot Leg RHR Suction Valve 8701 14-GM48SEZ Mtr. Op. Hot Leg RHR Suction Valve HCV-638 8-BAS 4D Air Op. RHR to Cold Leg 1 and 2 !
HCV-637 8-BA54D Air Op. RHR to Cold Leg 3 and 4 l l
IV. Component Cooling Water System FCV-430 30" 150# B' fly Mtr. Op. CCW Supply Hdr. A FCV-431 30" 150# B' fly Mtr. Op. CCW Supply Hdr. B FCV-355 2( " 150# B' fly Mtr. Op. CCW Supply Hdr. C FCV-364 12" 150# B ' fly Air Op. RHR l-2 Hx. CCW Outlet TCV-130 8" 150# B' fly Air Op. Letdown Hx. CCW Outlet FCV-365 12" 150# B' fly Air Op. RHR l-1 Hx. CCW Outlet FCV-361 4" 150# B' fly Air Op. Excess Letdown Hx. CCW Outlet V. Steam System Dump Valves PCV-19 8" 600# Air 0p Stm. Gen.1-1 10% Atm. Stm. Dump PCV-20 8" 600# Air Op. Stm Gen.1-210% Atm. Stm. Dump PCV-21 8" 600# Air Op. Stm. Gen.1-310% Atin. Stm. Dump
/7 PCV-22 8" 600# Air Op. Stm. Gen.1-4 10% Atm. Stm. Dump 72 O
O V
(November 1978) Amendment 72
d 1
- ABJ 7-6 (Continuel)
. Sheet 2 of 2
) ftRHARY-MITic Q'7 t? FU'ATVfl & AN i TT"r@ L CW3 1 IQ '!MNT j N 'l % T' k 2 F% C C H .,"/;P I E i *rr j fWENT P E TI 'A TUW H"CGRI EVAUJATI'.N d Accelerat!uns Ac cele rat t uns
! treation Freq. QI. g MF . g QL/HE fie nt i fienti on ETe v ./i lb . Fr, h v Me thM
- Ba s il t s H V H V 1
- 22) FS Flas Thist le 4 .Jualification of the reartcir vessel ass 2res adequacy af this com;oner't.) OK '
Tabtog
, l f3) Fall length Rod (auallrication cf tha react:;r vessel assares anettacy of this expanent.) OK Control Cluster A s semb ly
? Pl6) Fart langth Rml (Qaallrication of the reactor vessel assres adeqancy
- f this compos.ent.) OK Control Cluster Assemtty 1
, PS) Net Transfer Tube 100 ' /cor.t . 5.67T ? .(- !..' ( r /. ) Maximaa. stress: M of all warle OK 2.f 1.2 1.0 1.0 33A
]
1 4
P6) RTP Bypass (Gualification of the applienble piping system asaares alepacy of this integral OK i Manifold c cucpunent . )
I j 27) Evaporator Feed Ion ll5 ' /A ux . , 3 31i 1.c7 0.6 ( 7.6 ) And.or tolts mast Lig'.ly r or 1.c ? c.6 1.0 1.0 Exchange peatneraltzer 33Y streone11 area: Es4 of al.lowatle
( 11)
! PA) RCP Fwal Dypass Orifice (Qua.11rication of the applicatie piping system assuree adequacy of tnis integral OK ccznponent . )
- 29) FP Seal Starktripe (kaalification e' the applicat:Le piping system assures alegancy -f tnis interral OK i Orifice c wponent.)
i I f) Boric Acid Blervier (Qualification of the applicable pip Ltw system assu,res adeq racy of this integral OK caponent.)
l J
! 31) Safety Injection Wmp (Gualification of the applicable piping system assares 41 pacy of this integral OK
! Bypass Orifice c oa.ponent . )
, 32) Bample Heat Exchanger n (Qua11ricativn of the applicatie piping system assures adequacy of this Integral UK
\ mmponent. )
{a
- 31) Spray Idactor (Qualification of the applicatie piping system assures adequacy of this interral cwponent . )
OK
(
7, j 34) Contairunent Spray (Qualification of the applicarle piping system assures stepacy of tnis integral DK Nozzle e spanent ,)
- 33) Containment Hydrogen ll5 ' /Aax . M + 8.0 8.0 (1) Tray sapports most critical part OK 1.22 C.66 *6.6 12.1 Purge Filters J
72
! 36) Containment Hydroge.. 115'/ Aux. Q h5 h$ (S) Fwetlanal after drop test OK 1.2* 0.75 36.9 60 j Nrge Blowers I
j 37) Camponent Cooling 167'/A4x. 21 1.W (1) Force:1 vit: ration tent plannel 3.90 1.60 Water Surge Tank l 30) Make-up Water 200 ' /A ux . 31 1.40 1.40 (1) OK 1. 37 1.00 1. ti2 1.40 4 Transfer Pump s
) M) Fire har L IS ' /A ux . ?0 1.t'O 1.10 f^
- 10) Co,., Ptorage Tank I W /t.P. 15H Foreji Hodt Icatvigration on requ{entred conducted j
1
- See Notes which foll:w Table 7-8 e
i i
t t (November 1978) Amendment 72 i
I a
h i
e t
T l
I I
l.
l i
O O O Sheet 1 of 2 TABLE 7-7 St%tARY - SEISMIC QUALIFICATION MINDEJM REQUIRED ACTIVE VALVES FOR HOT SFUTDOWN AND/OR CCLD SHUTDOWN VALVE QUALIFICATION HOSGRI EVENT Accelerations Accelerations Freq., QL, E HE, g QL/HE Iocation Function H2 H
- Method Results H V H V LCV-lO6 Aux. Feedwater from Turbine AFP to S.G. 1 bl.9 '
- 16. 3* - (1) ge most highly stressed 2.25 1.72 .5.76*
ILV-107 Aux. Feedwater from Turbine AFP to S.G. 2 Ll.9 16.3* (1) Yoke most highly stressed u 2.25 1.72 5.76*
Aux. Feedwater from Turbine AFP to S.G. 3 II)
LCV-108 kl.9 16.3* g most highly stressed 2 3.12 1.96 4.k2*
LCV-109 Aux. Feedwater from Turbine AFP to S.F. 4 kl.9 16.3* III Yoke most highly stressed 4.55 1.96 3 29*
FCV-95 Steam to Aux. F.W. Pump Turbine 38.6 7.84* (1) Yo**emosthighlystressed ., 2.38 1.73 r>6?=
6 4 (1) Yoke most highly stressed 0.95 0.47 6.3 8.5 8104 Boric Acid to Chg. Pump >33 area: 93% of allowable Yoke most highly stressed 8145 Pressurizer Auxiliary Spray >33 6 4 (1) area: 50% of allevable 1.42 0.60 4.2 6.7 880$A RWST to Charging Pump Suction 24 3 2 (7) Yoke most highly stressed 2.31 c.D(18) 1.30 2.70 0.85(18) 1.03 2.35 D 8805B RWST to Charging Pump Suction RhR Suetion 24 26.3 3
3.h5 1.91 2 (7)
(1)
{o'[mo Yoke most highly stressed ihl h 2.90 2.47 1.88 1.4 1.c5 8701 area: 6h% of allowable RHR Suction 26.3 3.45 1.91 (1) Yoke I:nst highly stressed 3 52 2.27 0. o 8'02 area: 64% 'of allowable (12) 12) .84 E V-637 RHR to Cold leg >33 2 (1) Yoke most highly stressed r* 1.76 1.21 1.70 1.65 3 area: 90% o.f allowable RHR to Cold Leg >33 3 2 (1) Yoke most highly stressed 1.66 L23 1.$1 1.55 HCV-638 area: 901 of allowable' .
FCV-355 Component Cooling Water Header C 72.9 - 15.2* (1) ggmosthighlystressed 6.02
. 3de 2.17*
most highly stressed
~
FCV-364 RHR Heat Exchanger No. 2 CCW Return 93 2 15.h* (1) Y 2.63 2.15 4.56*
FCV-365 RHR Heat Exchanger No. 1 CCW Return 93.2 15.h* (1) Yggmosthighlystressed y,$9 g'.07 8.06+ , ,
6 4 (1) Ycke most highly stressed 3.54 1.39 1.69 PCV-19 SG 105 Atmos. Relief >33 ..,
area: 103% of allowable (14)
SG 10% Atmos. Relier >33 6 4 (1) Yoke most highly stressed C 5.63 1.39 -1.06 2.88 PCV-20 area: 103% of allowable (14)
SG 10% Atmos. Relief >33 6 4 (1) Yoke most highly stressed 4.83 1 20 1.24' 3.-3r PCV-21 area: 103% of allowable (14)
PCV-22 SF 105 Atmos. Relief >33 6 4 (1). Yoke most highly stressed ." 5.40 1.22 1.11 }.28' area: 103% of allowable (14) f
- Resultan+ in any direction f Nowmter 1978) Amendment 72
\ [
V V V Table 7-7 Sheet 2 of 2 VALyg 7ALIFI m :C. .~! EVALUATION Accelerations Accelerations Freq., 3,
- W. r C11ME location Function MZ H 4 Ethoa Entdu H . H V RY-3 min Steam Safety, Icad 1 > AoV 11 9 (3) Maximum stress 301 of allowable 2.59 1 39 4.25 6.47 RV h Main Steam Safety, Lead 1 >k!y 11 9 (3) Maximum stress 3 M of allo ntle 3.46 1.39 3.18 6.47 1
RV-5 Main Steam Safety, Icad 1 >d n 9 (3) Maximum stress 30T, of anowstle 7.10 2.61 1 55 5.81 RV-6 Main Steam Safety, Icad 1 , u 9 (3) Maximum stress 305 of allowable 7.30 2 38 1.51 5 96 RV-7 Main Steam Safety, Imad 2 ,[ n 9 (3) mximum stress 30% of anowable 2 98 1.39 3.69 6.47 RV-8 Main Steam Safety, Lead 2 > 11 9 (3) Maximum stress 30$ of allevable 2.M 1.39 3.69 6.47 Rv-9 min Steam Safety, lead 2 >4cy n 9 (3) Maximum stress 30L of allavable 6.ru 1 39 1.63 6.k7 RV-lO Main Steam Safety, Icad 2 >{$$ 11 9 (3) Maxirus stress 30% of allovutic 6.67 1 39 1.65 6.47 RV-n min Steam Safety, lead 3 3 11 9 (3) Maximum stress 30% of anovetle 4.82 1.69 2.28 3 95 RV-12 Main Steam Safety, lead 3 >[ 11 9 (3) Maximum etress 30% of allevable 5.15 1.69 2.14 4.21 RY-13. Main Steam Safety, Imad 3 > n 9 (3) Maximum stress 30% of anovable 5.68 1.69 1.94 4.64 RV-lk Main Steam Safety, Lead 3 > 11 9 (3) Maximum stress 30% of allowable 6.25 1.69 1.76 5.11 RV-58 Main Steam Safety, Lead 4 >d 11 9 (3) mximum stress 30% of allovable 5.27 1.75 2.09 4.31 RV-59 Main Ste- a Safety, lead 4 >k 11 9 (3) Maximum stress 3% of allovable 5.27 1.75 2.09 4.31 lbH RV-60 Nin Steam Safety, lead 4 >4 11 9 (3) mxi=um stress 30% of allovable 5.27 1.75 2.09 4.31 RV-61 Main Steam Safety, Lead k > c. 11 9 (3) Maximum stress 301 of allowable 5.27 1.75 2.09 4.31 l8 2.11 6.47 RV-222 Main Steam Safety, Iead 1 >g 11 9 (3) Maximum stress 30% of allovable 5.22 1 39
- RV-223 min Steam Safety, lead 2 >d y 11 9 (3) Maximum stress 30% of allevable 4.25 1 39 2.59 6.h7 RV 224 min Steam Safety, Lew 3 > 11 9 (3) mximum stress 30% of anovable 7.17 1.69 1.53 5.88 18H
' RV-225 Main Steam Safety, lead 4 >40V 11 9 (3) Maximum stress 30% of allowable 5 27 1 75 2.09 k.31 FCV-41 Win Stesa Isolation, Ioop 1 > 33 h.50 1.75 (1) _
FCV-42 usin Steam Isolation, loop 2 > 33 4.50 1.75 (1) 72 M -43 Main Steam Isolation, loop 3 > 33 h.50 1.75 (1)
FCV-44 Main Steam Isolation, Loop 4
- 33 h.50 1.75 (1) 8010A Pressurizer Safety >33 6 h (1) Max. stress in valve 4.9 0.7 1.2 5.7 bonnet: 33% of allowable 8010B Pressurizer Safety >33 6 h (1) Max. stress in valve 4.9 0.7 1.2 5.7 bonnet: 33% of allowable 8010C Pressurizer Safety >33 6 4 (1) Max. stress in valve 4.9 o.7 1.2 5.7 bonnet: 33% of alMwable
_gmber1978) Amendment 72
( TABLE 7.94 SDJMARY - SEISMIC QUALIFICATION VALVES REQUIRED IN CASE OF SINGLE FAILURE VALVE QUALIFICATION HOSGRI EVALUATION Accelerations Accelerations Freq., QL. g HE, g QL/HE O cation _
M etion H2 H V Mathod Results H V ' H V 1
ICV-110 Au. Feedwater from lbtor AFP to S.G.1 ED.3 9 7* (1) {pkemosthighlystressed 2.43 L72 3 26*2 IIV-lli Aux. Feedwater frcan Motor AFP to S.G. 2 28.3 97* (1) Dngmosthighlystressed 2 95 1 ,74 2.83*
LCV-113 Aux. Feedwater from Motor AFP to S.G. 4 2813 9 7* (1) {pggmosthighlystressed 2.03 1 97 3.43*
14V-115 Aux. Feedwater from Motor AFP to S.G. 3 28.3 9 7* (1) D)gmosthighlystressed 2.03 L.97 3.43' FCV-37 Steam to Aux. F.W. Pump Turbine, loop 1 3P 6 7.84* (1) Modification being designed 6.1 2.7 1,18*
72l FCV-38 Steam to Aux. F.W. Pump Turbine, Loop 2 38 6 77.84* (1) Modification being designed 4.53 1.91 J.59*
FCV-436 Raw Water To Aux. Feed Pump 64.6 12.6* (1) {pggmosthighlystressed 2.09 1.40 5.00*
FCV-437 Raw Water to Aux. Feed Pump 64.6 12.6* (1) {pggmosthighlystressed 2.09 1 40 5.00*
Manual Valves Raw Water Res,ervoir FCV-110A Borie Acid Blender Inlet > 33 6 4 (1) gpbogt gl,t sgogt , stressed - 1.54 1.41 3.90 2.84 7 1 HCV-142 Chg. Pump Discharge to Regen. Heat Exch. 20 4.2 3 (1) gpyt gt piggsggese+
9 1.32 0.79(18)3.1B .3.79 8107 Chg. Pump Discharge to Regen. Heat Exch. > 33 3 2 (1) Ygg,mo big 1.83 1.21 1.64 1.65
' 8108 Chg. Pump Discharge to Regen. Heat Exch. > 33 3 2 (1) E)*:" $3g h h g stressed h*d 1.83 1.21 1.64 1.65
- 8146 Chg. Pump to Ioop 4 Cold leg 20 4.2 3 (1) !@e""$ ko 3 E l***
Bonnet most ighmy a ressed 2.95 1.37(18)1.42 2.19 72 8147 Chg. Pump to Ioop 3 Cold leg 20 4.2 3 (1) area: llevable 3.22 3.54(18)1.30 0,85(17,19'.
Yoke mo b stresse 1*07 8387B FCV-128 Manual Bypass, Chg. Pucp 1 > 33 6 4 (1) area: ovable d 0.57 5.61 7.02 -
8387C FCV-128 Manual Bypass, Chg. Pump 2
- 33 6 4 (1) M*:" hh'ME*d 1.10 0.57 5.45 7.02 8403 HCV-142 Manual Bypass > 33 6 4 (1) Y ke,mo are high g g ed 0 99 0.78 6.06 - 5.13 .
8471 Boric Acid Blender Bypass > 33 6 4 (1) g .bo g t jo g g stressed 1.93 1.41 3.11 2.84 j
-- 8145 Manual Bypass 207 6 4 ,1)
( go g t g t,gi g gggesed 1.42 0.60 4.22 6.67 ,
^
88094 RHR Heat Exchanger to Cold Lege 2k 2*7 1*O (1) :
h yv #
s 8809B . FER Heat Exchanger to Cold Legs 24 27,1.76 1.8,2.45 (1) area: ,
1.75 1.00 1,co 2.45 ICV-430 CL2ponent Coeling Water Header A 97.8 6 2 (1) ge ? o Alls. 3.37 1 93 1.78 1.04
- FCV-431 Component Cooling Water Header B 97.8 6 2 (1) Yoke most big _. d: 2 99 1.47 2.01 1.36 of allowable
! PCV-602 Aax. S.W. to CCW Ht. Exch. No. I 135.1 6 2 (1) , sed: 2.10 2.86 eg g ghg 2.13 FCV-603 Aax. S.W. to CCW Ht. Exch. No. 2 135.1- 6 2 (1) Yoke most highly stressed: 2.10 2.13 2.86 l 72 9351A Reactor Coolant Systest Sample . 33 8.5 4 (1) fallowabf str m - 1.37 0.58 6.20 6.h i i
9351B Reactor Cor.lant system sample > 33 8.5 4 (1) hbhohaNowable area:
1.37 0.58 6.20 6.90 g bo g tj o,l g g g stressed ;
9356A Reactor Coolant System Sample , 33 8.5 4 (1) Body-bonnet bolts most stressed 1.31 1.43 6.48 2.84 area: 67% of allowable
?
E (November 1978)
- Resultant in any direction Amendment '72 i
1
\ N
[
Sheet 1 ob j}
1 7-8 L lC QJALIFICATION POR NORMAL
.% . M. f 4 COID SiUTDnWN i f H03GRI EVALUATIC'I _
VALVE QUALi'ICATION Accelerations Accelerations Ireq., @,g HE. g QL/HE M V H V W . tion Function NZ H V Method Results T
YALTES LIS*TD IN TA3LE 7-3C LCV-106 Aux. Feed *ater From Turtine AFP to S.fl.1 41.9 16.3+ s1) gg;unethighlystressed 2.25 1.72 5.76+ ,
LCV-107 Aux. Feedwater from Turbine AFP to S.G. 2 bl.9 16.3* (1) {pggmosthihlystressed 6 2.25 1.72 5.76*
IIV-108 Aux. Feedwater fmni Turbine AFP to S.G. 3 kl.9 16.3* (1) ggg most highly stressed 3.12 1.96 4.42+
LCV-109 Aux. Feedwater from Turbine AFP to S.G. 4 L1.9 16.3+ (1) ggg_mosthighlystressed 4.55 1.96 3.29*
ITV-Il0 Aux. Feedwater from Motor AFP to S.G.1 28.3 97* (1) {pggmosthighlystressed 2.b3 1.72 3.e6-LCV-lll Aux. Feedvater from Wtor AFP to S.G. 2 28.3 9 7+ (1) gggmosthighlystressed 2 95 1.74 2.83*
LCV-ll) Aux. Feedwater from Wtor AFP to S.G. 4 28.3 9 7* (1) g most highly stressed Y 2.03 1.97 3.43*
IfV-ll5 Aux. Feedwater from Motor "JP to S.t 3 28.3 9 7* (1) {pggmosthighlystressed 2.01 1.97 3.43' FCV-4 36 Raw Water to Aux. Feed Pump 6L.6 12.6* (1) {pggmosthighlystressed 2.09 1.LO 5.00*
FCV-4 37 Raw Water to Aux. Feed Pump 64.6 '2.6* (1) gggmosthighlystressed 2.09 1.40 5.00*
FCV-95 Steam to Aux. Feedwater Pu p Turbine 31.6 7.84+ (1) gggmoethighlystressed 2.38 1.73 1,43. ,
FCV-llCA Boric Acid Blender Inlet >33 6 L (1) B ar g bo g t jolt g g stressed 1.93 1.41 3.11 2.86 FCV-1103 Boric Acid Blender Outlet > 33 6 4 (1) ggg.no g h gh g t g ed 1.93 1,41 3.11 2.84 72 3CV-142 l.32 Cbg. Pump Discharge to Regen. Heat Exch. 20 k.2 3 (1) nan h obabcvNS" c.7p 3.18 3.79 8145 Pressurizer Auxiliary spray > 33 6 4 (1) If!I:" N Y E l' cM ii'd 1.h2 0.60 k.23 6.67 8146 Charging Pu=p to loop 4 Cold Leg 20 L.2 3 (1) a a e o N NU 2 95 1.37(18)1.L2 2.19 8147 Charging Pump to Loop 3 Cold Leg 20 4.2 3 (1) aN N k oE"a Yl ow N 3.22 3.54 1.30 0.85(17,L9)
Letdown Line Isolation 39.8 6 4 (1) Yoke most highly stressed: '6
.. 0.8 3.8 5.0 LCV-459 86% of allowable LCV-460 Istdown Line Isolation 58 6 4 (1) Flange most highly stressed: 1.41 0.60 3.80 5.95 112% of allowable. Adequate for 3g H, 2g V 8149A Letdown Line Orifice 39.8 6 4 (1) Yoke most highly stressed area: 2.59 0.67 2.32 5.97 865 of allowable 8149B Letdown Line Orifice 39.8 6 4 (1) Yoke most highly stressed area: 1.82 0.67 3.30 5.97 865 of allowable 8149C Letdown Line Orifice 39.8 6 4 (1) Yoke most highly stressed area: 2.51 1.16 2.39 3.45 86% of allowable 33 6 4 (1) Yoke most highly stressed area PCV-135 Letdown Heat Exch. To Demineralizer 2k% of allowable 1.57 o.99 3.82 4.oh 8104 Beric Acid to Charging Pucp >33 6 4 (1) Yoke most highly stressed 0.95 0.47 6.3 85 area- 93% of allowable 72 TcV-149 Letdown Hz totdown Bypass 33 6 4 (1) Yoke 'most highly stressed 1.57 0 99 3.82 4.04 8168 Mixed Bed Demineralizer Discharge 33 6 L (1) M *:moIt4 hSgh$ N Ned 1 93 1.kl 3.11 2.04 area: 4(4 of allowable (Novemb r 1978) M(esultarit in any direction Amendment 72
= _ .___....-_.-____.m._ _ _ _ __ . _ _ _ _ _ __ , _ _ _ , _ _ . ___ . _- _ _ m _ _ _ . . _ _ - . _ . - _
r TABIE 7-8 Sheet 2 of 7
)
r VALVE QTJA TION HOSGRI EVAIARTIJ3 l Accelerations. Accelerations Freq., QL, g HE g QL/HE l' Location Function H2 H V F Results H V H _
-V s ed ICV-112A Volume Control Tank Inlet from Demineralizer >33 '6 h ahhighla 1c 1 93 1.41 3.11 2.8h FUV- nlB Boric Acid alender to VCT >33 6 4 e most highly stressed 1,5h 1.41 3.90 2.8' ICVA55A Pressurizer Spray fr m Loop 2 >33 6 4 ,y,3 9 31 r all wable d 2.26 1.18 2.7 3k
- PCV 455B Pressurizer Spray fr m Loop 1 '33 6 4 (, r o@mo . y E d 1.49 1 92 4.0 2.1 l'
FCV- n1A ~ Primary Water to Boric Acid Blender *33 6 h (1) g"go gilmo 1.93 1.41 . 3.n 2.84 HCV-133 RHR Ex Letdown Recirculate '33 6 4 (1) g(moghghgsgged Y 1.57 0 99 3.82 1.05 HCV-670 RHR Hx Bypass >33 3 2 (1) Yoke most highly stressed 1.67 1.32 1.8 1.5 area: 90% of allowable (13) 8701 RER Suction 26.3 3.45 1 91 (1) Yoke most highly stressed 2.47 1.83 1.k0 1.05 area: 64% of allowable s 0 98 0.84 8702 RHR Suction , 26.3 3.45 1 91 (1) Yoke most highly stressed 3 52 2.27 !
area: 64% of allowable (12) (12)
HCV-637 RER to Cold Iag 3 & k >33 3 2 (1} Yoke most highly stressed 1.76 1.21 1.70 1.65 area: 90% of allowable (13) !
HCV 638 RHR to Cold Ing 1 & 2 >33 3 2 (1) Yoke most highly stressed 1.66 1.29 1.81 1.55 ,
area: 90% of anowable (13) i FCV 430 Caponent Cooling water Header A 97.8 6 2 (1) 3 37 1.93 1.78 1.04 7tv h31 Caponent Cooling Water Header B 97.8 6 2 (1) 2.99 1.47 2.00 1.36 FUV-355 Component Cooling Water Header C 72 9 15.2* (1) Yoke most highly stressed 6.02 3.56 2.17* I area j i
FUV-364 RHR Heat Exchanger No. 2 CCW Return 93.2 15.4* (1) Yoke most highly stressed 2.63 2.13 4.56*
area
' ICV-130 Intdown Heat Exchanger CCW Return 209 7 6 2 (1) Yoke most highly stressed 3 20 1.52 1.88 1.32 area 1
FEV-365 RHR Heat Exchan6er No.1 CCW Return 93 2 15.te (1) Yoke most highly stressed 1.59 1.07 8.06*
area FrV-361 Excesa Ietdown Heat Exch. CCW Return 839 49.6 * (1) Yoke most highly stressed 1.99 2 36 16.1*
area PCV-19 Stm. Gen. 1-1 10% Atm. Stm. Dump >33 6 h (1) Yoke most highly stressed 3.54 1 39 1.69 2.88 area: 103% of allowable (14)
ICV-20 Stm. Gen.1-210% Atm. Stm. Dump >33 6 h (1) Yoke most highly stressed 5.63 1.39 1.06 2.88 area: 103% cf allowable (14)
PCV-21 Stm. Gen. 1-3 10% Atm. Stm. Dump >33 6 k (1) Yoke most highly stressed 4.83 1.20 1.24 3.33
, area: 103% of allowable (14) ;
f ECV-22 Stm. Gen. 1 h 10% Atm. Stm. Dump >33 6 h (1) Yoke most highly stressed 4.80 1.22 1.25 3 28 i area: 103% of anowable (1k) t 4
(November 1978)
- Resultant in any direction Amendment 72
. _ _ - . - - - . __. _m . . .__ _ . _ _ . _ _ _ -
TABIZ 7-8 V Shee of 7 VALVE QIALIFICATION HOSCRI EVALUATION Accelerations Accelerations Freq., QL, g HE, g QL/HE Location Function HZ H V Method Results H V H V OTHER RD40TELY OPERATED CIASS I VALVES KV-37 Steam to Aux. F.W. Pump Turbine, Icop 1 38.6 7.84* (1) Modification being designed 6.1 2.7 1.18*
FCV-38 Steam to Aux. F.W. Pump Turbine, Loop 2 38.6 7.84* (1) Modification being designed 4.53 1.91 1.59' KV h1 Main Steam Isolation, Inop 1 >33 4.50 1.75 (1) 3.64 1.19 ~1.24 1.47 KV k2 Main Steam Isolation, Loop 2 22.6 2.5* 1.75 (1) 1.69 1.25 2.66 1.40 FCV h3 Main Steam Isolation Ioop 3 22.6 1.75 (1) 1.73 1.20 2.60 1.46 2.5*
M hk Main Steam Isolation, Icop 4 22.6 1.75 (1) 1.69 1.27 2.66 1.38 2.5*
KY-128 Centifugal Charging Pumps Discharge 24 8.5 4 Bonnet most highly stressed 0.90 0.51 9 kk 7 64 area: 7kf,of allowable KV-151 Steam Generator No. 1 to Blowlown Tank 19.3 5 91* (1) 1 94 1.41(18)2.k7* 72 FCV-15h Steam Generator No. 2 to Blowdown Tank 19.3 5.91* , (1) 1.85 1.41(18)2.55*
K V-157 Steam Generator No. 3 to Blowdown Tank 19.3 5.91* (1) 1.92 1.41(18)2.48*
KV-160 Steam Generator Nr. I to Blowdown Tank 19.3 5.91* (1) 1.90 1.41(18)2.50*
KV-307 CCW Out of Evap. Condenser 87.8 24.3* 2.58 1.26 8.47*
FCV-356 Caponent Coolirg Water to DC Pumps 81.7 6.3+ (1) 3 29- 2 34 1.56*
KV-357 BCP 'Ihermal Barrier CCW Return 16 6.51* (1,3) 3.53 1.69(18) 1.67*
NY-360 Containment Fan Cooler CCW Return 115 60* (1) 3.63 3.89 11.3
- KV-363 BCP Oil Cooler CCW Return 82 6.9* (1) 2.66 2.h6 1.90*
FCV-366 Containment Fan Cooler CCW Return 185 43* (1) 4.09 2.09 9.35*
KV 438 Main Feedwater Isolation, S.G. 1 28.4 37.5* (1) 2.08 1.25 15.4*
K V 439 Main Feedwater Isolation, S.G. 2 28.4 37.5* (1) 3.42 1.25 10.3*
FCV hbO Main Feedwater Isolation, S.G. 3 28.4 37.5* (1) 2 99 1.h5 11.2*
FCV-441 Main Feedwater Isolation, S.G. 4 28.4 37.5* (1) 2 99 1.60 11.1*
FCV-606 Cae:ponent Cooling Water Pump No.1 Recire. 19.h 6.36* (1) 2.61 0.92(18) 2.30*
FCV-607 Component Cooling Water Pump No. 2 Recire. 19.4 6.36* (1) 2.61 0 92(18) 2.30*
FCV-608 Component Cooling Water Pump No. 3 Recire. 19.4 6.36* (1) 2.61 0.92(18) 2.30*
FCV-749 RCP Oil Cooler CCW Return 82 6.9* (3) 2.55 4.11 1.43*
FCV-750 RCP Themal Barrier CCW Return 16 6.51 * (3) 4.00 2 93(18) 1.31*
FCV-760 Steam Generator No.1 Blowdown and Sample 18.0 5.52* (1) 3.54 0.81(18) 1.52' FCV-761 Steam Generator No. 2 Blowdown and Sample 18.0 5.52' (1) 2.85 0.61(18) 1.90*
FCV-762 Steam Generator No. 3 Blewdown and Sample 18.0 3.52' (1) 3 50 2.23(18) 1.33*
4 FCV-763 Steam Generator Na, 4 Blowdown and Sample 18.0 5.52* ( 1) 3.13 o.6o(18) 1.73*
FCV-1307 Waste Eirap. Cormienser CCW Return 87.8 24.3* 9.24*
(1) 2.31 1.26 i
(November 1978)
- Resultant in any direction Amendment 72
TABLE 7-8 Sheet 4 of 7 VALVE QUALIFICATICN HOGGRI EVALUATICN Accelerations Accelerations 5 0;., , HE. g QL/EE Locction Function ny '
Witod Pesults H V H V I
LCV-69 Component Coolir.g sater Make up 18.0 US (1) 4.65 0 99(18) 1.1G*
LCV 70 Ccq<anent Cooling Water Make up 18.0 5.52* (1) 5.84 0.99(18) c.93+
t LCV-112B Voluma Control Tank to Charging Pumps >33 6 6 (1) t mQ hig ly stressed 1,93 1,g1 3,u g,g IfV-112C Volume Control Tank to Chargirg P=ps >33 6 6 (1)
B nnet m st highly stressed 1.93 1.41 3.11 4.26
'#** 99$ *
- PCV-11.9 Air to Aux. Bldg. Inflatable Seals 4.70 1.57 4.70 1.57 1.00 1.00 72 PCV-4 MC Pressurizer Power Relief 20 4.2 3 (1,3) Bonnet most highly stressed 3.22 0.65(18)1.30 4.62 area: 80% of allcwable PCV-456 Pressurizer Power Relier 20 4.2 3 (1,3) Bannet nog highly stre wed 1.23 2.17(18)3.41 1.38 area: Se% of allevable
~
PCV-474 Pressurizer Pcwer Relief 20 4.2 3 (1,3) @et n& bishly .ststresser' 1.29 o.66(18)3.26 4 555 area: 80 8 of allowable -
t~
18H RV-3 min Steam safety, Lead 1 3 4cy 11 9 (3) mximum stress 305 of anowable 2.59 1,39 k.25 6.47 ,
RV 4 Main Steam Safety, Lead 1 (3) Maximum stress 30% of allowable 3 46 1 39 3 18 6.47
>@ 11 9, RV-5 Main Steam Safety, Icad'l . 11 9 (3) Maximum stress 30% of anovable 7.10 2.61 1.55 5.81 RV-6 Main Steam Safety, Imad 1
, H 9 (3) Maximum stress 30% of anovable 7.30 2 38 1.51 S.96 RV-7 Main Steam Safety, Ien.1 2 11 9 (3) Maximmt stress 30% of anovable 2.98 1 39 3.69 6.47 RV-8 min Steam Eafety, lead 2 , n 9 (3) Maximum stress 30% of allowable 2 98 1.39 3.69 6.47 Rv-9 min Steam Safety, Lead 2 n 9 (3) Maximum stress 30% of anovable 6.74 1.39 1.63 6.47 RV-10 RV-11 Main Steam thfety, Iead 2 min Steam Safety, Lead 3
@ n n
9 (3) Maximum stress 30% of anovable mximum stress 305 of allowable 6.67 4.82 1.39 1.69 1.65 2.28 6.47 3 95 H 9 (3)
RV-12 Main Steam Safety, Lead 3 11 9 (3) mximum stress 30% of allowable 5.15 1.69 2.14 4.21 RV-13 min Steam aarcty, Lead 3 ,1b 11 9 (3) Maximum stress 305 of anovable 5.68 1.69 1.94 4.64 RV-14 min Steam Safety, Lead 3
, 11 9 (3) mximum stress 30% of snowable 6.25 1.69 1.76 5.n
.RV-58 min Steam Safety, lead 4 , n 9 (3) mrimum stress 30% of allowable 5.27 1.75 2.09 4.31-RV-59 Main Steam Safety. Lead 4 >
18H
{ 11 9 (3) mximum stress 30% of allowable 5 27 1.75 2.09 4.31 RV-60 Main Steam Safety, lead 4 ,gy n 9 (3) Maximum stress 301,'of allowable - 5.27 1.75 2.09 4.31 18H RV-61 Main Steam Safety, Imad 4 >40V 11 9 (3) Maximum stress 30% of allowable 5.27 1.75 2.09 4.31 18H RV-222 Main Steam Safety, Imad 1 ,gy 11 9 (3) Maximum stress 30% of allowable 5.22 1 39 2.n 6.47 18H RV-223 Main Steam Safety, Iead 2 ,gy n 9 (3) Maximum stress 30% of anovable 4.25 1 39 2.59 6.47 18H RV-22k Main Steam Safety, Lead 3 > goy 11 9 (3) Maximum stress 30% of anovable 7.17 1.69 1.53 5.88 18H RV-225 Main Steam Safety, Lead 4 >40V 11 9 (3) Maximum stress 30% of allowable 5.27 1.75 2.09 4,31 (November 1978)
- Resultant in any direction Amendment 72
O O O TABIE 7-8 Sheet 5 cf 7 VALVE QUALIFICATION HCGGRI EVALUATION Accelerations Accelerations Freq., QL, g HE, g QL/EE Incation Function HZ H V Method Results H V H V WC-27 Consponent Cooling Water Return Header B 254.6 6 2 (1) Yoke ecst highly stressed 3.62 2.09 1.66 0 96 area: 11% of allowable (17)
WC-28 Ccx:ponent Cooling Water Return Header A L68.3 6 2 (1) Yoke most highly stressed 4.09 2.09 1.47 o.%
area: 19f, of allowable (17) 2.5 Yoke mo t h stre ed 1.00 1.00 8000A Pressurizer Power Relief Isolation 233 4.9 (1) , g 4.9 2.5 8000B Pressurizer Power Relief Isolation 333 g9 g.5 (1)
Yoke most highly stressed 49 2.1 1.oo 1.19 area: 82% of allowable 8000C Pressurizer Power Relief Isolation >33 4.9 2.5 (1) Yoke most highly stressed 4.9 1.3 1.00 1 92 8100 Reactor Coolant Pump Seal Water Return >33 6 6 (1) REnit M t sitHy=suessed 1.31 o.67 u.58 8 95 area: 99% of allowable 4.04 8105 Centrifugal Charging Pump Recirculation >33 6 4 (1) hi stre sed 1.57 0.99 3.82 YokemoNZokhlailowab!e area:
8106 Centrifugal Charging Pump Recirculation 233 6 4 (1) 1.57 0.99 3.82 4.04 Yoke mo g h ghg stregsed 8107 Chg. Pump Discharge to Regen. Jteat Exch. 233 3 2 (1) k most hi h res ed
- 1. 1.21 1.M W area: 83%ofaflowable 1.81 1.21 1.66 1.65 8108 Chg. Pump Discharge to Regen. Heat Exch.
233 3 2 (1) Yoke most highly stressed 6 6 area: 83% or allowable 1,44 o.93 k.17 6.45 8112 Reactor Coolant Pump Seal Water Return >33 (1) 8124 Letdown Heat Exchanger Outlet Relief >33 8.5 4 (1) g[onnetg111,"jmos hgkysEressedgstgesed ish 1.57 0.99 5.41 4.04 81klA Reactor Coolant Pump No.1 Seal Icakoff 20 8.5 4 (1) Qt 1.28 0.58 6.64 6.89 t hi [IN seed 81klB Reactor Coolant Pump No. 2 Seal Ieakoff 20 8,5 4 (1) B rt t i y k ssed 1.28 o.58 6.64 6.89 81blC Reactor Coolant Pump No. 3 Seal Irakoff 20 8.5 4 (1) t high 1.31 o.58 6.48 6.89
$[>E area: t $ of at[.[owab[eIk esed 6.48 81hlD Reactor Coolant Pump No. k Seal Icakoff 20 8.5 4 (1) Bonnet most highly stressed 1.31 o.58 6.89 8152 Ietdown Line orifice 20 1.8 1.2 (1) $5E*je[o}t[t%")Il essed 1.27 1.59 12 2.03 area: 76% of allowable str ed 0.94 8700A MIR Pump No.1 Suction from NCS >33 6 6 (1) Bnnetmgthgh 1.27 4.72 6.38 8700B RHR Pump No. 2 Suction from BC3 >33 6 6 (1) Bonnet most highly stressed 1.25 0.57 k.8o 10.52 area: 105% of allowable 8703 MIR Heat Exchanger to Hot Legs >33 6 h (1) 3 3h 2.85 *l.80 1.40 Yokemosthighl stressed
>33 2.5 2 (1) 1,67 1.58 1.49 1.26 8716A RHR Heat Exchanger No. I to RCS Hot legs 3h'*ne $styg$$v f
tressed RHR Heat Exchanger No. 2 to RCS Hot Legs 2.5 2 (1) sthih$y 2.06 1.32 1.21 1.51 8716B >33 Q:ne tressed area: 95% of allowable Amendment 72 (November 1978)
_m , _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ __
TABLE 7-8 Sheet 6 of 7 QUALIFICATION HOSGRI EVALUATION YAL*2 Accelerations Accelerations Freq., CL , c HE. g qL/HE H V H V
.1 Location Functice HZ H V Method Results Boron Injection Tank Discharge > 33 6 4 (1) 2.20 1.6' 2.73 2.50 8801A {og mogtg p g s g sed
,-88018 Boron Injection Tank Discharge >33 6 4 (1)
Y
- " $4 yhl&Qg'd are :
th 2 53 1.20 2 37 3 33 8802A ' Safety Injection Pump Discharge to Hot Iegs >33 6 4 (1) g eo g h gh g syged - 4.30 2.54 1.40 1.57 Safety Injection Pump Discharge to Hot Lega (1) 2*23 1*93 2.69 2.07 8802B.' >33 6 4 g eo g hgh g sgged Charging Pumps to Boron Injection Tank >33 6 4 (1) hi hl stressed 0.90 0.51 6.67 7.84 8803A" YokemoNIoka}1ovable area: .
Charging Pumps to Boron Injection Tank 6 4 (1) 0.90- 0.51 6.67 7.84 I
8803D >33 Yoke most hi h stressed area:
880hA ' RHR Heat Exch. to Charging Pump Suction >33 2.5 2 (1) @:"*g7% ok 95f"oEtN#M1 a o blowable stressed 1.97 1.26 1.27 1.59 Body neck most highly stressed RER Heat Dtch. to SIS Pump Suction > 33 2.5 2 (1) area: 95% of allowable 2.01 ~ 1.73 1.24 1.16 880LB ,
- I
' 8805A RWST to Charging Pump Suction 24 3 2 (7) Yoke most hiably stressed 2 31 1 area: -of allowable 0.7} M .30 2.70
{ 8805B R M to Charging Pump alction 24 3 p (7) a* al U90 0.85 . 1.01 2. M ~,
~ 880'IA
, Charging Pump - SIS Pusp Suction Crosatie > 33 6 6 (1) Bonnet most highly stressed area: 99% of allowable 1.48 0.92( W 4.C5 6.52 8807B- Charging Pump - SIS Pump Suction Crosetie > 33 6 6 (1) Bonnet most highly stressed area: 1.48 0.92 4.05 6.52 4
8808A Accumulator No. 1 to Loop No. 1 24 3 2 (1) Y k'
- 99% of 8 allowable *t hi hlY 8t#'88'd i area 2.59 1.23 1.16 1.63 '
- " ** h18hl# 't#'d 8808a Accumulator No. 2 to Loop No. 2 2h 3 2 (1) , g g ag I ke m at highly stressed 8808c Accumulator No. 3 to Loop No. 3 24 3 2 (1) , , ,
1.90 0.33 1 59 6.06
~8808D Accelator No. 4 to Loop No. 4 24 3 2 (1) $emathighlystnssed 4.16 0.80(18)
}.8809A RHR Heat Exchanger to Cold Legs 24 2.7 1.8 (1) Yoke most highly stressed 1.66 1.68 1,63 1.07 1.0, area: 99% or allowable 88093 RHR Heat Exchanger to Cold Legs ph 2.7s Yoke most highly stressed 1.76 1.76 2.45 (1) area: 4 2.45 1.00 1.00 of allowable SIS Pump Discharge to Cold Legs I tn sed 8821A 6 4 (1) 1.85 1.46 3.24 2.74
>33 ~ ,, ." 8 Yoke nos hkghIy n N essed 8822a SIS Pump Discharge to Cold Lege >33 6 L (1) 1.86 -1.66 area: 87% of allowable 3.23 2.41 .
! 8835 SIS Pump Discharge to Cold Legs >33 6 4 (1) Yoke most highly stressed 2.54 2.63 2.36 1.52 area: 87% of allowable
, 8856A RER Heat Exchanger Outlet Relief s33 8.5 4 (1) Inlet most highly stressed 1.88 1.41 k.52 2.84 area: 16% of allowable i 8856B RHR Heat Exchanger Outlet Relief > 33 8.5 4 (1) Inlet most highly stressed 1.88 2.29 4.52 1.75
- area: 16%ofallowable
$ 8923A Safety Injection Pump No.1 Suction > 33 8.5 4 (1) Bonnet most highly stressed 1.55 1.40 5.48 2.86 Safety Injection Pump No. 2 Suction ' 33 8.5 4 (1) Io t t ig seed j 8923B 1-8974A Safety Injection Pump Recirculation > 33 6 (1) Yoke most highly stressed area: 93% o f allowab - 1 57 0 99 3.82 4.c4 (November 1978) Amendment 7C
l
[] T} ,
f'\
%Y Y l
t l TABLE 7-8 Sheet 7 of 7 VALVE QUALIFICATION HOSGRI EVALUATION Accelerations Accelerations Freq., @, g HE, g @/HE Lccation Function HZ _H_ _V_ Method Results H V H V 837LB Safety Injection Pt: sip Recirculation >33 6 4 (') Yoke most highly stressed 1 57 *99 3.82 4.04 d FW3T to Safety Injection Pump Suction 24 area: 93% of allowable 1.97 2.05 0.94 (17) 8976 3 2 (1) Yoke most highly stressed 1.52 area: 73% of allowable 8980 RW3T to RHR Ptmp Suction 20 1.82 1.32 (1) Yoke most highly stressed 3.85 1.40(18) area:
8982A containment Sump to RHR Ptzp No.1 Suetim >33 3 2 (1) B nnet most highly stressed 1.43 0.90 2.10 2.22 area: 98f. of allowable p 8982B Containment Su=p to RHR Pung No. 2 Suction >33 3 2 (1) Bonnet most highly stressed 1.43 0 90 2.10 2.22 area 961,of allowable _
Yoke most highly stressed }
9001A Containment Spray Pump No. 1 Diocharge 2k 3 2 (1) area: 1001 of allowabl* 2.20 1.73 1.36 1.16 Yoke most highly stressed 9001B Containment Spray Pump No. 2 Discharge 24 3 2 (1) area- will be supported 4,45 1,gg Body most highly stressed 9003A RHR Heat Exch. No.1 to Corttain=ent Spray ' 33 3.21 1.91 (1) area : 95% of allowable 3.21 1.91 1.00 1.00
. Body most highly stressed 9003B RRR Heat Exch. No. 2 to Containment Spray
- 33 3.21 1.91 (1) area : 95% of allowable 2.18 1.73 1.47 1.10 (Novemter 1978) Amendment 72
(v^) (7
%J O
Sheet % if 2 NOTES FOR TABLES 7-5, 7-6, 7-7 & 7-8 (12) Additional piping supports were required in order to reduce accelerations on the valve to acceptable levels.
Although final accelerations on the valve were higher than valve qualification levels, there was sufficient margin (36%) in the valve qualification to accommodate the slightly increased accelerations from the piping analysis.
(13) Modifications to this valve (8 BAS 4D) were required in order to increase its fundamental frequency to greater than 33 Hz. Modifications are discussed in Chapter 13.
(14) The 3% overstress condition for this valve (6RV58M5B) is minimal and not sufficient to invalidate the qualification of this valve. Additionally, the calculated stresses are for 69H and 4gV accelerations which are substantially higher than the valve experiences during the Hosgri event.
(15) The RCS seal table and parts were qualified for loads obtained from the response spectra analysis of the RCS instrumentation and couplings (Item 20 in Table 7-6).
(16) The anchor bolts for the positive displacement charging pump were marginally overstressed for 39 horizontal and 29 vertical. However, for actual Hosgri accelerations, stresses in these bolts are 25% of allowable.
All other evaluations were for 3g horizontal and 29 vertical.
(17) Although the Hosgri acceleration is slightly greater than the qualification acceleration, there is more than enough margin between the calculated stress and the allowable stress to make this increase acceptable.
(18) These accelerations were calculated using a model which incorporated a flexible representation of the valve. f (19) Qualification stress shown includes operating loads which do not apply.
Amendment 72 (November 1978)
TABLE 7-11 LIST OF MECHANICAL EQUIPMENT SEISMICALLY TESTED A. Items Tested In-Place In The Plant
- 1. Diesel Generators & Accessories
- 2. Component Cooling Water Heat Exchangers
- 3. Component Cooling Water Surge Tank l
- 4. Boric Acid Tank -
- 5. Liquid Hold-Up Tank ,
- 6. CO2 Tank
- 7. SIS Accumulator Tank
- 8. Evaporator Feed Ion Exchanger Demineralizer 72
- 9. Seal Water Heat Exchanger
- 10. RHR Pump
- 11. Seven Limitorque Actuated Valves (on line)
- 12. One Rotork Actuated Valve (on line) l
- 13. Five Valves With Pnaumatic Actuators (on line) n 14. Five Piping Systems of Various Sizes U 15. Ten Pipe Hangers & Restraints B. Items Tested On Laboratory Shaker Table:
- 1. Two Main Steam Safety Valves
- 2. One 14" Limitorque Operated Valve
- 3. Two Pneumatic Operated Valves O
(November 1978) Amendment 72
The table is organized on the basis of the major systems included and lines are given in numerical order within each system. Modifications were made where seismic stresses exceeded allowable stresses, but some lines which were not over allowable stresses have also been affected. Some lines in ,
the table are not represented by specific stress values but have the note "See 8.1.1." This indicates that these lines.were qualified by the seismic ,
72 l restraint spacing criteria described in Sections 8.1.1 and 8.1.3. A few 1ines in the table have the note "N.R." This stands for, "Not Required",
and means that these lines are extremely short, embedded in concrete, or otherwise obviously adequately supported to assure low seismic stresses.
Lines in this category have been re-reviewed to verify this conclusion. l Piping in the reactor coolant loop is discussed further in Section 8.3.
i I
O J
O (November 1978) 8-9 Amendment 72
8.4 PIPE SUPPORTS AND HANGERS Pipe supports and hangers were designed to ANSI and AISC code criteria.
Stress levels are within material yield strengths for all members and welds for both DE and DDE conditions.
For evaluation of supports to determine the additional capability to withstand Hosgri loads, new acceptance criteria were developed. These are listed, as are the previous design criteria, in Table 8-9. Because standard designs and structural members were used, in most cases actual stresses are considerably below the maximum allowable stresses given in the table.
All hangers located by the spacing criteria shown in Table 8-1 and described in Sections 8.1.1 and 8.1.3 are being re-evaluated for the loading conditions shown in Table 8-10. If the re-evaluation requires physical modifications, they are made.
All hangers with loads generated by the response spectra piping analysis techniques described in Section 8.2 are being re-evaluated and modified, if necessary.
Approximately 900 of the 5,000 re-evaluated hanger assemblies have required some modification. As any changed design was completed, drawings were directly issued for construction. About 200 of these modifications were requirad to prevent uplift of the lines qualified by the seismic restraint spacing criteria. This was done as a result of the generalized modeling technique and does not necessarily indicate 72 that there are uplift seismic forces on these lines. Furthermore, nearly 150 hangers were upgraded from Design Class II to a Design Class I design to provide seismic-resistant pipe supports to Class II. This Class II piping was delineated by the NRC as necessary for a long-term cool-down water supply.
O (November 1978) 8-13 Amendment 72
l l
I TABLE 8-3 l
V (Page 1 of 18)
STRESS EVALUATION - CIASS I PIPING SYSTEMS Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi)- (psi) to Allowable Stress System 3 - Feedwater System and Auxiliary Feedwater 476 4 13610 30718 .443 488 6 604 34n2 .018 489 6 kn 34n2 .012 554 16 19594 30130 .650 SM 16 16340 29989 545 557 72 556 16 16586 29777 557 16 10376 30393 341 558 8 569 34054 .o17 559 6 628 34n2 .018 560 6 287 34n2 .008 561 8 466 34054 .014 l72 562 8 544 34054 .o16 563 4 5417 31450 .172 568 6 6235 307o4 .203 569 3 15462 31046 .498 570 3 16529 3n 54 531 571 3 10042 31943 314 572 3 18499 31035 596 72 573 4 14285 3o629 .466 574 4 21805 3125o .698 575 3 14143 31325 451 72 576 3 13401 31325 .428 577 3 8590 31489 .273 578 3 12732 31478 .404 l72 4292 4 3168 3o718 .103 (November 1978) Amendment 72
Table 8-3, Page 2
(~] Line Nominal Diameter Calculated Allowable Ratio of Calculated
(/ Hosgri Stress Hosgri Stress fjumber (inches) (psi) (psi) to Allowable Stress System 4 - Turbine Steam Supply System 225 28 3227 34741 .093 226 28 14431 34686 .416 227 28 12771 34554 370 228 28 11017 33931 325 593 4 17708 30979 572 72 594 4 14953 30979 .h83 760 4 5354 30229 .177 1040 2.5 21533 32498 .663 1041 2.5 14833 32498 .456 1042 2.5 8666 32498 .267 1043 2.5 19931 32498 .613 1045 lo 7433 34500 .215 1065 24 17179 29744 578
(
1066 24 3719 29639 .125 72 3915 3 NR - -
3916 3 NR - -
3917 3 NR - -
3918 3 NR - -
System 7 - Reactor Coolant System 1 29 5700* 54700* .104*
2 29 5700* 54700* .104*
3 29 5700* 54700* .104*
4 29 5700* 54700* .104*
5 31 5700* 54600* .104*
6 31 5700* 54600* .104*
7 31 5700* 54600* .104*
8 31 5700* 54600* .104*
e 9 27.5 13700* 54700* .25*
10 27 5 13700* 54700* .25*
- S h
= 3.6 (November 1978) Amendment 72
Table 8-3, Page 3 jO Nominal Calculated Allowable Ratio of Calculated V Line Diameter Hosgri Stress Hosgri Stress Number (inches) (psi) (psi) to Allowable Stress System 7 - Reactor Coolant System (Cont.)
11 27.5 13700* 54700* .25*
12 27.5 13700* 54700* .25*
13 4 18453 34079 541 14 4 19964 33195 .601 25 4 19068 33409 571 '72 16 14 15600* 53500* .292*
17 6 1290 35300 .037 19 8 3945 36166 .109 20 8 2887 36176 .080
- 72 21 8 4556 36131 .126 23 12 10967 34828 .315 24 3 24657 34596 .713 fD J
109 14 9800* 54600* .179*
727 6 11542 34864 331 728 6 10307 35023 .294 729 6 10175 34566 .294 730 6 2670 35367 .075 1141 3 8375 35362 .237 1147 3 4186 35032 .119 1153 3 13833 35644 388 1158 3 5192 34936 .149 1171 3 8522 34824 .245 1172 3 7269 35337 .206 72 1195 3 14501 35178 .412 2057 12 10967 34828 315 2753 3 10851 34781 312 2754 3 9729 35404 .275 2758 3 12127 33940 357 2998 4 3591 36900 .097 3 2999 4 8288 36623 .226 i
(Q' 3000 3 See 8.1.1 - -
- S h = 3.6 (November 1978) Amendment 72
Table 8-3, Page 4 Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 7 - Reactor Coolant System (Cont.)
3248 4 7906 35539 .222 3468 3 5188 35518 .146 3489 3 2165 34541 .o62 3495 3 8390 35165 .239 72 3496 3 9047 35304 .256 3798 3 5126 35911 .143 3799 3 1462 35483 .041 3800 3 10421 35881 .290 3801 3 2679 36097 .o74 4081 4 6264 34468 .182 System 8 - Chemical and Volume Control System O 25 3 11602 38692 300 41 4 5557 41295 .135 42 6 3852 41760 .092 72 43 6 4682 42123 .111 44 4 4668 42362 .110 45 4 5972 39947 .149 46 4 3559 39679 .090 47 3 1220 40096 .030 48 3 3581 40096 .o89 49 3 28o1 38190 .073 72 50 3 9984 34794 .287 52 3 1249 40096 .o31 53 3 1o75 40096 .027 62 4 3616 40886 .088 l72 65 4 See 8.1.1 - -
75 4 5538 42056 .132 p' 76 77 4 4 4863 5773 42056 42056
.116
.137 l72 78 4 5773 42056 .137 79 4 See 8.1.1 - -
(November 1978) Amendment 72
Table 8-3, Page 5 Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stres6 System 8 - Chemical and Volume Control System (Cont.)
80 4 3702 42056 .088 PV) 3 23689 34794 .681 72 4.32 3 See 8.1.1 - -
535 3 See 8.1.1 - -
536 3 See 8.1.1 - -
537 3 See 8.1.1 - -
744 3 See 8.1.1 - -
746 3 See 8.1.1 - -
747 3 See 8.1.1 - -
751 3 See 8.1.1 - -
905 4 See 8.1.1 - -
1119 3 See 8.1.1 - -
1312 3 1989 40096 .oso 1343 4 See 8.1.1 - -
1345 4 See 8.1.1 - -
1347 4 See 8.1.1 - -
1453 4 See 8.1.1 - -
1454 6 7846 39973 .196 1455 3 See 8.1.1 - -
1456 8 7386 42363 .174 72 1458 4 See 8.1.1 - -
1474 3 10664 41078 .260 1475 3 4170 40731 .102 1503 3 See 8.1.1 - -
1504 3 See 8.1.1 - -
1508 4 See 8.1.1 - -
1509 3 See 8.1.1 - -
1519 3 See 8.1.1 -
1 1522 3 See 8.1.1 -
1527 3 See 8.1.1 - -
~1528 3 See 8.1.1 - -
_ (November 1978) Amendment 72 l
Table 8-3, Page 7
['v] Line Nominal Diameter Calculated Hosgri Stress Allowable Hosgri Stress Ratio of Calculated i
j Number (inches) (psi) (psi) to Allowable Stress System 9 - safety Injection system 221 18 6389 40623 .157 222 4 25243 42238 598 72 223 12 3619 36h26 .099 235 6 14000* 55000* .255*
236 6 16500* 55000* 3*
237 6 9500 55000* .173*
238 6 14400* 55000* .262*
253 lo 17300* 55100* 314*
254 lo 20600* 55100* 37h*
255 lo 15100* 55100* .274*
256 10 14300* 55100* .260*
261 12 24548 42240 .581 282 3 7699 39826 .193 283 3 6618 40660 .163 508 8 17100* 59400* .288*
509 8 16700* 59400* .281*
512 14 1947 40640 .048 513 14 7109 ho6ho 175 734 8 12104 40582 .298 72 1016 3 4756 41h24 .115 1294 10 13900* 56200* .247*
1295 lo 20600* 56200* 367*
12 % lo 12400* 56200* .221*
1297 lo 12800*' 56200* .228*
1972 4 4504 41193 .109 72 1974 4 7536 39271 .192 1975 4 6266 39271 .160 1977 4 15485 37842 409 1978 4 6905 38764 .178 3 1979 4 6209 39239 .158 (O *S h
= 3.6 Amendment 72 (November 1978)
Table 8-3, Page 8
[' Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 9 - Safety Injection System (Cont.)
1980 4 30050 38299 .785 1981 4 9921 37150 .267 1982 6 2574 40686 .063 1983 6 1532 40686 .038 1984 8 19932 36358 548 1985 4 4393 41193 .107 72 1986 8 13172 37272 353 1987 8 11680 40144 .291 72 1988 8 9484 38695 .245 "995 3 See 8.1.1 - -
2003 3 5319 36747 .145 2004 3 11867 36747 323 2032 6 5064 40615 .125 2033 4 8365 38964 .215 2641 4 See 8.1.1 - -
2749 14 3352 43283 .077 2750 14 See 8.1.1 - -
3136 4 NH 42617 -
3844 6 12900* 59700* .216*
3845 6 22400* 59700* 375*
3846 6 10700* 59700* .179*
3847 6 13900* 59700* .233*
3848 4 15443 37936 407 3849 4 8558 38120 .225 3850 4 4991 38077 .131 3854 4 18919 39483 479 3860 4 28138 40575 .693 3868 3 9792 41423 .236 3869 3 15065 41423 364 4296 4 3906 41193 .095
- S h = 3.6 (November 1978) Amendment 72
I Table 8-3, Page 9 O Nominal Calculated Allowable V Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 10 - Residual Heat Removal System no 14 5588 32900 .170 111 14 5951 32852 .181 n2 8 3716 33656 .no n3 8 4790 33656 .142 j 118 8 5363 32517 .165 l 8 8730 33323 .262 72 119 120 12 1313 37920 .035 224 8 30224 40366 .749 279 8 4557 33814 .135 280 8 3949 33215 119 72 735 8 4456 35619 .125 927 14 9800* 54600* .179*
930 3 10464 36312 .288 931 3 9096 35988 .253 72 985 12 17270* 60800* 517*
1661 8 7933 33656 .236 1663 8 5672 72 33656 .169 1665 14 9800* 54600* .179*
1669 8 10480 32616 321 1971 8 5081 33861 .150
..e 2212 8 10063 33225 303 2458 8 6207 34032 .182 2575 8 14000* 54700* .256*
2576 8 16400* 54700* 3*
3551 14 2446 32900 .074 System 12 - Containment Spray System 262 lo 5270 43030 .122 263 lo 5149 43030 .120 264 8 17830 39146 455
- S h
= 3.6 (November 1978) Amendment 72
Table 8-3, Page lo O Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress l l
1 System 12 - Containment Spray System (Cont.)
265 8 13545 27926 .485 266 3 See 8.1.1 - -
267 3 See 8.1.1 - -
268 3 See 8.1.1 - -
l 270 10 32770 34699 944 l 271 lo 23852 37970 .628 325 8 20603 31690 .650 326 8 5060 382n .132 907 4 16718** 4o439 .413** l 2371 6 n654 40204 .290 2372 6 14145 40204 352 2373 4 14318 40399 354 2374 8 6992 4o117 .174 l 2375 6 20865 40204 519 72 '
2376 8 16718 kon7 .417 2377 6 16718** 4o256 .415**
2509 3 See 8.1.1 - -
2510 3 See 8.1.1 - -
l 2511 3 See 8.1.1 - -
2519 8 8272 37732 .219 2520 8 13365 37732 354 2648 3 See 8.1.1 - -
2649 3 See 8.1.1 - -
2650 3 See 8.1.1 - -
2651 3 See 8.1.1 - -
3153 3 See 8.1.1 - -
i System 13 - Spent Fuel Fool Cooling System 725 4 See 8.1.1 - -
1068 3 See 8.1.1 -
- These lines are enveloped by line 2376.
(November 1978) Amendment 72
Table 8-3, Page 11 Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 14 - Component Cooling Water System 81 20 4446 32612 .136 l 82 20 7899 32612 .242 89 20 5885 33869 .174 90 20 3137 32612 .o96 91 20 3863 32612 .119 94 12 3721 33337 .112 95 30 18403 34493 533 96 30 7755 34493 .225 97 20 10873 32612 333 72 98 24 4940 32212 .153 99 12 6950 33337 .208 101 30 7417 34493 .215 102 30 22484 34493 .652 103 20 20927 32612 .642 104 20 12128 32612 372 105 18 16771 32812 5u 106 18 8460 32812 .258 n6 6 10 % 1 34576 317 117 6 10 % 1 33724 325 121 6 13342 33685 396 123 6 15950 33435 .477 124 12 8205 33337 .246 125 8 3937 33607 .117 126 8 11734 33607 349 127 12 14502 33337 .435 72 129 4 21584 33899 .637 130 4 5553 33899 .164 133 10 14264 33507 426 134 4 2473 34436 .o72 f 135 4 24926 34148 730 136 4 17155 33899 506 (November 1978) Amendment 72
Table 8-3, Page 12 Nominal Calculated Allowable Line Diameter Hosgri Stress Hoagri Stress Ratio of Calculated ,
Number (inches) (psi) (psi) to Allowable Stress System 14 - Corzponent Cooling Water System (Cont.)
137 4 19836 33899 585 142 6 11007 33724 326 143 6 13735 33724 .407 72 144 4 5567 33899 .164 (
145 4 6235 33899 .184 146 16 9131 33012 .277 147 16 8478 33012 .257 72 148 10 4303 33507 .128 149 10 5848 33507 .175 150 8 3266 33607 .097 l
151 8 3541 33607 .105 152 12 6831 33337 .205 O 12 33337 .106 72' V 153 3543 157 6 10961 33724 325 180 10 6497 33507 .194 314 12 19404 33337 372 315 12 12848 33337 385 r 15469 33337 .464 72 316 317 12 13404 33337 .402 318 12 21113 33337 .633 319 12 19126 33337 574 l72 320 12 12848 33337 385 321 12 15469 33337 .464 322 12 13404 33337 .402 72 323 12 23702 33337 711 1357 6 12586 30556 .412 1575 4 2601 33899 .o77 1576 4 2601 33899 .o77 1577 4 2601 33899 .o77
' 6 .258 1759 8695 33724 1760 6 8942 33724 .265
, (November 1978) Amendment 72 l
l
Table 8-3, Inge 13
, Nominal Calculated Allowable
\ Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 14 - Component Cooling Water System (Cont.)
1763 4 8987 33899 .265 1764 4 7353 33899 .217 1767 6 6205 33724 .184 1768 6 6526 33724 .194 1775 4 2242 33899 .066 1776 4 1422 33899 .042 1951 10 2585 33507 .077 1%8 6 13989 33724 .415 2207 10 3691 33507 .110 72 2211 6 14926 33724 .443 2277 20 13759 32612 .422 2278 18 9098 32812 .277 2279 12 7357 33337 .221 2280 12 3867 33337 .116 2281 18 4726 32812 .144 2282 20 8777 32612 .269 2285 20 10374 32612 318 2286 3 See 8.1.1 - -
2290 3 See 8.1.1 - -
2292 3 See 8.1.1 - -
2293 3 See 8.1.1 - -
2296 8 18954 33607 564 2297 6 12316 35069 351 2298 6 18018 33724 534 2299 3 3475 35043 .099 2300 3 13438 34003 395 2301 3 6606 34003 .194 2302- 3 14457 35040 413 72 2303 4 14730 33899 .435
( 2304 4 12030 33899 355 -
72 2305 4 17080 33899 504 (November 1978). Amendment 72
Table 8-3, Page 14 C
Nominal Calculated Allowable Line Diamot,er Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress, System 14 - Component Cooling Water System (Cont.)
2306 4 5975 33899 .176 2311 6 7865 33724 .233 2312 8 21336 33607 .635 2313 3 11314 32007 353 72 2314 3 23915 31300 764 2315 25 16353 34079 .4 83 2324 25 16527 34079 .485 2340 6 1485: 3056 .486 2341 8 8118 33607 .242 2342 3 10623 31300 339 72 2343 3 4996 31300 .159 2369 18 4823 32812 .147 2378 4 5921 34185 .173 2399 16 1h07 33o12 .043 l72 2400 4 6889 32429 .212 2555 3 See 8.1.1 - -
2836 4 9135 33899 .269 2994 20 5556 33260 .167 3004 6 9999 33724 .296 3005 4 8785 33899 .259 72 3007 10 6918 33507 .206 3036 20 10725 32612 329 3037 3 See 8.1.1 - -
3038 3 17143 34003 504 3039 20 18585 32612 57o 3179 6 6164 33724 .183 3270 4 9694 35006 .277 3272 4 1574h 37269 .422 3279 12 D6285 33337 .788 3280 12 3473 33337 .104 3281 12 5783 33337 .173 (November 1978) Amendment 72
Table 8-3, Fage 15 Nominal Calculated Allowable
/ Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) _.
(psi) (psi) to Allowable Stress System 14 - Component Cooling Water System (Cont.)
3282 12 7858 33337 .236 72 3283 12 27023 33337 .811 3284 12 13775 33337 .413 3285 12 2015 33337 .060 3286 12 19580 33337 587 3287 12 14536 33337 .436 3288 12 6981 33337 .209 3289 25 12485 34079 366 3290 2.5 14208 34079 .417 3291 4 7458 33399 .220 72 3292 4 18940 33899 ,559 3480 3 See 8.1.1 - -
3481 3 See 8.1.1 t ]'
3744 3 2351 32358 .073 3745 3 3930 31300 .126 3746 3 3895 31300 .124 3747 3 10419 32483 321 l 3901 3 8145 32893 .248 3902 3 14289 33092 .432 System 16 - Makeup Water System 377 4 See 8.1.1 - -
380 10 1964 34335 .057 383 4 See 8.1.1 - -
384 4 See 8.1.1 - -
385 4 See 8.1.1 - -
386 4 See 8.1.1 - -
638 8 816 33994 .024 1900 4 NR 34400 -
1906 4 See 8.1.1 - -
1917 -4 See 8.1.1 - -
(November 1978)- Amendment 72
l Table 8-3, Page 16 i Allowable O Line Number Nominal Diameter (inches)
Calculated Hosgri Stress (psi)
Hocgri Stress (psi)
Ratio of Calculated to Allowable Stress System 16 - Makeup Water System (Cont.)
2122 4 See 8.1.1 - -
2123 4 See 8.1.1 - -
I 2127 4 See 8.1.1 - -
2242 3 See 8.1.1 - -
2829 4 See 8.1.1 - -
2835 4 See 8.1.1 - -
3006 4 11607 40227 .289 ,
3181 4 See 8.1.1 - - !
3269 4 9046 39592 .228 i 3271 4 14289 42387 337 72 3830 4 NR 33946 -
System 18 - Fire Protection Systems
[]
986 4 NR 33919 -
3831 3 NR 33946 -
3833 3 See 8.1.1 . -
i System 19 - Liquid Radwasto System 1
528 25 See 8.1.1 - -
1248 4 See 8.1.1 - -
1249 4 See 8.1.1 - -
l 1578 3 18873 40125 .470 1597 25 See 8.1.1 - -
1609 3 25788 40125 .643 1653 3 See 8.1.1 - -
1800 3 See 8.1.1 - -
1806 4 See 8.1.1 - -
1807 3 See 8.1.1 - -
1808 3 See 8.1.1 - -
2142 25 See 8.1.1 - -
(November 1978) Amendment 72
Table 6-3, Page 17 O Nominal Calculated Allowable Line Diameter Hosgri Stress Hosgri Stress Ratio of Calculated Number (inches) (psi) (psi) to Allowable Stress System 19 - Liquid Radwaste System (Cont.) l 2565 3 See 8.1.1 - -
2993 4 See 8.1.1 - -
3001 4 See 8.1.1 - -
1 3048 2.5 16102 38739 .416 )
3072 4 See 8.1.1 - -
3331 25 See 8.1.1 - -
l 3332 25 See 8.1.1 - -
3333 25 See 8.1.1 - -
3334 25 See 8.1.1 - -
3335 25 See 8.1.1 - -
3339 3 See 8.1.1 - -
3352 25 See 8.1.1 - - j 3378 3 See 8.1.1 -
1 3379 3 See 8.1.1 - -
)
3486 25 See 8.1.1 - -
3490 4 See 8.1.1 - -
3491 4 See 8.1.1 - -
3729 25 15424 38894 397 System 23 - Containment H Purge System 2
647 12 NR 34477 -
438o 4 See 8.1.1 - -
4381 4 3710 34480 .108 4382 4 3710 34480 .108 4384 4 See 8.1.1 - -
4385 4 3710 34480 .108 72 4386 4 3710 34480 .108 l 4387 4 3710 34h80 .108 1
See 8.1.1 4388 4 - -
Os 4389 4 See 8.1.1 - -
(November 1978) Amendment 72 v wr- - ' g-
O O O TABLE 10-1
SUMMARY
- SEISHIC QUALIFICATION OF CLAS5 IE INSTRUMENTATION Sheet 1 of 3 AND ELECTRICAL EQUIPMENT QUALIFICATION HO5GRI EVALUATIOM Zero Period Zero Period Accelerations Accelerations Para. Item Equipment Elev./ Bldg. QL. (g)* Method Remarks and References OL/HE H V H V H V 10.3.1 1. Annunciator. Main (B.O.P. 128'/ Aux. 3.5 1.5 (5.10) 1.35 0.58 2.6 2.6 la. Tested Components 2.50 1.2 (8), pages 143-213 1.35 0.58 1.85 2.1 10.3.2 2. Auxiliary Safeguards 128'/ Aux. 2.00 1.33 (1.3) WCAP 8021 1.47 0.58 1.36 2.29 10.3.3 3. Battery Chargers (B.O.P.) 115'/ Aux. 2.0 1.3 (8), pages 201-315 1.12 0.56 1.96 2.3 10.3.4 4. Station Battery 115'/ Aux. 2.5 1.1 (4.8) pages 240-254 Group IV 1.14 0.56 2.2 2.0 Battery Racks (B.O.P.) 115'/ Aux. 3.0 0.65 (5.10) 10.3.5.1 5. DC Motor Control Center (TLOP) 115'/ Aux. 2.1 1.3 (4.8). pages 215-239 1.14 0.56 1.84 2.32 10.3.5.2 DC Switchgear 2.5 1.2 (8), pages 255-280 l72 115'/ Aux. 1.12 0.56 2.0 2.14 (B.O.P.)
10.3.5 6. 01esel Generators (B.O.P.) 85'/Turb. 0.85 0.85 (8), pages87-142 0.54 0.50 1.57 1.7 6a. Excitation Cabinet 0.85 0.85 O.54 0.50 1.57 1.7 6b. Engine Control Cabinet 0.85 0.85 0.54 0.50 1.57 1.7 10.3.7 7. Electrical Penetrations 130'/ Cont. 10.5 7.0 (5. 10) see 10.3.7 2.0 1.5 5.25 4.67 (B.O.P.)
~
10.3.8 8. Fire Pump Controller (B.O.P.) 115'/ Aux. 2.1 1.3 (8), pages 214-239 1.02 0.56 2.05 2.32 10.3.9 9. Hot Shutdown Panel (B.O.P.) 100'/ Aux. 3.0 2.0 (5.10)see10.3.9.1 0.97 0.54 3.09 3.70 (Fisher Controller) (8) pages 445-469 10.3.10 l72 10.3.10 10. Static Inverter (NSSS) 115'/ Aux. 1.50 1.00 (1.4) WCAP 7821, pages 447-479 1.16 0.58 1.29 1.72 Tested at resonance 10.3.11 11. Instrument AC Panel 115'/ Aux. 2.9 1.5 (5,6.10) 10.3.11 Wyle Labs Report 1.13 0.56 2.47 2.67 (Breakers) (B.O.P.) 53744-2 10.3.12 12. Instrument Panels 128'/ Aux. 2.0 0.72 (5) see 10.3.12 1/1 0.58 1.4 1.24 PIA B&C (B.O.P.)
10.3.13 13. Local Instrument Panels (B.O.P.) Various Various see 10.3.13.1 Various >l.0 >1.0 (Includes Solenoid Valves) 10.3.14 14. Le<a1 Starters (B.O.P.) Various 2.5 2.5 (8), pages 412-446 1.0 1.5 2.5 1.67 14a. Local Starters Various 2.0 1.2 (8), pages 201-315 1.14 0.56 1.75 2.14 10.3.15 15. Main Control Board (B.O.P.) 140'/ Aux. 2.2 1.6 (1.5.10) see 10.3.15 1.57 1.5 1.4 1.07 15a. Switches and Indicators 140'/ Aux. 2.6 2.6 (6.8) pages 512-529 1.57 1.5 1.66 1.73
- QL denotes maximum acceleration of test table or other input.
- See text. page 10.22b.
(NovemLer 1978) Amendment 72
TABLE (Continued)
SUPHARY - SEISMIC QUALIFICATION OF Ct165 IE INSTRLHENTATION Sheet 2 of 3 AND ELECTRICAL EQUIPMENT QUALIFICATION H05GRI EVAtUATION Zero Period Zero Period Accelerations Accelerations Para. Item Equipment Elev./ Bldg. QL (g)* Method Remarks and References QL/HE H V H V H V 10.3.16 16. Nuclear Instrumentatica 140'/ Aux. 2.00 1.33 (1.4) WCAP 8830, 8831 1.75 0.58 1.14 2.29 System (N555) Tested at resonance 10.3.17 17. P & 6P Transmittu s (N555) 122'/ Cont. 2.0 1.33 (1) see 10.3.17 0.89 0.50 2.25 2.67 (Worst Case) 10.3.18 18. P & AF Transmitters (00P) Various 18 18 see 10.3.18 Various >1.0 >l.0 10.3.19 19. Process Control & 128'/ Aux. 2.00 1.33 (1.4) WCAP 8832. 8833. 8831. 1.47 0.58 1.36 2.29 Protection Equipment (N555) 8624 Tested at resonance 10.3.20 20. Reactor Trip Switchgear (N555) 115'/ Aux. 1.40 0.93 (1.4) WCAP 7821 1.16 0.58 1.21 1.60
. Tested at resonance 10.3.21 21. Safeguards Relay Bd. (B.O.P.) 119'/Turb. 2.7 3.4 (8) pages92-138, 323-339 1.47 1.5 1.83 2.27 72 10.3.22 22. Solid State Protection System 140'/ Aux. 2.00 1.33 (1.4) WCAP 8941, 8694. 8831 1.75 0.58 1.14 2.29 (N555) Tested at resonance 10.3.23 23. Ventilation Control. 140'/ Aux. 3.0 1.3 (8.10) pages 174-201 1.42 0.81 2.1 1.60 Logic (B.O.P.)
10.3.24 24. Ventilation Control. 128'/ Aux. 3.0 1.3 (8.10) pages 143-173 1.47 0.58 2.0 Relay (B.O.P.) 2.24 l72 I
10.3.25 25. Vital Load Center (480 VAC MCC) 100'/ Aux. 1.7 1.2 (8) pages 316-410 0.95 0.53 1.79 2.26 (8.0.P.)
25a. Auxiliary Relay Panels 1.7 1.2 (8) pages 447-479 25b. Fan Cooler Starter 72 1.7 1.2 (81 pages 344-386 25c. 4160 - 480 VAC Transformer 1.7 1.2 (3) see 10.3.25c 10.3.26 26. Vital Seitchgear (4.16KV) 119'/Turb. 3.0 3.0 (1.3.8) pages 19. 25-27 30-33 1.47 1.5 2.0 2.0 (B.O.P.) 36-92. 139-322 GP.I. 58255-1 10.3.27 27. Resistance Temp. 107'/ Cont. 2.00 -
(1.4.10) WCAP 8234A Uncalculated >1.0 >1.0 Detectors (N555) Tested at resonance to 200g 10.3.28 28. Safeguards Test 140'/ Aux. 2.0 1.33 (1.4) WCAP 8021 1.75 0.58 1.14 2.29 Cabinet (N555) Tested at resonance 10.3.29 29. Cable Trays (B.O.P.) Various (5) see 10.3.29 Various 10.3.30 30. Limit Switches (B.O.P.) Various Various Various (4.8) 10.3.30 Various >l.0 >1.0 i
10.3.31 31. Potential Transformers (B.O.P.) 119'/Turb. 2.7 3.4 (8) pages 28,92-138.GP.I. 58255-1 1.47 1.5 1.83 2.27 72 10.3.32 32. Emergency Light Battery Pack Various 2.7 3.4 (8) pages 21-24. Group I 1.47 1.5 1.83 2.27 (B.O.P.) Wyle Report 58255-1 (November 1978) Amendment 72
O O O TABLE 10-1 (Continued) sheet 3 of 3 Notes for Qualification Method Column:
(1) Sine beat vibration tests.
(2) Shake table vibration test for similar equipment.
(3) Results from testing other similar equipnent were used.
(4) One unit or a typical unit was tested.
(5) Mathematical analysis to verify seismic capability of structure only.
(6) Shake table testing of components only.
(7) Additional laboratory testing scheduled.
(8) Additional laboratory testing completed, Wyle Laboratories Report 58255. and 58255-1.
- (9) Table changed to reflect additional laboratory testing completed.
(10) Cabinet, panel or component is rigid; frequency >33 Hz.
)
(November P378) Amendment 72
Table 10-11 Sheet 1 of 2 VITAL LOAD CENTER AUXILIARY RELAY PANEL A Auxiliary Relay Panel 480 Volt Bus G - Relay Summary Relay Relay Contacts Safety Contacts Designation Type Used Requirements fionitored Results 43X-2G-49 Electro.(LOR) 6 N.0.,5 N.C. No Chatter 1 N.0.,1 N.C. No Chatter 43X-2G-57 5 N.O.,4 N.C. --- ---
43X-2G-44 5 N.0.,4 N.C. " --- ---
43X-2G- 1 "
4 N.O.,5 N.C. " --- ---
43X-2G- 2 "
4 N.0. 5 N.C. --- ---
43X-2G- 4 "
5 N.0..,5 N.C. --- ---
42X-2G-49 C-H (Type R) 3 N.0. No Chatter 1 N.0. No Chatter 72
- e 2-2G-64 AGASTAT(Timer) 1 N.C. No Chatter 1 N.0.,1 N.C. No Chatter 2-2G-41 --- ---
, DPX-2G- 1 GE(IC2820) 1 N.C. No Chatter 1 N.0. ,1 N.C. No Chatter DPX-2G- 2 --- ---
K632AX C-H (Type R) 1 N.0, No Chatter 1 N.0. ,1 N.C. No Chatter
- K632BX --- ---
2G2A AGASTAT 2 N.0. No Chatter 1 N.0. ,1 N.C. No Chatter 2G2 1 N.0. --- ---
l 2G1A 2 N.O. --- ---
2G1 1 N.O. --- ---
2G49 2 N.0. --- ---
l CIAX1-G C-H (Type M) 10 N.O. No Chatter 1 N.0. No Chatter CIAX2-G 7 N.O. --- ---
(Novembee 1978) Amendment 72
l Table 10-11 Sheet ? of 2 B Auxiliary Relay Panel 480 Volt Bus H - Re1ay Summary O l Relay Relay Contacts Safety Contacts Designation Type Used Requirements Monitored Results 43X-2H- 1 Electro (LOR) 5 N.0.,5 N.C. No Chatter 1 N.0. ,1 N.C. No Chatter 72 2H1 AGASTAT 1 N.0. No Chatter 1 N.0. ,1 N.C. No Chatter 2HlA 2 N.0. --- ---
2-lH36 "
1 N.0. --- ---
OPX-2H-1 GE(IC2820) 1 N.C. No Chatter 1 N.0.,1 N.C. No Chatter CIAX-H C-H(TypeM) 9 N.0. No Chatter 1 N.O. No Chatter l
l O
I l
O (November 1978) Amendment 72
CHAPTER 11 OUTD0OR WATER STORAGE TANKS INDEX 11.1 Introduction 11.2 Description of Tanks 11.3 Analysis Criteria 11.3.1 Seismic Input 11.3.2 Dynamic Effects of a Horizontal Earthquake 11.3.3 Gravity Load and Hydrostatic Pressure 11.3.4 Vertical Earthquake 11.3.5 Load Combination 11.3.6 Allowable Stresses 72 11.4 Axisymmetric Analyses 11.4.1 Refueling Water Storage Tank 11.4.2 Firewater and Transfer Tank 11.4.3 Condensate Tank 11.5 Non-Axisymmetric Analysis of the Refueling Water Storage Tank 11.5.1 Purpose 11.5.2 Finite Element Model 11.5.3 Analysis Approach 11.5.4 Discussion of Results 11.6 Foundation Analysis 11.7 References (November 1978) 11-i Amendment 72 l j
l l
INDEX OF FIGURES 1
11-1 Outdoor Water Storage Tanks: Site Plan I 11-2 Sections of Tanks 11-3 Blume and Newmark 7.5M Hosgri Spectra - 7% Damping 11-4 Blume and Newmark 7.5M Hosgri Spectra - 5% Damping l
11-5 Refueling Water Tank - Axisymmetric Model 72 l 1
11-6 Firewater and Transfer Tank j 11-7 Refueling Water Tank - Perspective View of Half Tank Model 11-8 Refueling Water Tank - Node Numbers .
11-9 Refueling Water Tank - Steel Shell Elements 11-10 Refueling Water Tank - Concrete Shell Elements 1
11-11 Typical Reinforcement Details in Concrete Shell '
.O 4
i O
(November 1978) ll-ii Amendment 72
CHAPTER 11 OUTDOOR WATER STORAGE TANKS
11.1 INTRODUCTION
This chapter summarizes the dynamic seismic analyses of Class 1 outdoor water storage tanks of Units 1 and 2 of the Diablo Canyon Nuclear Power Plant. The outdoor water storage tanks which are considered Class 1 structures are the refueling water storage tanks, the firewater and transfer tank, and the condensate tanks. Each tank consists of a steel liner and a concrete cover.
The firewater and transfer tank, in addition, has an inner steel tank. A site plan for the tanks is shown in Figure 11-1.
The re-evaluation of the tanks has been accomplished according to the February 8,1977, Diablo Canyon, Specification for Sciamic Review of Major Structures for 7.f>M Hoagri Earthquake. ) Maximum shears, overturning moments ,
and shell forces were calculated at specific nodal point elevations of the mathematical models of the tanks. Based on the results of this investigation, stresses in the existing steel liner were checked, the required steel reinforcement in the concrete cover was determined and the foundation redesigned.
The tanks were analyzed for effects of gravity loading, hydrostatic pressure, and the two horizontal components and one vertical component of the 7.5M Hosgri ground motion.
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11.2 DESCRIPTION
OF TANKS
' The Class 1 outdoor water storage tanks are located just outside the fuel handling building. They will be covered with concrete which varies in I thickness from three feet at the base to twelve inches up to mid-height, then eight inches up to the dome. Studs will be provided to tie the steel liner and the concrete cover together. The tanks will be tied to bedrock with rock anchors and the existing earth fill under the foundation will be replaced with concrete.
There are two refueling water storage tanks, one to service each unit of the pl ant. Each refueling water storage tank is forty feet in diameter, and is intended to store water up to a depth of 51.75 feet. The tank's overall height is approximately 58 feet. The thickness of the steel liner varies from 0.578 '
inch at the base to 0.25 inch at the dome. 72 The firewater and transfer tank, intended to service both Units 1 and 2 of the l plant, is made up of two concentric steel cylindrical tanks connected by a common dome roof, and a concrete cover on the outer tank. The inner cylindrical tank, called the firewater tank, is 32.67 feet in diameter and made of steel plates with thickness varying from 0.802 inch at the base to 0.375 inch at the top. The outer tank, called the transfer tank, is 40.0 feet in diameter and made of steel plates with thickness varying from 0.627 inch at the base to 0.25 inch at the top.
The " - e configuration of the cont. oate tank is similar to the refueling water .,e tank. Each of the two condensate tanks is forty feet in diameter and is intended to store water up to a depth of 46.5 feet. The thickness of the cylindrical steel liner varies from 0.60 inch at the base to 0.25 inch at the top. The dome is made of 0.263 inch plates.
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On Figure li-2 are shown typical sections of the refueling water storage tank, firewater and transfer tank and condensate ta.:k, respectively. Thicknesses of
() the steel _ liner plates are indicated. Figure 11-2 also shows the 14-foot x 14-foot vault opening in the refueling water storage tank concrete shell. The vault openings in the other tanks are slightly smaller.
72 l The steel plates in the refueling water storage tanks are made of SA-240 Type 304L plates, whereas the steel plates in both firewater and transfer tank and condensate tanks are made of SA-516-55 to SA300 plates. The concrete cover will have a minimum compressive strength of 4,000 psi at 28 days. The concrete fill under the existing base slab will have a minimum compression strength of 3,000 psi at 28 days.
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11.3 ANALYSIS CRITERIA l
11.3.1 SEISMIC INPUT Two different postulated 7.5M Hosgri free-field smooth ground response spectra have been developed for the Diablo Canyon Nuclear Power Plant site: one by URS/ John A. Blume & Associates, Engineers (URS/Blume), and another developed independently for the staff of the United States Nuclear Regulatory Commission (NRC) by Nathan M. Newmark Consulting Engineering Services. Figures 11-3 and i 11-4 illustrate the Blume and Newmark 0.759 free-field horizontal spectra for 7% and 5% damping, respectively. Also shown in the figures is the 5% damped i 0.49 ODE free-field spectrum. The vertical input to the outdoor water storage tanks was taken as two-thirds of the 0.759 horizontal free-field ground 72 spectrum.
Comparison of the Blume and Newmark spectra on Figures 11-3 and 11-4 shows that the Newmark spectrum is generally higher than the Blume spectrum. Thus, for the analysis of the outdoor water storage tank, only one set of analysis using the Newmark spectrum was made.
11.3.2 DYNAMIC EFFECTS OF A HORIZONTAL EARTHQUAKE The tanks were analyzed following the method developed by A. S. Veletsos and V. Y. Yang (6) in order to consider the effects of tank flexibility. In their paper, Veletsos and Yang presented an approach to analyze a tank-fluid system as a single-degree-of-f reedom system by assuming that the system vibrates in a fixed configuration along its height. A half-sine vibration configuration, the most conservative of three configurations suggested by Veletsos and Yang, was used in the analysis.
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11.3.3 GRAVITY LOAD AND HYDR 0 STATIC PRESSURE >
Gravity load refers to the self weight of the tank whereas the hydrostatic I
pressure at a point is equal to the product of the liquid density and the height, h, of-the liquid. .Both gravity load and hydrostatic pressure are j static loads and are constant around the circumference of the tank. The net
- inplane shear and torsion due to these loads are zero. The hydrostatic l pressures which were input to the computer program are shown in detail in r Appendix A to Reference 11.
- 11.3.4 VERTICAL EARTHQUAKE l l
3 It was found that the fundamental mode of the empty refueling water storage
{
j tank in the vertical direction is 0.033 second. Thus, it was assumed initially
- that the tank and the fluid act as a rigid mass during vertical motion so l effects of vertical earthquake were obtained by scaling the stresses caused by
- gravity load and hydrostatic pressure by 0.5 (2/3 of maximum horizontal ground acceleration, i.e., 2/3 x 0.75g = 0.50g). This assumption was used in the l
axisymmetric phase of this investigation.
O i At the present time, there is no. accepted procedure to analyze the fluid motion l 1~ i fn a tank due to a vertical earthquake. To consider the possibility that the 1
) fluid may not act as a rigid mass during vertical motion, an amplification !
I factor of 2.0, i.e., the acceleration at zero period of 0.59 is amplified to a i
value of 1.0g, was used in the non-axisymmetric phase of this investigation.
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Effects of vertical earthquake were then obtained by scaling the sum of the dead load and hydrostatic pressure stresses by 1.0.
11.3.5 LOAD COMBINATION l Responses due to the two horizontal components and one vertical component of the 7.5M llosgri ground motion are combined by the square-root-of-sum-of-squares (SRSS) method. That is:
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EQ = (HEI )2 + (HE2 )2 + (VE)2 (1)
O where EQ = total earthquake response HE1 , HE2 = the responses due to the two horizontal components of the ground motion, respectively, and VE = response due to the vertical component of the ground )
, motion.
1 Each of the responses HE and HE 2 , is obtained by taking the absolute sum of the responses due to structure inertia forces, impulsive pressure and convective pressure.
72 The total load used to calculate stresses is obtained as follows:
TL = DL + HS + EQ (2) where TL = Total Load DL = Dead Load HS = Hydrostatic Pressure, and EQ = Earthquake Response, and defined by Equation 7. I 11.3.6 ALLOWABLE STRESSFS Reinforced Concrete The capacities of reinforced concrete structural members were determined in accordance with " Structural Analysis and Proportioning of Members - Ultimate Strength Design" given in AC1 318-71(7), except that unit load factors were used for combining loads.
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s Structural Steel The capacities of structural steel members other than plates were based on the 72 seventh edition of the American Institute of Steel Construction (AISC) Code (8),
The capacities of structural steel plates were based on ASME Boiler and Pressure Vessel Code,Section VIII, Division 2(9).
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11.4 AXISYMMETRIC ANALYSIS O ,
11.4.1 REFUELING WATER STORAGE TANK l
Finite Element Model l
The basic axisymmetric finite element model used in the analysis of the 1 refueling water storage tank is shown on Figure 11-5. The tank is assumed j fixed at the base. The model consists of 30 nodes, 29 steel shell elements and ;
29 concrete shell elements. The steel shell liner and the concrete cover were modeled as axisymmetric shell elements parallel to each other. The global i coordinates (R, Z) of the nodal points and material properties for steel and l concrete are provided in Appendix A of Reference 11. The model was used to l analyze for effects of gravity loading, hydrostatic pressure, structure inertia forces and hydrodynamic loads consisting of impulsive and convective pressures caused by a horizontal earthquake. AXIDYN, a computer program for the static and dynamic analysis of axisymmetric structures by the finite element method, 72 was used in all the analysis runs. I O
Discussion of Results Longitudinal forces and moments, circumferential forces and moments, and in-plane shears at critical points in the steel liner and concrete cover obtained !
from the analysis runs indicate that the impulsive pressure loading and, to a lesser degree, the hydrostatic pressure loading, contribute the most to the total forces and moments. Moments in the steel liner are negligible.
Longitudinal and circumferential forces and moments due to the impulsive pressure loading are shown in Appendix A to Reference 11.
Design of the concrete cover and checking of stresses in the steel liner were made using the results of the axisymetric analysis. Stresses in steel and concrete elements are shown in Tables 1 through 7 of Reference 11.
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11.4.2 FIREWATER AND TRANSFER TANK Finite Element Model The firewater and transfer tank has been analyzed for effects of gravity loading, hydrostatic pressure, horizontal earthquake, and j vertical earthquake. The same analytical procedures used in the analysis of )
the refueling water storage tank are used in the analysis of the firewater and transfer tank and are not discussed here.
The firewater and transfer tank model and analysis results presented here are based on the preliminary sizing of the thickness of the concrete cover.
Thickness was changed; originally the 12-inch section starts approximately eight feet above the base. Based on the analysis results of the refueling water storage tank, the concrete cover should have a 12-inch thickness up to the mid-height of the tank, which is about 25 feet above the base. 72 i
l The basic axisymmetric finite element model used to analyze the firewater and l transfer tank is shown in Figure 11-3. The steel sections and the concrete cover are modeled as axisymmetric shell elements. The model consists of 46 nodes, 27 concrete shell elements and 45 steel shell elements.
The exterior tank is assumed fixed at the base whereas the inner steel tank is considered pinned at the base.
Two loading conditions are considered:
- 1. Case 1 where both inner and outer cylindrical tanks are filled with water up to design level; and
- 2. Case 2 where inner tank is filled to design level while the outer tank is empty.
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A set of analyses for effects of hydrostatic pressure, horizontal earthquake l and vertical earthquake is made for each case. For the first case, hydrostatic pressure, impulsive pressure and convective pressure are assumed exerted by the l
- fluid on the outer tank only. For the second case, hydrostatic pressure,
[ impulsive pressure and convective pressure are exerted by the fluid on the j inner tank only.
Because fluid impulsive pressure was being exerted essentially only on the
- exterior concrete tank and because large inertia forces will also be generated I on the concrete shell, damping of 7% (which is the NRC Regulatcry Guide 1-61
] specified damping for concrete structures) was used in the Case 1 dynamic analysis for the combined effects of structure inertia forces and impulsive pressure.
4 For the Case 2 analysis, however, large inertia forces will be generated 72 ;
on the concrete shell while impulsive forces will be also acting on the ]
interior steel tank. Thus, a modified damping value of 6.3% was calculated j to consider the lower damping value of the steel tank. Because there is no I spectrum specified for 6.3% damping, the more conservative 5% Newmark I
spectrum was used in the analysis for Case 2.
Discussion of Results I
l Longitudinal forces and moments, circumferential forces and moments and inplane l shears at critical points in the two steel tanks and concrete cover for Case 1 and Case 2 indicate that the impulsive pressure loading and the hydrostatic pressure loading contribute the most to the total forces and moments. Moments
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in the steel sections are negligible. l I
The maximum stress intensity at each steel liner section is determined and compared with the allowable stress intensity at the section. Calcul ations indicate that the steel s3ctions are adequately designed. l O
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Comparison of the design forces and moments for concrete elements with those for the corresponding preliminary refueling water storage tank model indicate that the values for the former are less than for the latter tank, that is, at corresponding concrete elements. Therefore, the same steal reinforcement provided for the concrete cover of the refueling water storage tank will be adequate for the firewater and transfer tank. Stresses in steel and concrete elements are shown in Tables 1-7 of Reference ll.
11.4.3 CONDENSATE TANK The structure configuration of the condensate tank is very similar to the y refueling water storage tank. Both tanks have the same inside diameter. The height of the structure, as well as the design liquid depth of the condensate tank, is less than those for the refueling water storage tank by 5.25 feet.
Because of the similarity between the two tanks, it is felt that analysis results of the refueling water tank should apply to the condensate tank. The steel plates specified for the condensate tank are slightly thicker and made of stronger material than those for the refueling water storage tank, thus, they should be adequate. A concrete cover with steel reinforcement, the same as those specified for the refueling water storage tank, would also be adequate O for the condcasate tank.
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11.5 NON-AXISYMMETRIC ANALYSIS OF THE REFUELING WATER STORAGE TANK 11.5.1 PURPOSE In the earlier dynamic analysis, the tank was assumed to be an axisymmetric :
l shell structure and was analyzed using the AXIDYN program. That analysis, j however, did not take into account a 14-foot x 14-foot vault opening in the concrete shell. To get a clearer understanding of the stress distribution in the concrete and steel shell elements around and in the opening area, a more comprehensive three-dimensional model, with thin shell elements, and including the concrete shell opening, was made and analyzed using the computer program SAPIV. Loads were applied statically.
Two distinct features of the SAPIV static analyses of the tank were (1) the consideration of the 14-foot x 14-foot o,aening in the concrete shell, and (2) l the simulation of the impulsive pressure exerted on the structure by applying 72 normal fluid pressures, that varies cosinusoidally along the circumference.
The impulsive pressures were calculated using the square-root-of-sum-of- l squares (SRSS) values of accelerations obtained from the earlier AXIDYN analysis.
11.5.2 FINITE ELEMENT MODEL I 1
Symmetry of the tank about the centerline through the opening was utilized and only one-half of the tank was modeled. Quadrilateral shell elements were used to model all steel and concrete members except for the concrete framing around the opening. Beam elements were used to model the heavy concrete framing. The three-dimensional half-tank model is shown in Figure 11-9. It consists of 427 nodes, 360 steel shell elements, 348 concrete shell elements and 13 concrete beam elements. The tank was fixed at the base against all six degrees of freedom. Because of computer program restrictions, and for modeling convenience, a small one-foot diameter opening was left at the crown of the dome. (See Figures 11-11 and 11-12.)
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Appropriate boundary conditions at the plane of symmetry of the tank structure, i.e., at nodes located on the z-y plane (see Figure 11-10), were determined by the nature of loading considered, i.e., whether the loading is symmetric or anti-symmetric. Dead load and the horizontal earthquake along the z-z direction are loads symmetric with respect to the z-y plane. On the other hand, horizontal earthquake along the x-x direction is anti-symmetric with respect to the z-y plane.
For symmetric loading, joints on planes of symmetry are restrained such that they displace only in those planes. Thus, for dead load and horizontal earthquake along the z direction, joints of the y-z plane, for example, were allowed to displace only in the y and z directions and to rotate only about the x direction. For anti-symmetric loading, joints on planes of symmetry were compelled to displace in an anti-symmetric manner. Thus, for the case where the horizontal earthquake is acting along the x direction, joints located on the y-z plane were restrained against translations in the plane of anti-symmetry and rotational normal to the plane. This modeling technique is discussed further in Reference 10.
O 11.5.3 DISCUSSION OF RESULTS Direct stresses and moments, obtained from the SAPIV analysis runs for dead load, hydrostatic pressure and horizontal earthquake along the x-x and z-z directions, at critical points in the steel liner and concrete cover are given on Tables D-12 through D-25 of Reference 11. The vertical earthquake responses correspond to a value of 1.09 The vertical earthquake responses are thus equal to the sum of the dead load and hydrostatic pressure analysis results.
Tables D-32 through D-33 of Reference 11 give the stress intensity values of critical points in the steel liner section, including the concrete steel opening area. The allowable stress intensity is exceeded at two sections in the concrete shell opening area. To take care of this overst.ress, the plate will be reinforced where the overstress occurs.
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The concrete cover, including the concrete framing around the vault opening,
. was designed for the more critical of the two analyses (axisymmetric and finite element). Design calculations are given on Appendix 0 of Reference 11. A 72 typical section of concrete cover, showing the required steel reinforcement, l is shown on Figure 11-1t,. l l
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11.C FOUNDATION ANALYSIS O
V 11.6.1 ANALYSIS The foundations were redesigned to resist the overturning moments and base shears determined by the axisymmetric tank analysis. To resist overturning each tank will be anchored to bedrock by 46 rock anchors, a safety factor of 1.GS is provided against overturning. Sliding is resisted by friction between the concrete foundation and tne iyedroci:. A safety factor of 1.63 72 is provided against sliding.
High soil pressures required the removal of earth fill from under the existing foundations and replacement with concrete fill down to bedrock. The resulting bedrock pressure is approximately one half of its ultimate capacity.
The concrete foundation was checked for bending and shear stresses and these were found to be within allowables.
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11.7 REFERENCES
O 1. usa /ae"n ^. .me & assec4etes. see4neers. o<eble Ce vc". Sgcetr<eet<c-for Scior? tic Review of bbjor Structures for 7.5M Hoogri Earthquake (Preliminary) revised February 8,1977.
- 2. Ghosh, Sukamar, and Edward Wilson, vynamic Stress Analysis of Axisymmetric l l
Structtwes Under Arbitrary Loading, EERC 69-10, Earthquake Engineering ,
1 Research Center, College of Engineering, University of California, I Berkeley, September 1969.
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- 3. Bathe, K-J. , E. L. Wilson, and F. E. Peterson, SAPIV, A Structural Analycio Prognm for Static and Dynamic Responce of Linear Systems, l EERC 73-11, College of Engineering, University of California, Berkeley, June 1973. l 72 !
- 4. Harding-Lawson Associates, Geotechnical Studies - Intake Structure, Water Stomge Tanko, Dieul Fuel Oil Storage Tanko - Diablo Canyon Nuclear Power Plant, San Luio Obiopo County, California, Apri1 12,1978.
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- 5. U.S. Atomic Energy Commission, Nuclear Reactors and Earthquakes, TID 7024, Washington, D.C. , August 1963.
- 6. Ve1etsos, A. S. , and J. Y. Yang, Dynamics of Fixed-Base Liquid Storage
'lunks, Proceedings of U.S. - Japan Seminar on Earthquake Engineering Research with Emphasis on Lifeline Systems, Tokyo, Japan ~, November 1976.
- 7. American Concrete Institute, Building Code Requiremente for Reinforced concrete, AC1 318-71, Detroit, Michigan,1971.
- 8. American Institute of Steel Construction, Specification for the Design, Fabricat :on and Erection of Structural Steel for Buildings, New York, N.Y., Fe r uary 12, 1969. j Oi V
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- 9. American Society of Mechanica) iineers, ASMF Boiler and Pressure Vessel O Code,Section VIII, Division 2, Alternative Rules for Pressure Vessels, V 1974.
- 10. Heaver, Hi11iam Jr. , Computer Programa for Structural Analysis, D. Van Nostrand Co. , Inc. , Princeton, N.J. ,1967.
- 11. URS/ John A. B1ume & Associates, engineere, outdoor Water Storage Tanke Dynamic Seisnic Analysea for the 7.5M Hongri Criteria, November 1978.
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CHAPTER 12A
,r mg V SEISMIC EVALUATION OF BURIED PIPING SYSTEMS l l
l INDEX 72 12A.1 Method 12A.1.1 Seismic Inputs 12A.1.2 Material Properties and Allowable Stresses 12A.1.3 Analysis Procedures 12A.2 Results 12A.3 Reference c)v f ).
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INDEX OF TABLES 12A-1 Results of Seismic Evaluation of Buried Piping for 7.5M Hosgri Earthquake l
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( CHAPTER 12A SEISMIC EVALUATION OF BURIED PIPING SYSTEMS l I
The auxiliary saltwater (ASW) piping is a buried Design Class I piping system. The piping is 24 inch diameter ASTM A53 Grade B steel pipe. The buried portion of the system is approximately 1600 feet long. The system, as well as the buried Design Class Il reinforced concrete, circulating water intake (CWI) conduits which provide support for the ASW piping, have been 72 reviewed for 7.5M Hosgri earthquake.
The buried Design Class I diesel fuel oil (DF0) piping running along the west side of the turbine building has been re-routed in a concrete trench.
The work is done in conjunction with the buttresses addition to the turbine building. The short buried suction line between the diesel fuel oil tank and its respective fuel oil pump vault is designed conservatively to allow for 2 inches of maximur relative displacement between them. The line, approxi-mately 15 fee' om is 3-inch diameter ASTM Grade B steel pipe. The piping will not be '
assed during the Hosgri earthquake.
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12A.1 METHOD O
O The method and criteria employed in the seismic evaluation of the buried piping are based on the approach jiven in BC-TOP-4A(1), Section 6, entitled
" Analysis of Long, Buried Structures."
12A.1.1 Seismic Inputs The following are the seismic input parameters:
- 1. Maximuni horizontal ground acceleration (Ah ) = 0.75g
- 2. Maximum vertical ground acceleration (Ay) = 0.50g
- 3. Maximum horizontal ground velocity (V h ) = 3.0 fps
- 4. Maximum vertical ground velocity (V y ) = 2.0 fps
- 5. Shear wave velocity (Vs ) = 3600 fps (elevation 50 to 75 f t. above MSL) 72 12A.1.2 Material Properties And Allowable Stresses n
fy Actual material properties as determined by properly substantiated test results were used to formulate acceptance criteria for the buried piping as follows:
The compressive strength of concrete (f'c) was taken as the average of the 28-day or 60-day test values, depending on the original curing period specification. The substantial additional margin of strength associated with the gain in compressive strength by aging was not considered.
Steel yield strength (f )y was taken as the average of actual test values.
The yield strength value used in strength computations was always less than 70 percent of the corresponding average ultimate strength value.
The following codes, in conjunction with actual material strengths, were used in establishing allowable stress levels:
- 1. For reinforced concrete: ACI Code 318-73
i 12A.1.3 Analysis Procedures i The ASW pipes are anchored to the CWI conduits at 40 ft. intervals. In order to assure the structural integrity of the conduits during strong earthquake motion, they were reviewed for the same criteria as the ASW piping.
The portion of a buried pipe far from the ends, and free of any external support other than the surrounding soil, was assumed to move with the ground under the propagation of seismic shear and compressional waves.
With.this assumption the stresses in the pipe were computed as the products of soil strains and the modulus of elasticity of the pipe material. Since i shear waves generally transmit the greatest proportion of the earthquake's I energy, they were used in determining the stresses as follows:
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- 1. Calculate axial and bending stresses due to propagation of a shear wave. l
! a. Maximize the combined stress for an incident angle ($) br en 0 and 45 from the longitudinal axis of the pipe.
- b. Assume vertical seismic component of two-thirds
- izontal component te act simultaneously and combine the rew uses by the square root of the sum of the squares (SRSS).
- 2. Compute stresses due to internal pressure.
- 3. Combine earthquake and pressure stresses (see below) to determine maximum axial and bending stresses.
- 4. Check bolt stresses at flange connections.
Near entry points into buildings, additional stresses are normally induced in pipes by differential movements between the building and the soil. No differential movement occurs in this case since all pipes enter the turbine building and the intake structure at the foundation level. These structures t are treated as fixed based mathematical models in the Hosgri evaluation (see Chapter 4).
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For the CWI conduits, the concrete was calculated to crack and the rebar was assumed to carry the entire tension load.
Internal pressure and stresses induced by 7.5M Hosgri earthquake were combined as follows:
72 Su"Sp+ Seq where:
Su = total stress to be resisted Sp = internal pressure stress Seq = total combined seismic stresses due to horizontal and vertical inputs l
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I 12A.2 Results '
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Results of the Hosgri evaluation are given in Table 12A-1. The maximum stress induced in the ASW piping was calculated as 20 ksi. The maximum I stress in the CWI conduit reinforcing steel was determined as 28 ksi. These maximum stresses were in all cases less than the specified minimum yield strengths for the material of construction.
The steel used f the ASW piping is a ductile material. These steel pipes 72 and fittings are joined by bolting, which provide ductility in the connec-tions. Isolation sleeves or flexible couplings are used where the pipes enter the buildings to accommodate relative displacement between the soil and the buildings. These design considerations aid in making these buried l piping systems earthquake resistant. Their use together with the above l maximum stress results Itad to the conclusion that the Design Class I buried piping system would remain elastic under the Hosgri seismic input. The system is, therefore, considered qualified for the Hosgri event.
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12A.3 Reference
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" Seismic Analysis of Structures and Equipment for Nuclear Power Plant," 72 i
Bechtel BC-TOP-4A (Revision 3, November 1974), Topical Report, Bechtel Power Corporation, San Francisco, California.
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, . . .--...,-,.,,--......._-.....,m , , ~ . . -......,......,.....m..._.._,. m._,,,.,.-..re,...,,,, .,...,y,_,rmnw...-m.,w.,-
O O O TABLE 12A-1 RESULTS OF SEISMIC EVALUATION OF BURIED PIPING FOR 7.5M H0SGRI EARTHQUAKE Maximum Stresses Size & Allowable Yi 1Q Stress Axial Bending Combined Iten Material Stress (ksi)9(I l Induced by (ksi) (ksi) (ksi)
ASW 24 in. dia. steel S p+3eq 12.4S Horizontal Motion 12.5 0.4 Pipes ASTM A53 S = 15 Vertical Motion 8.3 15.5 Grade B Sy = 2.4 (S) Internal Pressure 1.0 0.2(2) 3.3 4.3
= 36 Total 19.8 CW Reinforced Concrete f'. = 3.0 Horizontal Motion 12.5 0.3 Intake Double Box Culvert f s~ = 40.0 Vertical Motion 8.3 0.2 15.5
= 0.9 (fs) Internal Pressure 12.2 Conduits 11.75 ft. sq. Sy 12.2 -
Concrete Re-bars = 36 Total 27.7 ASW Fl anges Sp + Seq I 2.4 St Combined Axial Bolt tensile stress = 22.2 Pipes Class 125# St = 15 and Internal ASTM A181 Sy = 2.4 (St ) = 36 Pressure Grade II 20 1/4 in. & Bolts ASTM A307 Grade B (1) S = code allowable stress Seq = earthquake stress Sp = internal pressure stress St = tensile stress Sy = allowable yield stress f'c = compressive strength of concrete fs = yield strength of rebar (2) Local secondary stress (November 19/8) Amendment 72
CHAPTER 128 BURIED TANKS INDEX 128.1 Method 128.1.1 Seismic Inputs 72 12B.1.2 Material Properties and Allowable Stresses 128.1.3 Analysis Procedures 128.2 Interpretation of Results 128.3 References O
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INDEX OF TABLES 12B-1 Summacy of Cross-Section Data 72 128-2 Summary of Factor of Safety For Tank (Y-Z Plane) i 1
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INDEX OF FIGURES 128-1 Horizontal Contrcl Motion 128-2 Typical Cross Section and finite Element Model 128-3 Schematic Diagram of Procedure 1 128-4 Plan and Elevation 128-5 Cross Section C-C 72 128-6 Cross Section F-F 128-7 Finite Element Model, Cross Section C-C 128-8 Finite Element Model, Cross Section F-F 128-9 Deconvoluted Motion 128-10 Finite Element Model, Cross Sections Z-Z and Y-Y 12B-11 Notation of Model Point and Element Number for Beam Element 128-12 Plane Notations O
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l CHAPTER 12B g
U SEISMIC ANALYSIS OF BURIED TANKS 1 The Diesel Fuel Oil (DF0) Storage Tanks which supply diesel oil to the
! emergency diesel engine generators are Design Class I buried tanks. The I l
two 40,000-gallon capacity tanks are located about 40 feet west of the .
Turbine Generator Building. They are buried end-to-end in a 20-foot deep 72 :
by 14-foot wide trench. The trench was backfilled first with bedding sand and then with sandy clay containing rock fragments. The tanks are fabri-cated of welded carbon steel and protected against corrosion with coal-tar epoxy coating and cathodic protection. An analysis was performed to evaluate the response of the tanks under both static and dynamic loading conditions for the Hosgri earthquake. !
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128.1 METHOD O A static analysis was performed to determine the in situ static stresses.
The results of this analysis were then used to evaluate the response of the i
tanks, under both static and dynamic loading conditions.
!' I 128.1.1 SEISMIC INPUTS 1 1
The Hosgri 7.5M/Newmark Motion, an artificial accelerogram with a peak acceleration of 0.75 g, was used as the control motion for horizontal excitation. The same control motion, with a peak acceleration of 0.5 g, was used for vertical excitation. The time history and response spectrum of the control motion for the horizontal component are shown on Figure 128-1.
j 12B.1.2 MATERIAL PROPERTIES AND ALLOWABLE STRESSES 72 The tanks are fabricated of ASTM A-36 steel plates in accordance with UL l Standard 58 and ASME Boiler and pressure Vessel Code Sections VIII and IX.
Allowable yield stress for structural steel SA-36 is based o Table UCS-23 of ASME Code,Section VIII, 1977.
l 128.1.3 ANALYSIS PROCEDURES 4 1
i Static Analysis 1
The fcrces and displacements in the tanks under static loads were computed l by the two-dimensional finite element analysis. Computer Code SSTIP, an improved version of ISBILD (18) was used. The tank and surrounding soil l were modeled by a finite element mesh as shown on Figure 128-2. Linear ,
elastic properties were assigned to the tank elements and hyperbolic stress-strain and volume change relationships (7 and 24) were used for the soil elements. The stresses, strains and displacements in the soil were calcu-lated by simulating the actual sequence of construction operations. The non-linear stress-dependent stress-strain properties of the soils were
~
approximated by performing the analysis in increments of fill placement.
(November 1978) 128-2 Amendment 72
The values of tangent modulus and Poisson's rr a for each element were re-evaluated during each increment of load tr .onform with the stresses in the element for that increment (7). The internal forces and moments of the structural elements were also computed for each increment.
Case Analyzed The model analyzed (Figure 128.2) consists of eight layers of compacted J
fill, two layers of sand and a layer of concrete. Since the tanks were placed in a relatively narrow trench with rigid bedrock boundaries, only the geometry of the trench was needed for a complete finite element mesh.
Appropriate boundary conditions were imposed on the two sides of the model so that only vertical soil movements were allowed. Due to the symetry, only one-half of the section was modeled and analyzed.
The finite element mesh consists of 32 triangular and rectangular soil elements and 48 nodal points. The tank half-section was simulated using
! eight beam elements. A rigid base was assumed below the concrete foundation.
Incremental analyses were performed to compute the stresses and strains in the soil, and forces and moments in the t::nk. A total of eleven successive layers were used for the stress-strain calculations. The final forces and moments exerted on the tank were computed after the placement of the last fill layer.
Dynamic Response Analysis 1
Dynamic soil-structure interaction (SSI) studies were conducted for the Fuel Oil Storage Tanks to evaluate their behavior under seismic loading conditions.
Approximate three-dimensional (3-D) finite element and one-dimensional wave propagation methods were used.
The general procedures in the SSI studies are summarized in two main categories, Procedures I and II (16). A diagram of Procedure I is shown on Figure 12B-3 and the sequence of steps is given below:
O (November 1978) 12B-3 Amendment 72
- 1. A typical free-field profile and appropriate cross-sections were determined as representative soil-structure models.
- 2. A control motion was defined at finished grade (ground surface Elevation 85) in the free-field profile. The Hosgri 7.5M/Newmark time history of accelerations was used as the control motion in the horizontal direction.
- 3. Using deconvolution analyses, the motion was computed at the base elevation of the soil-structure model.
I
- 4. The base motion obtained in Step 3. vias then used as a horizontal excitation to the soil-structure firite element models. Successive iterations were performed until the noduli and damping ratios i 1
throughout the soil system were compatible with the developed strain 72 levels.
- 5. Steps 2. to 4. were repeated for the vertical excitation analyses using O the dynamic soil proper les obtained from the horizontal excitation analyses without further iterations. The Hosgri 7.5M/Newmark scaled to a j maximum acceleration of 0.E0g was used as the vertical control motion at finished grade.
The sequence of steps employed in Procedure II:
- 1. Typical cross-sections were determined as representative soil-structure j model s. '
- 2. The horizontal control motion was used directly as excitation at the base of the soil-structure finite element models. Successive iterations were performed until a strain-compatible solution was attained.
- 3. Step 2. was repeated for the vertical excitation with the dynamic soil properties obtained from Step 2. without further iterations.
O (November 1978) 128-4 Amendment 72
i The prime products of the SSI analyses were acceleration and moment time
.l histories at selected nodal points. Digitized time histories were then used to compute relative displacements (9) and response spectra. Based i
j on these resultu, an evaluation of the dynamic behavior of the structures was made. One-dimensional analyses (19) were also used to provide additional data for the evaluation.
Computer Codes 2
The computer programs FLUSH (12) and SHAKE (19) were used for the SSI studies. The programs use the complex response method which pennits the
' use of separate damping properties in each element or layer of the model.
The equivalent linear methou (6) was also_ incorporated to simulate non-linear soil behavior.
I
- Soil-Structure Model l
! 72 I i
A plan and profile illustrating the site and subsurface conditions are i shown on Figure 12B-4. The typical cross-sections shown on Figures 12B-5 i and 12B-6 are labeled C-C and F-F. Cross-section C-C has two princi' pal layers; approximately 10 feet of compacted fill underlain by sandstone bedrock (or concrete channels). Cross-section F-F consists of the same two layers but of different thicknesses. The compacted fill begins at Elevation 85 (ground surface) and extends to the point where concrete structures or sandstone are encountered at Elevation 62.
One-dimensional profiles were chosen based on Cross-sections C-C and F-F.
The free-field profile (outside of tank and backfill) was represented by sandstone throughout because it is consistent with the overall site conditions in this area.
Soil Properties and Parameters: The properties which are the basis for the one-and two-dimensional SSI analyses are shear modulus at small strain levels, G max (or shear wave velocity, Vs); damping ratio 6; Poisson's ratio, p; and unit weight, Y. The values of shear _ modulus and damping ratio are strain-dependent. The variations of shear modulus and damping with strain have been developed for the various soil types.
(November 1978) 128-5 Amendment 72 i
Finite Element Model O The finite element models used are shown on Figure 128-7 for Cross-section C-C and on Figure 12B-8 for Cross-section F-F. The models consist mainly of the compacted fill, sandstone, and a fuel oil storage tank.
The effects of the Turbine Generator Building and Intake and/or Discharge Conduits on the response of a Fuel Oil Storage Tank were also included by modeling those structures in the finite element mesh.
Cross-section C-C was divided into 249 triangular and rectangular elements with a total of 298 nodal points. Cross-section F-F was divided into 249 triangular and rectangular elements with a total of 295 nodal points. The fuel oil stcrage tank was modeled using 16 beam elements. The third dimension used was 66 feet, which is the length of the tank foundation.
Cases Analyzed N To simulate the horizontally layered system, an energy transmitting boundary was attached at the right hand sides of the models. A rigid base was assumed at Elevation 20 because the shear wave velocity at this elevation is more than 5,000 feet per second. The SSI analyses were conducted using Procedure I described previously for Cross-section C-C and F-F. The design response spectrum and its accelerogram (control motion) were defined at finished grade in the free-field. The base input motion at Elevation 20 (deconvoluted motion) and its response spectrum which produces the given control motion at ground surface in the free-field profile, are shown on Figure 128-9.
The tanks were also analyzed using Procedure II, which proved to be more conservative than Procedure 1. Cross-sections Z-Z and Y-Y (shown on Figure 12B-10), which have identical geometrics, were analyzed using Procedure II.
The case represented by Section Z-Z, useo upper-bound variations of properties and Section Y-Y used measured properties. The control motion shown on Figure 12B-1 was applied directly at the concrete foundation of the tank (i.e., no A
(November 1978) 128-6 Amendment 72
l l deconvolutional procedure was used). The left-hand sides of both Sections j Z-Z and Y-Y were attached' to the energy transmitting boundaries. A suninary l of the features of each cross-section is tabulated in Table 128-I.
l _ Approximate Three-Dimensional Finite Element Analyses: A storage tank was l J 1 j modeled using 16 beam elements. The nodal point and element numbers for
- j. the finite element model are shown on Figure 128-11. Maximum moment, shear, j and axial forces induced' in cach beam element during horizontal and vertical
- excitations were calculated. Based on the results, the following conclusions i can be made
l 1. Conservative values of maximum axial forces are observed it.
1 i Cross-section Z-Z.
- 2. Conservative values of maximum moment are observed in Cross-sections d
Y-Y and Z-Z.
l
- 3. The maximum shear forces throughout Cross-sections Z-Z and Y-Y are
. similar.
J
- 4. The smallest responses, of the four cross-sections, always occur in
) Section C-C.
i i
i i ,
- l 1
4 4
1 l O
(November 1978) 128-7 Amendment 72
128.2 INTERPRETATION OF RESULTS O Introduction Evalua'. ion of the response of the tanks is performed based on comparing the total stresses induced by both static and dynamic loadings and relative displacements with the strength of the tank material.
Definition of Planes Several planes and directions have been defined for clarity. The orientations of planes X-Y, Y-Z and X-Z with respect to the tank are shown '
on Figure 12B-12. The total forces and moments due to static and dynamic loads, and acting on a plane perpendicular to the longitudinal tank axis, were evaluated. Therefore, stress analysis was performed on the Y-Z plane.
Relative displacements of the tank in the X-Z and X-Y planes were also 72 eval uated.
Stress Analysis The axial and shear forces and moments on the beam elements were converted into their corresponding stresses. The converted stresses were used to calculate the factors of safety against shear, buckling and bending failures.
The factors of safety against buckling and bending failures for each element, were computed for two conditions, 1) combined compression with bending and
- 2) tension with bending. Finally, the shear stresses were compared with the !
shear strerigth of the structural steel to compute the factors of safety against shear failure.
Static and Dynamic Loading: Static and dynamic forces and moments on each nodal point of the tank elements were added algebraically. The total forces and moments were converted to their corresponding stresses. Table 12B-11 summarizes the total stresses for each case analysis. Section C-C is not l
presented because it is the least critical case. Furthermore, in each l section, only the tank element with the maximum induced stress is shown in '
the table. The factors of safety against different modes of failure are (November 1978) 128-8 Amendment 72 l
l U
l I
also summarized in the table. Comparison of the results shows that l y Section Z-Z is the most critical. The minimum factor of safety against )
i (G tension or compression with bending is 1.8. l 1
i Relative Displacement between Tank and Pump Vault i
A three-inch diameter pipeline extends between each tank and the fuel oil l pump vault. During ground shaking the pipeline is expected to move with the soil which surrounds it and the design allows for 2 inches of maximum relative displacement between tank and pumpvault. ,
l The calculated maximum relative displacement between the top of a tank and the foundation level of the Turbine Generator Building is 0.26 inches 72 in the horizontal direction and 0.04 inches in the vertical direction for l the Cross-section F-F. The resultant of the horizontal and vertical relative displacement is therefore 0.26 inches. The pump va 11t is only ]'
20 feet from the tanks. Therefore the relative displacement between tank and vault is less than 0.26 inches. Essentially no relative displacement Q is expected within a few feet. Therefore the pipe to tank connection is not overstressed.
j Diameter Change of Tank The computed maximum diameter change is less than 0.1 inches for Cross-section Z-Z. A diameter change of 0.1 inches will produce about 4.0 X 106 psf of stress (5) at the connection between the tank and I tank cap.
I O
O (November 1978) 12B-9 Amendment 72
12B.3 REFERENCES
- 1. Bowles, J.E. , " Foundation Analysis and Design," pages 379, 380, 387 and 388, McGraw-Hill Book Company, Second Edition,1977.
2 Christian, J.T., " Relative Motion of Two Points During an Earthquake,"
Journal of the Geotechnical Engineering Division, ASCE, Vol.102, No. GT11, Technical Notes, November, 1976, pp. 1191-1194. !
3 GE0-RECON, INC., Letter, dated September 18, 1968 to John A. Blume &
Associates; Diablo Canyon Nuclear Power Plant Project, Determination of Compressional and Shear Wave Velocities.
4 Harding-Lawson Associates, " Geophysical Investigation of Compacted Earth Fill," A Report to PG&E, March 8,1978.
72 5 Harding-Lawson Associates and John A. Blume & Associates, Engineers, p Telephone Conversations and Memos of March 27, 1978, R. Villatuya and T. Udaka. l
- 6. Idriss, I.M. , Dezfulian, H. , and Seed, H.B. , " Computer Programs for Evaluating Seismic Response of Soil Deposits with Non-Linear Characteristics Using Equivalent Linear Procedures," Research Report, Geotechnical Engineering, University of California, Berkeley,1969.
- 7. Kulhawy, F.H., Duncan, J.M. and Seed, H.B. , " Finite Element Analyses of Stresses and Movements in Embankments During Construction,"
Report No. TE-69-4, Office of Research Services, University of California, Berkeley, California,1969.
- 8. Lambe, T.W. , " Soil Mechanics," pp. 182-185, John Wiley & Sons, Inc. ,
1969.
- 9. Lysmer, J. and Tsai , C.F . , " FORCE 2 An Auxiliary Program for the Computer Program LUSH 2," Draft Report, EERC, University of California, O- Berkeley, 1974.
(November 1978) 128-10 Amendment 72
- 10. Lysmer, J. , Seed, H.B. , Udaka , T. and Hwang, R.N. , " Efficient Finite Element Analysis of Seismic Structure-Soil-Structure Interaction,"
Second ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, December,1975.
- 11. Lysmer, J. , Udaka , T. , Seed, H.B. and Hwang, R. , " LUSH - A Computer Program for Complex Response Analysis of Soil-Structure Systems,"
Earthquake Engineering Rese6rch Center, University of California, Berkeley, Report No. EERC 74-4, April,1974.
- 12. Lysmer, J. , Udaka , T. , Tsai , C.F. , and Seed , H.B. , " FLUSH - A Computer Program for Approximate 3-0 Analysis of Soil-Structure Interaction Problems," Earthquake Engineering Research Center, University of California, Berkeley, Report No. EERC 75-30, November,1975.
- 13. Manual of Steel Construction, Seventh Edition, American Institute of Steel Construction, Inc., 1970. 72 I
U 14. Mononobe, N. and Matsuo, H., "On the Determination of Earth Pressures During Earthquakes," Proceedings, World Engineering Conference, Vol . 9,1929.
- 15. Newmark, N.M., " Problems in Wave Propagation in Soil and Rock,"
Proceedings of the International Symposium on Wave Propagation and Dynamic Properties of Earth Materials, Albuquerque, N.M.,1967, pp. 7-26.
- 16. Nuclear Regulatory Commission, Standard Review Plan: Section 3.7.1, Seismic Input, June 1975.
- 17. Okabe, S., " General Theory of Earth Pressure," Journal of Japanese Society of Civil Engineers, Vol.12, No.1,1926.
!O l
(November 1978) 128-11 Amendment 72
- 18. Ozawa, Y. and Duncan, J.M., "ISBILD: A Computer Program for Analysis
> of Static Stresses and Movements in Embankments," Report No. TE-73-4, i Office of Research Services, University of California, Berkeley, Cali fornia , 1973.
- 19. Schnabel , P.B. , Lysmer, J. , and Seed, H.B. , " SHAKE - A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites,"
Report No. EERC 72-12, Earthquake Engineering Research Center, University of California, Berkeley,1972.
1 20. Seed, H.B. and Idriss, I.M., " Soil Moduli and Damping Factors for Dynamic Response Analysis," Report No. EERC 70-10, Earthquake Engineering Research Center, University of California, Berkeley,1970.
- 21. Seed, H.B., and Whitman, R.V., " Design of Earth Retaining Structures y for Dynamic Loads, Specialty Conference - Lateral Stresses and Earth
, Retaining Structures," American Society of Civil Engineers,1970.
O 22. Udaka, V T., " Analysis of Response of Large Embankments to Travelling Base Motions," Doctoral Dissertation, College of Engineering, University of California, Berkeley, Fall,1975.
I 23. Udaka, T., Lysmer, J. and Seed, H.B., " TRIP and TRAVEL - Computer ,
l Programs for Soil-Structure Interaction Analysis with Horizontally i
Travelling Waves," Earthquake Engineering Research Center, University of California, Berkeley, Report No. EERC 75-32, December,1975. l I
- 24. Wong, K.S. and Duncan, J.M. , " Hyperbolic Stress-Strain Parameters for Non-Linear Finite Element Analysis of Stresses and Movements in )
l Soil Masses," Report No. TE-74-3, Office of Research Services, University of California, Berkeley, California,1974. !
l i
- O l (November 1978) 128-12 Amendment 72
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i 5 TOTAL FORCE TOTAL STPISS j DYNAMICS STATICS AND MO?ENT (PSF) FACTOR OF SAFETY E
M COMBINED STRESSES OF g MOMENT AXIAL SHEAR MOMENT AXIAL SHEAR MOMENT AXIAL SHEAR MOMENT AXIAL SHEAR COMFRESSION OR TENSION 8 -(FT-LB) (LB) (LB) (FT-LB) (LB) (LB) (FT-LB) (LB) (LB) WITH BENDING MOMENT
-U FF 1.946x10 1.23Tx104 21.56 313. 6230 237.0 337.5 1.86x104 258.6 1.55x106 5.29x105 3.956x104 2.5 Z?, 4.0x10 3.5x104 77.26 318. 6230 237.0 358.1 4.2x104 314.26 1.64x106 1,19xto6 4.808x104 1.8 YY 2.43x10 2.81x104 62.2 318. 6230 237.0 342.3 3.4x104 299.2 1.57x106 9.76x105 4.578x104 2.0 Allowable Yield Stress = 36 Ksi (5.184x106 psf)
TABLE 12B-2
SUMMARY
OF FACTOR OF SAFETY FOR TANK (Y-Z PLANE) f 1
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(November 1978) Amendment 72
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i FIGURE 12B-1 q HORIZONTAL
! . CONTROL MOTION j
, FUEL Cll STORAGE TANKS i
i 58
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DECONVOLUTED MOTION FUEL OIL STORAGE 1 ANKS C2
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Chapter 13 MODIFICATIONS
%s)
The preceding chapters have identified Seismic Class I structures, systems, and components including those required for a safe shutdown of the plant, and have presented evaluations of stresses induced by a postulated earth-quake on the Hosgri fault. In some cases, it was shown that overstress conditions could have occurred at the seismic levels analyzed. Modifica- 72 r
tions made to reduce such overstresses to within allowable limits are discussed in this chapter.
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w i (November 1978) 13-1 Amendment 72
13.1 MODIFICATIONa TO STRUCTURES O Structures are discussed in Chapter 4.
V Det ailed discussion of modifications will be found in Chapter 4 and include:
- 1. Turbine Building
- a. Buttresses and si added. i
- b. Strengthening of ruc'"> (columns , cross-bracing , i t ru s w. ) ,
- c. Checker plate addeo,
- d. Enlargement of turbir , cal gap,
- e. Installation of post , sning system in turbine pedestal .
- f. Added pilasters to some shear walls. 72 i
- g. Some shear walls thickened and rock anchors added. )
- 2. Containment
- a. Modification of annulus steel.
- b. Modification of manipulator crane.
- 3. Fuel Handling Bridge Crane Modified
- 4. Fuel Handling Building
- a. Minor structural steel modifications. ,
13.2 MODIFICATIONS 70 THE REACTOR COOLANT SYSTEM l
The reactor coolant systerr is discussed in Chapter 6.
All components and supports for the reactor coolant system satisfy criteria l demonstrating qualification for the Hosgri event without modification.
(November 1978) 13-2 Amendment 72 l
13.3 MODIFICATIONS TO MECHANICAL EQUIPMENT
. Evaluation of mechanical equipment is discussed in Chapter 7.
Modifications necessary to satisfy appropriate criteria for overstressed components are presented in this section.
13.3.1 AIR OPERATED VALVES Several valves have been identified which are necessary for or would be convenient in the shutdown of the plant as described in Section 5.1 and which are air operated. The four steam generator 10% atmospheric relief valves and the pressurizer auxiliary spray valve are required to operate.
A valve in the charging pump discharge line is required to remain open.
Two pressurizer power operated relief valves and a number of valves in the 72 letdown lines would.be convenient to operate. The plant compressed air system is not Design Class I. If the plant were to be shut down without off-site power, the air compressors would not be energized and the system gradually would lose pressure. To insure the reliable operation of the valves discussed above, modifications will be made to provide a separate backup supply of compressed gas to them from the plant nitrogen supply or from separate accumulators. To further insure operation, another separate 72 backup supply of compressed air operable from the control room will be installed for each of the steam generation 10% atmospheric relief valves and a manual by-pass valve will be installed around the pressurizer auxiliary spray valve. There is an existing manual by-pass valve around j the valve in the charging pump discharge line. Accumulators will be added j for valves 9356A and 9351 A and B so that samples may be obtained to confirm j boration of the reactor coolant system. !
i 13.3.2 EQUIPMENT REQUIRED FOR SHUTDOWN
- 1. The regenerative heat exchanger required a continuous weld on the shell/ saddle interface at the middle shell fixed support and the 72 support was reinforced. ,
1 O 13-3 Amendment 72 (November 1978)
- 2. The seal water injection filter baseplate was stiffened by welding additional plate. 72 O
- 3. The residual heat removal pump motor hold-down bolts are overstressed
- and will require replacement by bolts of higher capacity. A support bracket for the conduit box will be added to increase the fundamental frequency of the box to greater than 33Hz.
- 4. Residual heat removal heat exchanger bottom supports required an additional horizontal seismic restraint.
- 5. The component cooling water heat exchanger supports required reinforcement. Bracing frames were added at the fixed end support 72 both in the longitudinal and the lateral directions and brought the stresses of the support and the loads of the hold-down bolts within code allowables. The sliding support was modified in the lateral direction and guides were added to allow the heat exchanger to expand axially while the postulated seismic uplift forces of the hold-down bolts were kept low.
\
13.3.3 EQUIPMENT NOT REQUIRED FOR SHUTDOWN 1
- 1) The letdown heat exchangers required additional gussets on the saddle support. The baseplates have been extended and anchor bolts added.
72
- 2) The following demineralizers or ion exchangers required horizontal restraints at centers of gravity and that baseplates. be stiffened:
- a. The mixed bed demineralizers
- b. The cation bed demineralizers
- c. The boric acid evaporator feed ion exchangers
- d. The deborating demineralizers
- 3) The spray additive tanks required support modifications, the addition of baseplate gussets, and unslotting the baseplate holes. 72 O
(November 1978) 13-4. Amendment 72
- 4) The seal water heat exchanger required additional gussets on the supports.
13.3.4 VALVES
- 1) The bracket on the operator of Valve 8BA54D (HCV 637, 638, 670) is s
being replaced to increase the fundamental frequency of the valve assembly to greater then 33Hz.
- 2) All Limitorque operators on Class I valves have had hardware changes to prevent loosening of internal parts (of limit switches).
- 3) Main steam isolation valve operating cylinder supports are being reinforced.
- 4) Valves in the steam lines to the auxiliary feed pump drive turbine are being reinforced to reduce yoke stresses. 72 13.4 MODIFICATIONS T0 PIPING SYSTEMS
't piping systems other than the Reactor Coolant System are discussed in Chapter 8.
Approximately 900 piping hanger assemblies have been modified. An additional nearly 150 hangers were modified to upgrade them from Design Class 11 to Design Class I to support Design Class II piping required for long term cooling. Several piping connections have been added to facilitate the use of possible alternate long-term cooling water sources.
4 4
13.5 MODIFICATIONS TO ELECTRICAL EQUIPMENT Electrical equipment is discussed in Chapter 10. Modifications include:
Hold down clips were added to starter cells in 480 volt vital load center.
(November 1978) 13-5 Amendment 72
J Kickout springs in some motor controllers were replaced.
O Bracing was added to battery racks.
V Some raceway supports were revised.
1 Some relays were replaced.
4.16 kV Class IE switchgear was reinforced structurally. 72 Potential transformers were relocated.
Flexible connections were installed between switchgear and overhead bus ducts.
The saeguard relay boards were strengthened by welding at the base channels.
l 1 l
+
13.6 MODIFICATIONS TO OUTDOOR TANKS Outdoor tanks are discussed in Chapter 11.
l Modifications include:
f a. Foundations extensively modified, including enlargement, grouting, and rock anchors. 72 2
- b. Concrete jackets added.
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O O
(November 1978) 13-6 Amendment 72
~
AMENDMENT 73 INSTRUCTION SHEET (File this instruction sheet in the front of Volume 1 as a record of changes.)
The following instructions and check list are provided as a guide for the insertion of new pages for Amendment 73 in the FSAR for Units 1 and 2 Diablo .
Canyon Site. The new pages are marked " Amendment 73" and "(November 1978)"
and contain both amended and supplement:ry material. This material is indicated by a vertical bar with the figure "73" inscribed in the adjacent margin of the page. Where such marks appear adjacent to a blank portion of a page, a deletion is indicated. Where pages have been changed only to reposition material, with no change in content, only the amendment number and the date are given.
l
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For a brief description of the changes made by Amendment 73, see the l
SUMMARY
of AMENDMENT 73 which precedes the ". REMOVAL-INSERTION" INSTRUCTIONS.
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_1-
SUMMARY
OF AMENDMENT 73 l Location of Change Comment 2.1-3 Revises text to be consistent with Figure 2.1-2.
I Figure 8.4-1 Shows addition of redunlant air flow switch.
Figure 9.4-7 Shows addition of fire dampers.
9.5-11, 9.5-11a Provides update to reflect that diesel fuel oil 9.5-llb, 9.5-12, piping is now in a trench.
9.5-13, 9.5-13a, l
2 9.5-13b, 9.5-14, 9.5-15 Figure 9.5-6 Updates figure showing diesel fuel oil transfer pump vaults.
Figure 9.5-7 Shows pipe in trench.
I 13.1-1 through Updates discussion of conduct of operations and 13.1A-96 organizational structure. l Figure 13.1-4 Update of figure.
13.2-1 through Updates discussion of training program.
13.2-11 and table 1
13.2-1(sheets '
I through 7) 13.4-2, 13.4-2a and Updates discussion of review and audit committees.
Figure 13.4-1 (sheets 1 through 6)
Figure 13.4-2 (sheets 1 and 2) 1 1
REM') VAL-INSERTION INSTP.UCTIONS Remove Insert New Material O Volume 1 2.1-3/2.1-4 2.1-3/2.1-4 Volume X Figure 8.4-1/ blank Figure 8.4-1/ blank Volume XI Figure 9.4-7/ blank Figure 9.4-7/ blank 9.5-11/9.5-11a 9.5-11/9.5-11a 9.5-11b/9.5-12 9.511b/9.5-12 9.5-13/9.5-13a 9.5-13/9.5-13a i 9.5-13b/9.5-14 9.5-13b/9.5-14 9.5-15/9.5-16 9.5-15/9.5-16 Figure 9.5-6 Figure 9.5-6 Figure 9.5-7 Figure 9.5-7 O Volume XII 13.1-1/13.1-2 through 13.1-1/13.1-2 through 13.1-13/13.1-14
] 13.1-13/ blank Figure 13.1-3 Figure 13.1-3 i Figure 13.1-4 Figure 13.1-4 13.1A-4/13/1A-4 13.1A-3/13.1A-4 through 13.1A-95/13.1A-96 through 13.1A-65/
13.1A-66
) 13.2-1/13.2-2 through 13.2-1/13.2-2 through 13.2-13/ blank 13.2-10/13.2-11 Table 13.2-1 (Sheets 1 Table 13.2-1 (Sheets 1 through 7) through 7)
- Figure 13.2-1 Figure 13.2-1 13.4-1/13.4-2 13.4-1/13.4-2 1
13.4-2a/ blank Figure 13.4-1 (Sheets 1 Figure 13.4-1 (Sheets 1 through 6) through 6)
, Figure 13.4-2 (Sheets 1 Figure 13.4-2 (Sheets 1 and 2) and 2)
The offshore area is at times entered by commercial or sports fishing boats.
In the event of emergency, such boats could be notified and requested to O 1 eave the area. As described in Section 13.3, arrangements have been made Q
1 with appropriate authorities for notification and removal of such boats and of persons occupying the breakwaters, offshore rocks, or shoreline in the event of emergency.
Boundaries for Establishing Effluent Release Limits On land, the boundary line of the restricted area (as defined in 10 CFR 20) coincides with the exclusion area boundary. The site boundary coincides with these two boundaries except at the southeastern section as shown in 73 Figure 2.1-2. Control of access to the land area within this boundary will be as described for e>talusion area control. As therein described, no special provisions have been made for control of access, during normal operation, to the offshore area below the mean high water line. Occupancy 1
of this area by any membe. of the public is expected to be of short duration, I resulting in exposures during normal operation within the limits established n by 10 CFR 20.106(b) and related " low as practicable" provisions.
( )
%.J 2.1.3 POPULATION AND POPULATION DISTRIBUTION Population data are based on th 1970 Census and on projections prepared by the State of California Department of Finance. The portion of California which lies within 50 miles of the site is relatively sparsely populated, j having approximately 210,000 residents in 1970. A circle with a 50-mile l radius includes most of San Luis Obispo County, about one-third of Santa Barbara County, and a minor, sparsely-populated portion of Monterey County.
The 1970 Census population of this region is less than had been projected in the PSAR, and the subsequent projections by the Department of Finance are also reduced.
P (November 1978) 2.1-3 Amendment 73
About 55 percent of the area within the 50-mile circle is on land; the balance falling in the Pacific Ocean. Table 2.1-1 shows population trends for San Luis Obispo and Santa Barbara counties and for the State of California.
Table 2.1-2 shows the growth since 1960 of the principal cities within 50 miles of the site. Table 2.1-3 lists all communities within 50 miles having a population of 1,000 or more, giving distance and direction from the site and population in 1970.
Population Within 10 Miles Approximately 6,260 persons reside within 10 miles of the site. The nearest residence is 1-3/4 miles north northwest of the site. Two persons occupy this dwelling. Figure 2.1-4 shows the 1970 population distribution within a 10-mile radius. This area is divided into 22-1/2 sectors, with circles
, of radius 1, 2, 3, 4, 5, and 10 miles. Figures 2.1-5, -6, -7, and -8 show projected population distributions for 1980, 1990, 2000, and 2010, respec-tively. The distributions are based on the assumption that land usage will not change in character during the next 40 years, and that population growth within 10 miles will be proportional to growth in San Luis Obispo County as a whole. No major plans exist that would affect this assumption. Plans for a new community in the Los Osos Valley that was to have been financed under Title 4 of the 1968 Department of Housing and Urban Development Act (and referred to in the Unit 2 PSAR) did not materialize.
Other potential residential additions within the 10-mile radius are as follows:
- 1. Baywood Park (7-9 miles N). There are some 1,500 separately owned undeveloped parcels of land, each consisting of 2, 3, 4, and sometimes more lots, that can be developed by the owners at any time.
- 2. Los Osca (6-7 miles N). This area has some 3,000 undeveloped lots, ranging from small to 5 to 10 acres in size. The latter are zoned so that apart-ments can be built, or they can be abdivided into smaller lote, (In the Baywood Park-Los Osos area, during the 39-month period f rom A1 il 1970 through July 1973, 767 dwelling units were added.
2.1-4
O O O LEGEND g amper D opens when exhaust fan E-27 io shutoff by thermostat.
Outside Air Intake
.]
7 r ( S-27) -
f sE-27x' r Exhaust Air Discharge j__ _/w r N t7I f
W i with Damper Supply Air
-El. 163'-4"
'-Air Temperature Thermostat -+ Exhaust Air
" in Exhaust Duct. See Note.
u CkJ Fan El. 140'-0" Roughing Filters h
y y\p Fire Damper O' Air Flow Switch
-El . 128'-0"
(
{
- 4} ~ + ,_ + m- + . ,y [ NOTE Thermostat controls exhaust fan Y 500 cfmO [ 500 cfm 6 500 cf g E-27 and supply fan S-27. Supply r
h 500 cfm 500"cfm 500*cfm' an - n continuously, e M er BATTERY ROOM BATTERY ROOM BATTERY ROOM # " "P' ' "" * ""* #"
E El* 115'-0" E-27 off on low temperature.
l - -From Switchgear Areas To Switchgear Area a e UNITS I AND 2 El. 100'-0" DIABLO CANYON SITE
'. , .lo k' FIGURE 8.4-1 BATTERY ROOMS VENTILATION SYSTEM AREA-H PARTIAL AUXILIARY BUILDING AUXILIARY BUILDING Amendment 73 November 1978
_.m._ _ _ . . _ . . . . . _ _ . _ , _ . . . . - _ - . _ _ _ . . _ . _ ___ _ ..m_ . .. . _ _ _ _ _ __ _ _ . __ _
O O O- >
l s
i EXHAUST STACK WITH WIRE ,
MESH COVER i LEGEND 5
d GW g
f
- SUPPLY AIR DUCT
< TURBINE BUILDING OPERATING FIDOR
/
El. 140 OUTSIDE AIR INTAKE LOUVER l l <l 1 SUPPLY FAN
/ 6 A SUPPLY AIR OUTLET
( S-69 ) I
- SUPPLY FAN ' " ^
6,000 CFM ON VITAL 4-i BUS "F" r {, FLOOR GRATING
~ El. 119 h4 ROUGH UG FIL'IER 6 #-+
' FLOOR GRATING FIRE DAMPER i
h i I
O/A + CABLE SPREADING ROOM FOR VITAL BUS ABOVE s .a
! El. 107 i
l
% - A r -Ar %
- Jc'g ,
r l 4 Kv v1TAL BUS "F" UNITS I AND 2 I i 4 KV BUSES "G" AND "H" ARE SIMILAR DI ABLO CANYON SITE FIGURE 9.4-7 VETIILATION SYSTEM 4 KV SWITCHGEAR ROOMS i
Amendment 73 November 1978
- 2. Two diesel fuel oil transfer pumps located below ground level, each adjacent to a storage tank but in separate compartments. Each transfer pump delivers more than 55 gpm at a discharge pressure of 50 psig, pumps are of the positive displacement rotary screw type with 5 hp motors.
One pump is more then adequate to supply the five diesel generators of Units 1 and 2 running at rated load. A cartridge type fuel oil filter is located at the discharge of the fuel oil transfer pumps to prevent any fuel oil contamination from reaching the engine base mounted diesel fuel oil tanks. The diagram of the engine fuel oil transfer system is shown in Figure 3.2-21 (Sheet 1). The physical arrangement of the fuel oil transfer system is shown in Figures 9.5-6 and 9.5-7.
- 3. Two diesel fuel oil supply headers to each unit routed in separate trenches.
- 4. A diesel fuel oil tank (day tank) of 550-gallon capacity built into the base of each diesel engine generator. This is sufficient capacity for two and one-half hours of full load operation. N
? 5. Two redundant supply headers between the fuel oil storage tanks an,d the base-mounted day tanks.
- 6. A direct engine driven positive displacement fuel cil booster pump delivering fuel oil to the engine injection system, taking suction from the base mounted tank and passing the oil through two duplex absorbent type filters. The diagram of the engine fuel oil system is shown in Figure 3.2-21 (Sheet 5). The physical arrangement of the engine generator units and the engine fuel oil system is shown in Figures 9.5-8, 9.5-9, and 9.5-10.
N (November 1978) 9.5-11 Amendment 73
System instrumentation and control is provided on the tanks and pumps as v
follows :
- 1. The base-mounted day tanks have two separate redundant transfer pump start-stop level switches. Each level switch starts a transfer pump and opens the supply header solenoid valve corresponding to the respective transfer pump, "A" or "B." The start setting for the header "A" level switches is slightly different then those for header "B," allowing one to be a backup.
- 2. The start of transfer pump "A" or "B" is indicated both locally and in the control room.
- 3. Local controls at each diesel generator and manual crosstie valving between headers allow manual starting of either transfer pump and filling of the base mounted tanks from either header system, "A" or "B."
D U
O (November 1978) 9.5-11b Amendment 73
- 4. High and low level alarm switches are installed on all base mounted l
tanks, which activate alarms both locally and in the control room to alert the operators.
- 5. Illgh and low level alarm switches are installed on both fuel oil storage :
tanks, which will activate alarms in the control room. Additionally, dipstick type indicators are provided on each storage tank.
Safety Evaluation {
1 The two diesel fuel oil storage tanks are buried, and are designed to the criteria of Part 2, Bulk Underground Storage, of NFPA No. 30, " Standard for the Storage, Handling and Use of Flammable Liquids." The National Fire Code j does not require any specific fire protection system for this type of under-ground storage tank. Physical separation of the two tanks precludes the possibility of a fire in one storage tank from spreading to another tank.
Fire risk is further minimized by the fact that the only source of oxygen to support a fire is through the 4-inch tank vent. If sufficient heat were generated to damage and collapse the storage tank, dirt would cave in and help smother the fire. Two yard hose reel stations are available for fire fighting any above ground fires at the tank location. The exterior surfaces of the tanks have a corrosion-resistant coating of coal-tar epoxy. Cathodic protection is provided to protect the tanks and underground piping from stray 73 electric ground currents. Steel piping is wrapped in accordance with AWWA C203.
The Diesel Generator Fuel Oil System is designed to remain operable after sustaining a single failure of either an active or a passive component. The capability to meet the single failure criterion is met by providing redundancy in tanks, pumps, valves, piping, and power supplies. The system arrangement provides sufficient separation of the tanks and their associated transfer pumps that the possibility of damage to both simultaneously as a result of a single event is considered highly unlikely. The f uel oil transfer piping, transfer pump and tank manifolding are arranged so a single failure of any pipe, valve, tank or pump will not disable the system. Safety of the reactor Amendment 73 9.5-12 November 1978
facilities is not impaired by the sharing of the fuel oil systems as any combination of one storage tank and one pump is capable of serving all five day tanks. Each unit normally supplies power to one transfer pump from one vital 480 volt bus. Provisions are made to manually switch both pumps to the vital buses of either unit, in case one unit should be placed in a prolonged cold shutdown condition.
To assure good fire protection practice, the diesel engine generator instal-lation was designed in accordance with National Fire Protection Association (NFPA) Standard No. 37, Standard for the Installation and Use of Stationary Combustion Engines and Gas Turbines. This standard permits a maximum capacity of 550 gallons for an integral fuel oil day tank, mounted on the engine base.
This capacity is adequate for normal engine operation. It also provides suf-ficient tima for all necessary operator actions to assure that diesel generator operation is not interrupted in the event of any malfunctions in the system which transfers fuel oil from the underground storage tanks to the day tanks.
As discussed in the preceding paragraph, the entire Diesel Generator Fuel Oil System is redundant up to and including fill valves and connections on the engine day tanks, so that a single malfunction will not prevent the transfer of oil. In the unlikely event of malfunctions in both redundant fuel oil headers, such as a pump failure in one and a piping blockage in the other, low level will be alarmed when sufficient fuel oil remains in the base-mounted day tank for one hour's operation of the engine at full load. One hour is ade-quate for an operator (1) to correct a malfunction on one of the two redundant transfer headers or (2) to line up manually the valves of the two headers into one path which will transfer oil. All the valves necessary for this action are readily accessible in the compartments for the diesel fuel oil transfer pumps.
The design considerations to prevent water from flooding or ground water from entering the fuel oil storage tanks, concrete vaults, and pipe trenches were: 73 As discussed in Section 2.4, the risk of surface water flooding at this site is essentially zero. No ground water has been encountered at or below the buried tanks, pump vaults, or pipe trenches. Therefore, the source potential pd for water flooding the fuel oil system is negligible.
74 (November 1978) 9.5-13 Amendment 73
In addition the below ground system is completely sealed with the vent extended 16 feet above ground.
1 i
The two transfer pumps are in separate, underground, reinforced concrete I l
vaults with solid covers elevated above the general ground level such that surface water will flow around them. These vaults are drained to the building sump and are protected with backwater valves.
The two reduadant fuel oil supply hearlers are in separate, below-ground reinforced concrete pipe trenches with solid covers elevated above general I ground level. These trenches are drained to sumps well below the trench elevation.
73 Diesel fuel oil piping within the trenches is classified as Seismic Class I.
Supports for the diesel fuel oil piping were designed using the methods described in Section 8.4, in Chapter 8 of the Hosgri Seismic Evaluation Repoct.
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(November 1978) 9.5-13b Amendment 73
The diesel fuel supply headers are hydrostatically tested during construction and all active system components, pumps, valves, and controls are functionally tested during startup and periodically thereafter. The diesel fuel oil in storage will be periodically tested for any possible contamination or )
deterioration.
Experience with the Company's transmission system indicates that in the event of complete loss of off-site power, restoration of normal power sources could ,
be accomplished within a few hours. However, seven days of on-site power generation has been used as a conservative upper limit for design and safety evaluations of fuel storage capacity for the tanks, even though it is highly improbable that the diesel generators would be required to furnish plant aux- )
iliary power for this long a period. The plant diesel fuel storage supply could be replenished within this time for the following reasons:
l l
- 1. The possibility that delivery trucks from fuel oil suppliers could not reach the plant under any weather conditions is very remote, since hard-surfaced all-year roads exist between the plant and the suppliers.
- 2. If the roads should become impassable, an agricultural spray firm in the area has four helicopters that could readily transport diesel fuel.
In addition, boats or barges could be used to supply the plant in such an emergency.
There are six major petroleum distributors within forty miles of the plant.
Their normal diesel fuel stocks on hand and their delivery capabilities are shown in Table 9.5-3.
Corrosion Control - Compliance with NACE Recommended Practice RP-01-69 Insulation for the diesel fuel oil lines begins immediately inside the west wall of the Turbine Building. The vent lines have insulating flanges where they enter the building. All diesel fuel oil lines entering or leaving the 73 transfer pump vaults are contained in trenches to protect grounded pumps and accessory equipment from the cathodic protection system.
g Amendment 73 9.5-14 (November 1978)
/9 Both storage tanks and all buried piping have protective coatings. The tanks are coated with al-tar epoxy to a minimum dry-film thickness of 16 mils. They were checked for faults with a holiday detector and all coating defects were corrected. The steel piping is coated with a coal-tar enamel in accordance with AWWA C-203.
Field tests were made at the site to (1) assure the integrity of the electrical isolation, (2) ascertain protective current requirements, and (3) determine the most suitable means of providing cathodic protection current (galvanic anodes or impressed current system).
The system was found to be electrically isolated.
A temporary d-c power source was used to polarize the diesel fuel oil tanks and their associated piping to a negative voltage of at least 0.85 volts, as measured between the structure surface and a saturated copper-copper-sulfate half-cell contacting the adjacent soil. The required shift in potential was i,x_,) acquired with 44 mi111 amperes.
Cathodic protection is provided by a galvanic anode system which was chosen because of the low current requirements, low-to-average soil resistivities, congested area surrounding tank and piping, and ease of maintenan;e.
Three magnesium anodes are installed at a depth of eight feet and are back-filled with a mixture of gypsum, bentonite, and sodium sulfate. The anodes are located so as to provide good distribution of current with no shielding effects. Each anode is installed through an open-Lottom concrete vault to easy access for replacement when necessary.
f3 N)
(November 1978) 9.5-15 Amendment 73
i To provide for periodic monitoring of the anode array, all anodes are fastened to a common header cable (AWG #12, Type TW) . The cable is soldered to a 2-ohm, 10-watt, wire-wound ceramic shunt located in one of the pump vaults. The j other end of the shunt is connected to the steel piping of the fuel oil system.
An indicating structure-to-soil voltmeter is located in the same vault to allow seriodic checking of the protection level of the system structures.
The inoicating meter employs a five-pound zine reference electrode, buried at a ten-foot depth adjacent to one of the storage tanks.
l l
9.5.5 DIESEL GENERATOR AUXILIARY SYSTEMS l
The diesel generator starting system, cooling system, lubricating oil system l and controls and instrumentation are described in Subsection 8.3.2. These systems are shown schematically in Figure 3.2-21. Their physical arrangement is shown in Figures 9.5-8, 9.5-9 and 9.5-10. i l
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- d. M ER9LILE O' CPUMM UNITS I AND 2 DIABLO CANYON SITE FIGURE 9.5-7 DIESEL-GENERATOR FUEL OIL PIPING r Amendment 73 November 1978 .
_ ,. _ _ _ _ _ _ . . _ - - - - - _ - - - - - - - - - - - - - - - - - - - ' - - -- -~ ^-
I l
13.0 CONDUCT OF OPERATIONS The purpose of this chapter is to describe the manner in which the Company intends to operate and maintain the plant.
I The Company has been involved in the operation of nuclear power plants since I
- 1957. An approach to the operation of nuclear plants has been developed j during this period which is compatible with the Company's organizational (
concepts and operational philosophy which has been successfully employed for l many years in its conventional thermal power plants. The Nuclear Regulatory j Commission has had the opportunity to observe this operational approach at the Company's Humboldt Bay Nuclear Unit which commenced operation in 1963. The successful operation of this Unit demonstrates that the Company has a technically competent and safety-oriented operating organization. '*l
~
1 13.1 ORGANIZATIONAL STRUCTURE 4
13.1.1 CORPORATE ORGANIZATION The structure of the Pacific Gas and Electric Company's corporate organization is shown in Figure 13.1-1. The manner in which the various Company departments
- function in carrying out the design, construction, quality assurance, and initial startup and testing of the Diablo Canyon Units 1 and 2 are described in Chapters 14 and 17. The working interrelationships and organizational interfaces I
between the Company and Westinghouse (the NSSS manufacturer) and other suppliers and contractors are described in Chapter 1.
This section of Chapter 13 describes the Department of Steam Generation, which
, is the General Office department that supports the operation of the Company's nuclear power plants. Thr Department of Steam Generation is beaued by the Manager of Steam Generation. The Manager of Steam Generation reports to the Vice President-Electric Operations, who in turn reports to the Executive Vice President (Operations).
O (November 1978) 13.1-1 Amendment 73
The basic function of the Departnent of Steam Generation is to provide functional direction, guidance and assistance on operation and maintenance of the Company's fossil fuel, geothermal, and nuclear power plants. The depart-ment coordinates operation, maintenance, and performance of these facilities 4
as an integrated system. Other departmental functions include:
- 1. Developing policies, procedures, and standards for activities under the department's functional jurisdiction; communicating these policies, l procedures, and standards to other General Office departments and to the Division operating organizations; and auditing for compliance with these
- policies, procedures, and standards.
- 2. Advising and assisting Division operating organizations in performing activities under the department's functional jurisdiction, including:
- a. Construction, operation, and maintenance of facilities.
- b. Preparation and approval of cost estimates to perform work,
- c. Organization and personnel changes including procuremenu, training and utilization of engineers and supervisors and long-range organizational planning.
- d. Selecting and obtaining materials, tools and equipment for work performed.
- e. Preparation and administration of licenses, permits, agreements or contracts.
R
- f. Investigation of complaints or claims involving facilities.
- 3. Working with other General Office departments on projects in which the department has a functional interest, including:
- a. Planning, designing, constructing and testing of facilities.
O Amendnent 73 13.1-2 (November 1978)
1
- b. Developing specifications for materials, tools, and equipment.
Disposing of obsolete materials, tools, and equipment.
c.
- d. Developing' safety and training programs, job definitions, and lines of progression for personnel.
- e. Investigating complaints or claims involving facilities.
- f. Analyzing proposed governmental regulations and legislation which may affect facilities.
- 4. Providing advice and assistance to the Division operating organizations in bringing new plants into operation.
- 5. Operating a central chemical icboratory for the thermal power plants.
- 6. Conducting investigations and performing research on metallurgical, water chemistry, and air and water pollution problems relating to thermal power production.
- 7. Performing studies of the cost of operating and maintaining facilities under the department's functional jurisdiction, and making recommendations for improvements.
- 8. Maintaining records, diagrams, maps, specifications, and test data or other material on facilities.under functional jurisdiction for Company use or preparation of reports to governmental agencies.
Figure'13.1-2 presents a current organization chart of the Steam Generation Department which shows the relationship of the General Office Department of i Steam Generation to the Division operating organizations. As indicated in this figure, three of the Company's geographic divisions have two or more thermal power plants. . In these divisions, administrative coordination of the various_ power plants is provided by a Division Steam Superintendent and his O -
-(November'1978) 13.1-3 Amendment 73 r 1-- - 1m:w - 9 e no. -e -
w+e
staff. The Coast Valleys Division operating organization consists of the Division Steam Superintender.t and his staff, and the plant operating organizations for the Moss Landing Power Plant (2060 MWe net), Morro Bay Power Plant (1002 MWe net), and Dia'lo o Canyon Units 1 and 2 (2205 MWe net).
The Manager of Steam Generation is in charge of the General Office Department of Steam Generation. The Manager of Steam Generation has reporting to him an administrative and clerical staff and a technical staf f. This technical staff provides support for the operation of the Company's thermal power plants using I
the resources of the Department of Steam Generation and technical specialists from a variety of other Company departments and consultants. One of the Supervising Steam Generation Engineers reporting to the Manager of Steam Generation functions as the " Engineer in Charge" of the technical support activities for nuclear power plants. The qualifications of the incumbent in this position substantially exceed the minimum requirements of Section 4.6.1-(Engineer in Charge) of ANSI /AN3 3.H978 (Revision of N18.1-1971) " Selection and Training of Nuclear Power Plant Personnel."* Referring to Figure 13.1-1, the principal Company departments which assist in the technical support of nuclear power plant operations are Law, Claims and Safety (industrial and radiological safety and fire protec; ion), System Protection (protective relaying), Communications, Gas Control (meteorology), Automotive and Equipment (internal combustion engines for emergency power systems), Security, Engineering Research (environmental radiological enitoring, terrestrial and aquatic biology, metallurgical engineering, ncadestructive testing, and in-service inspection of components), Siting (ge ilogy and seismology), Computer Applications, Civil Engineering, Electric Distric 4 tion Engineering (in-plant electrical systems),
Electrical Generation and Transmission Engineering, Mechanical and Nuclear Engineering, Engineering Services (inspection and quality control in vendor shops), General Construction, Materials (purchasing), Quality Assurance, and Environmental Quality.
- "4.6.1 Engineer in Charge - The engineer in charge shall have a minimum of a Bachelor s Degree in Engineering or the Physical Sciences and have a min-imum of three years of professional level experience in nuclear services, nuclear plant operation, or nuclear engineering, and the necessary overall nuclear background to determine s hen to call consultants and contractors for dealing with complex problems beyond the scope of owner-organization expertise."
O Amendment 73 13.1-4 (November 1978)
l
~ The Company's overall technical competence is such that most technical support functions for nuclear power plants will be handled within the Company ,
organization. This technical competence results in part from the fact that the Company has performed the architect-engineering design and construction manage-ment of Diablo Canyon Units 1 and 2 and has successfully operated the Humboldt Bqy Nuclear Unit since 1963. l There are presently no plans to assign responsibility for any specific area of technical support to an outside consultant. As required, outside consultants will be used to assist .on special problems. On the basis of the operating experience at Humboldt Bay, it is anticipated that this consultant support will principally involve assistance in certain areas of fuel management and safety analysis by Westinghouse. In every case, consultant personnel will work with Company specialists in each of these areas.
The Manager of Steam Generation and his technical staff are senior technical people with extensive experience in all aspects of power generation technology.
Commencing in 1957 when the department first became involved in nuclear power plant operations, additions of new personnel and training for existing O personnel has been carried out to provide the required technical support capability in the nuclear power field. This program will continue as addi-tional nuclear generation is added to the Company's system. The educational backgrounds of the technical staff include mechanical, electrical, chemical and nuclear engineering and chemistry and physics. Most of these individuals have work experience in nuclear facilities. Several individuals have received one or more NRC operator licenses for Company nuclear power plants.
The department collectively has the capability required to provide operational technical supptet and/or coordinate such technical support with specialists in other Company departments and consultant organizations in the following areas:
- 1. Nu' clear power plant operations, including quality assurance aspects of I
operation, maintenance, refueling, repairs and modification of nuclear power plants.
- 2. Nuclear engineering,' including nuclear fuel management and safety analyses.
l (November 1978) 13.1-5 . Amendment 73 e -- b-r - - , , < . , , . , , , - , ,_ g _, ,,,
1
- 3. Chemistry and radiochemistry.
- 4. Metallurgy.
- 5. Instrumentation and control, including process computers.
- 6. Radiological safety and radiological environmental monitoring.
l l
- 7. Mechanical, electrical and chemical engineering. I
- 8. Training of technical, technician, operations and maintenance personnel.
In summary, the General Office Department of Steam Generation, with assistance from other Company departments-and outside consultants as required, provides technical support of the operation of the Company's nuclear power plants in a manner fully consistent with the requirements of NRC Regulatory Guide No.1.8, Revision 1-R September 1975, re-issued May 1977.
O O
Amendment 73 13.1-6 (November 1978)
, 13.1.2 OPERATING ORGANIZATION
- The structure, functions, and responsibilities of the Company's operating organization for Diablo Canyon Units 1 and 2 are described in this section of Chapter 13.
- Organizational Arrangement Figure 13.1-4 illustrates the plant organization for the operation of the two Unit plant. This organization chart is in accordance with the Company's organizational practices for staffing conventional generating plants, with increased emphasis on the technical functions required in the operation of a nuclear unit. The organization is generally similar to the Humboldt Bay organization.
The organization for the startup and initial operation of Unit 1 is similar to that for operation of the two units except that the number of operations personnel per shif t will be less. The numbers applicable to the phase of operation are given in the legend on the figure.
V The plant organizational requirements during startup of Unit 2 with Unit 1 in operation are basically similar to those for operation of the two units.
The functional positions on Figure 13.1-4 will be filled at the time of initial loading of Unit 1, except that in the case of physical force maintenance personnel the Company will fill these positions as required based upon the work load and the availability of qualified Company General Construction personnel ar,d traveling maintenance crews from other Company power plants.
It is recognized that additional supervisorial and physical force personnel will be required to augment the normal crews during certain operating periods such as refueling, and cold and hot startups. Company practice at the Humboldt Bay Nuclear Unit and at conventional units is to extend the normal working day, and/or to require personnel to work on days they would normally have off. In addition, maintenance personnel from the Company's other thermal plants will be used as required for major repairs.
\_/
(November 1978) 13.1-7 Amendment 73
Finally, the Company will provide additional persons in certain positions shown on the organization charts in order to provide for replacement of '
personnel when required and to build up a pool of trained individuals who can be utilized in the staffing of new plants or in other functions. This practice is evidenced by Appendix A to this Chapter, which identifies the personnel currently holding various supervisorial positions in the plant operating organization. It should be noted that additional supervisory personnel have l been provided in nearly every area.
l Functions, Responsibility, and Authority The functions, responsibilities and authorities of key supervisorial and technical positions in the Diablo Canyon operating organization are summarized briefly in the following paragraphs. Detailed job descriptions for these positions have been prepared in accordance with the requirements of ANSI N18.7- l 1976, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants." Each individual in these positions is provided with a current written copy of his job description and is responsible for familiarity with its provisions.
1 Plant Superintendent Overall responsibility for all on-site activities in connection with the safe and efficient operation, maintenance, and administration
- of the plant.
The Plant Superintendent will be assisted by the Supervisor of Maintenance, Supervisor of Operations, and Power Plant Engineer in carrying out these responsibilities. One of these individuals will be selected in advance to act as Plant Superintendent in his absence.
"Alfministration" as used in these summaries applied to all facets of administrative control, including personnel management, personnel training, review and audit, security, and quality control.
O Amendment 73 13.1-8 (November 1978)
Supervisor of Maintenance 4 Responsible for administrating, coordinating, planning and scheduling all mechanical and electrical maintenance activities at the plant. He provides
- direction, assistance and guidance to the mechanical and electrical maintenance supervisors in preventive and corrective maintenance techniques and programs.
Mechanical Maintenance and Electrical Maintenance Supervision R.sponsible for direct supervision of the mechanical and electrical maintenance activities respectively, and for providing administrative support in these areas. I Supervisor of Operations Responsible for administrating, coordinating, planning, and scheduling all operating activities at the plant. He is the responsible supervisor for assur- I ing that appropriate operating procedures are available and that operating personnel are familiar with the procedures. In carrying out these responsibilities
, he provides direct supervision to the Relief Shif t Supervisor, the Shif t 1
Foremen and the plant's training staff.
4 Relief Shift Supervisor Provides administrative support, guidance and coordination for the routine operation of all power production facilities in the plant and provides staff assistance to the Supervisor of Operations.
Shift Foreman Responsible for providing direct supervision of the plant operators and the work which they perform, and for providing administrative support in this area.
In the absence of higher supervision, the Shift Foreman is in full charge of the plant and has the authority to take any action which he deems necessary in the event of an operating emergency. It the Shif t Foreman on duty becomes incapacitated, the Senior Control Operator on duty will assume the dutier of the Shift Foreman. A replacenent Shif t Foreman will be provided as soon as prac tical .
(November 1978) 13.1-9 Amendment 73 J
Training Staff P
Responsible for assisting the Supervisor of Operations in the conduct of the NRC licensed operator training and retraining programs. Also assists in the conduct of routine training programs for all plant personnel as necessary. '
Power Plant Engineer i 1
1 Responsible for administrating, coordinating, planning and scheduling all l technical activities at the plant. He provides direct supervision to members of the plant technical staff in the areas of nuclear engineering, control and instrumentation, and chemistry and radiation protection. '
1 Nuclear Engineers j Responsible for providing administrative and technical support in the areas of reactor operations, including the analyses of core nuclear, thermal and hydraulic performance, management of the fuel cycle, and continuing evaluation of the operational nuclear safety matters.
Instrument and Controls Supervision Responsible for the maintenance, calibration and testing of plant instrument and control systems and administrative support for these activities. Provides direct supervision of the Control Technicians and Instrument Repairmen performing this work.
Chemical and_ Radiation Protection Supervision Responsible for providing administrative and technical support in the chemical, radiochemical, chemical engineering, and radiation protection aspects of plant operation. Provides direct supervision of the Radiation and Process Monitors performing chemical and radiation monitoring work.
O Amendment 73 13.1-10 (November 1978)
Quality Control Supervisor
() Responsible for coordinating and supervises the plant's quality control program, including the inspection program for safety related activities and records management at the plant. Provides direct supervision of the Quality Control Inspectors.
Quality Control Inspector.
] Responsible for planning and conducting inspection and surveillance of safety related activities at the plant.
Security Supervision Responsible for development, planning and coordination of all activities associated with the plant security program. Provides direct supervision of the plant guard force.
Assistant to the Plant Superintendent a
O Provides general assistance to the Plant Superintendent in the development of f
technical programs and procedures required to implement applicable regulatory requirements; planning and scheduling of outages; and in various administrative activities where engineering education and experience are necessary skills.
Provides interface, as required, between the Plant Superintendent and various regulatory and governmental agencies on licensing and other technical matters.
Staff Engineer Provides general assistance to the Assistant to the Plant Superintendent.
Finally it should be noted that the detailed job description of each supervisor gives him the authority to take -action in an emergency to the extent justified.
O J (November 1978) 13.1-11 Amendment 73
Shif t Operating Organization
, A normal six man shif t organization will be utilized for the initial startup and normal operation of Unit 1. This organization will be expanded to eight men for normal operation of both units. These organizations are shown on Figures 13.1-4. The establishment of these shift organizations is based upon consideration of the Company's power plant staffing philosophy, our evaluation of the operating practices of U.S. pressurized water reactor plants, and our experience with the Humboldt Bay Nuclear Unit and large (up to 750 MWe) fossil fuel units.
During preoperational and initial startup of Unit 1, each shif t will include six men; one Shift Foreman with a Senior Operator's License (SOL), one Senior Control Operator with an Operator's License (L), one Control Operator (L), one Assistant Control Operator (L), and two Auxiliary Operators who are not
, required to be licensed.
Prior to the initial startup and operation of Unit 2, in conjunction with the operation of Unit 1 (i.e., two unit operation), an additional Control Operator (L) and an additional Auxiliary Operator will be assigned to each shif t, for a total of eight men on each shift. For two unit operation, the Senior Control Operators will also hold Senior Operators' Licenses.
l The Control Operator (s) is assigned to a specific unit in the main control room. The Assistant Control Operator is normally stationed in the control room and is available to assist on either unit. The Senior Control Operator generally assists the Shif t Foreman and also directs the activities of the Auxiliary Operators. The Shift Foreman's office is adjacent to the control room and he is normally available to direct activities in the control room.
The eight man shift organization (two units) provides adequate manpower to cover the operating contingencies which can reasonably be expected to occur during normal operation of the plant. An organization of this size is also ddequate to monitor operation of engineered safety systems in the event of any of the design basis accidents. In our opinion, this number of personnel is diso suf ficient to perform those operations required to implement the required portions of the Emergency Plan.
Amendment 73 13.1-12 (November 1978)
All licensed operators will be qualified as radiation protection monitors as part of their operator license training. This means that they will be ade-quately instructed in the performance of radiation monitoring and other radia-tion protection functions normally encountered during both normal and nonroutine operations. Should additional assistance be required in the radiation protec-tion area, operating personnel are instructed to contact radiation protection supervision for assistance.
As previously discussed under Organizational Arrangement in this section, the i I
normal plant staff will be augmented with off-duty personnel or personnel from
- other Company plants during operations when they are required. l 13.1.3 QUALIFICATION REQUIREMENTS FOR PLANT PERSONNEL Standards !
I The Company is currently using ANSI /ANS-3.li978, " Standard for Selection and i
Training of Personnel for Nuclear Power Plants" as the basis for establishing l c minimum qualifications for all management, supervisorial and professional- l technical personnel in the plant organization.
The minimum qualification processes for physical force personnel (operators, technicians, monitors and repairmen) will exceed the qualifications for these positions stated in ANSI /ANS-3.1-1978 due to established Company training and apprenticeship programs. These programs include documented academic and on-the-job training plus comprehensive qualification examinations. The length l of these programs varies from a minimum of 24 months to a maximum of 48 months with most requiring 30 months. There is no formal apprenticeship program for the Radiation and Process Monitors. An applicant must pass a qualification examination which assures that he has the required background before he will be considered for the job. He receives academic and on-the-job training under the Chemical and Radiation Protection Engineer for a two-year period before he is considered fully responsible in this position.
(November 1978) 13.1-13 Amendment 73
Experience The key management, supervisorial and technical positions in the plant organization will initially be filled by individuals who have been actively engaged in the nuclear power field, including assignments on the plant staff at the Company's Humboldt Bay Nuclear Unit. The experience of senior personnel, at the time of Unit 1 startup will generally be in the range of 15 to 33 years. These personnel include the Plant Superintendent, Supervisor of Operations, Supervisor of Maintenance, Power Plant Engineer and the Relief Shift Supervisor. In addition, most of the Shift Foremen and at least one individual in each of the major technical disciplines (nuclear enoineering, chemistry and radiation protection, instrumentation and controls) will have had at least five years of similar experience. Most of these key personnel have had senior operators' licenses or operators' licenses on the Humboldt Bay reactor and/or the Vallecitos Boiling Water Reactor.
Resumes for personnel holding the key positions in the initial plant organization are included as Appendix A to this chapter.
l l
O Amendment 73 13.1-14 (November 1978)
i
)
l O
l Figure 13.1-3 has been deleted
, (November 1978) Amendment 73
~ -- - - . . - - _ . . _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ __ _ __ _, _
i
/ PLANT S I
l MAINTENANCE dPERATIONS CLERICAL QUALITY C SUPERVISOR OF SUPERVISOR OF MAINTENANCE OPERATIONS 1 1 SOL RELIEF SHIFT QUALITYh SUPERVISOR y SUPERV:
SOL I
l l ELECTRICAL MEACKANICAL TRAININC SHIFT QUALI MAINTENANCE MAINTENANCE STAFF FORE EN CONTBs 5
SUPERVISION SUPEP. VISION y 1 (1) LNSPE@(
. SOL
- SQL
~
5 j SR CONTROL CLERICAL 2I 2 ELECTRICIANS -
MACHINISTS STAFF 7
(1) OPERATORS 5 4
4 3 l
ELECTRICIAN _
CERTIFIED CONTROL 3 HELPER 1 WELDERS 2 (2) OPERATORS 8
l
- RICGER ASST. CONTROL MATERIALS 1 II)
OPERATORS 0
FACILITY FEKSONNEL 1
l
- TOOL CLERK AUXILIARY 1 (3) OPERATORS 13 MECH. MAINT. 38**
I EELPERS l 5 (8 MEN PER SHIIT) 25 trcggp:
SOL - NRC SENIOR OPERATOR LICENSE REQUIRED L - NRC OPERATOR LICENSE REQUIRED
- - OPTIONAL
- - The numbers shown are for the operation of two units For the startup and initial operation of Unit 1, the total complement and men per shift are as follows:
SF-5(1); SCO-5(1); Co-4(1); ACO-4(1); A0-9(2) .
TCTIAL PLANT COMPLEMENT - 138 o . __
/
6
I i
KINTE:NDENT y 50L*
I I I ROL SECURITY TEGNICAL SIAPF POWER PLANT ASSISTANT TO THE ENGINEER PLANT SUPERINTENDENT 1
1 -- SOL
- SOL
~
l
_ _ __. _ _ _J NUCLEAR GMN INSTRUMENT & l STAFP SECURITv ENGINEERS RADIATION CONTROLS UOMED SUPERVISkON 2 PROT M ION SUPERVISION 1 1
1 SUPERVISION g 50L*
RADIATION & CONTROL i PROCESS TEGNICIANS HONITORS 4 3
INSTRUMENT REPAIRMAN 1
13 GUARD (12) PORCE 53 54 (12 MEN PER SilIFT)
UNITS I AND 2 DIABLO CANYON SITE FIGURE 13.1-4 OPERATING ORGANIZATION Amendment 73 November 1978 (
l
\ - _ _ _ _ _ _ _ _
G APPENDIX 13.lA RESUMES O
f 13.1A-1
CHAPTER 13 - APPENDIX A - RESUMES Number Indivi dual Position 1 Raymond D. Ramsay Plant Superintendent 2 Robert Patterson Supervisor of Operations 3 James D. Shiffer Assistant to the Plant Superintendent 4 Donald A. Backens Supervisor of Maintenance 5 Not Selected
- Power Plant Engineer 6 Carl P. Schulze Relief Shif t Supervisor ;
7 Donald C. Bashaw Shift Foreman 8 Oliver A. Cole Shift Foreman 9 William J. Dilbeck Shift Foreman 10 Ronald L. Ewing Shift Foreman 11 Oscar E. Sundquist Shift Foreman 12 William F. White Shif t Foreman 13 Not Selected
- Shift Foreman l 14 Tim J. Martin Shift Foreman (Training Coordinator) 15 Stephen R. Fridley Assistant Training Coordinator i 16 John M. Gisclon Senior Nuclear Engineer 17 Richard C. Anoba Nuclear Engineer 18 William B. Kaefer Nuclear Engineer 19 David B. Miklush Nuclear Engineer 20 James A. Sexton Nuclear Engineer 21 Lawrence F. Womack, 11 Nuclear Engineer 22 Not Selected
- Nuclear Engineer 23 Jerome V. Boots Senior Chemical and Radiation Protection Engineer 24 William A. O'Hara Chemical and Radiation Protection Engineer 25 George P. M. Lyon Radiation and Process Monitor Foreman 26 Kenneth C. Doss Senior Instrument and Control Supervisor 27 Not Selected
- Instrument and Control Supervisor 28 Mark W. Stephens Instrument Maintenance Foreman 29 Not Selected
- Instrument Maintenance Foreman (November 1978) 13.lA-3 Amendment 73
Number Individual Position 30 Doyle L. Abney Mechanical Maintenance Foreman 31 William Robert Ryan Mechanical Maintenance Foreman 32 Gerald M. Zocher Electrical Maintenance Foreman 33 Robert M. Nanninga Maintenance Engineer 34 Jaunito S. Diamonon Quality Control Supervisor 35 Not Selected
- Quality Control Engineer 36 Bryan A. Dettman Security Supervisor 37 Larry C. Fisher Security Shif t Supervisor 38 Lawrence G. Lunsford Security Shift Supervisor 39 Vernon L. Sciocchetti Security Shif t Superviser 40 Robert R. Smith Security Shift Supervisor 41 Ronald G. Todaro Security Shif t Supervisor 42 Not Selected
- Senior Staff Engineer 43 Not Selected
- Staff Engineer
- Resumes for the positions will be submitted in future amendments when the positions are filled.
O Amendment 73 13.1A-4 (November 1978)
[.
- RESUME 1 PLANT SUPERINTENDENT 1
Raymond D. Ramsay Sheet 1 of 4 l
i
- 1. Birthdate '- December 23, 1921 r 2. Citizenship - USA l
! 3. Education
- a. High School l
- b. Mechanical Engineer, International Correspondence Schools
- 4. Employment History - Joined PG&E in September,1946 2
i j a. .0ctober,1940, to April,1944 - Apprentice Machinist, Mare Island )
[ Naval Shipyard.
1
- b. April, 1944, to May, 1946 - U.S. Navy.
- c. September,1946, to January,1952 - Joined PG&E and worked as Machinist,
- Lead Machinist, and Subforeman in power plants.
i l d. January,1952, to October,1957 - Power plant Mechanical Foreman.
l
- e. Octobc ,1957, to November,1960 - Assigned to Vallecitos Nuclear Power
- Plant as a Shift Foreman, subsequently promoted to Plant Superintendent.
~
Included assignments to Dresden and Shippingport.
I
- f. November,1960, to June,1961 - General Mechanical Foreman on Division
. staff overseeing maintenance activities of seven steam electrical gen-eration stations, including Vallecitos. Included an assignment to Dresden.
EO (November 1978) 13.lA-5 Amendment 73
Raymond D. Ramsay Sheet 2 of 4
- g. June,1961, to March,1962 - Assigned to Humboldt Bay Power Plant O
as Assistant Station Chief.
- h. March,1962, to September,1963 - Nuclear Maintenance Supervisor during construction, startup and initial operation of Humboldt Bay Unit 3.
- i. September,1963, to October,1964 - Plant Superintendent -
Vallecitos Nuclear Power Plant.
J. October,1964, to June,1970 - Plant Superintendent - Potrero Power Plant, a major fossil fueled station in San Francisco.
- k. June,1970, to August,1971 - Plant Superintendent, Humboldt Bay Power Plant. Participated in Diablo Canyon startup preparation including assignments at Connecticut Yankee, R. E. Ginna and g
H. B. Robinson Nuclear Power Plants. T
- 1. August,1971 - Assigned to Diablo Canyon as Plant Superintendent.
- 5. Nuclear Experience
- a. Vallecitos - Assigned to Vallecitos Nuclear Power Plant as a Shif t Foreman from October,1957, to October,1959. Participated in the startup and initial operation of VBWR. Supervised turbine overhaul in 1959 and general plant maintenance from 1957 through 1960. Pl ant Superintendent of Vallecitos from October,1959, to November,1960.
Received AEC Operator's License and qualified as a VAL Radiation Moni tor. Reassigned to Vallecitos Nuclear Power Plant as Plant Super-intendent from September,1963, through October,1964. Participated l in the preoperational testing, startup and initial operation of the nuclear superheat reactor and plant.
O' Amendment 73 13.1A-6 (November 1978) i _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ -
Raymond D. Ramsay Sheet 3 of 4
- b. Dresden - Assigned to Dresden for five weeks, September to October,1960, to observe overall plant operation and for three weeks in February,1961, to observe control rod drive modification work.
i
- c. Shippingport - Assigned to Shippingport for two weeks in 1960 to observe overall plant operation and maintenance.
l
- d. Humboldt Bay - Assigned to Humboldt Bay to follow Unit 3 construction, participated in the preoperational testing, initial loading, low-level testing, and power operation testing. Receiyu an AEC Senior Reactor Operator's License. Directed all maintenance activities at the plant. Reassigned to Humboldt Bay in June,1970, as Plant Superintendent with responsibility for all on-site activities.
O e. Connecticut Yankee - Assigned to Connecticut Yankee for two months, September, and October,1970, to observe overall plant operation and j maintenance. During this period short visits were made to the H. B.
Robinson Plant to observe startup operations and to the R. E. Ginna Plant to observe overall plant operation.
4
- f. Diablo Canyon - Participated in the review of licensing material for Units 1 and 2 including the FSAR and Technical Specifications.
Supervised plant staff engaged in the preparation of training mate-rial for operators, technicians, and maintenance personnel. Al so involved in preparation and review of administrative procedures, inservice inspection procedures, and operational quality assurance
. programs.
- g. Fonnal Training Courses
- 1) Radiation Monitors Course, VBWR, 1958.
- 2) Fundamentals of Nuclear Technology, APED,1959.
(Novembei 1978) 13.lA-7 Amendment 73
Raymond D. Ramsay Sheet 4 of 4
- 3) Introduction te Nuclear Power, Humboldt Bay Power Plant,1962.
0
- 4) Radiation Protection Training Course, Humboldt Bay Power Plant, 1962.
J
- 5) Humboldt Bay Equipment Description and Operation Course, Humboldt Bay Power Plant,1962.
- 5) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Wi nter,1971.
O
't O
Amendment 73 13.lA-8 (November 1978)
d 4
-RESUME 2 SUPERVISOR OF OPERATIONS c( )
Robert Patterson Sheet 1 of 4
- 1. Birthdate - April 4, -1931
- 2. Citizenship - USA
- 3. Education - BME, Cooper Union School of Engineering,1953 -
- 4. Employnent History - Joined PG&E in June,1953
- a. June,1953, to September,1953 - General indoctrination _ and training in power plant engineering.
- b. September,1953, to September, -1955 - U.S. Army Service. Assigned to the Ballistic Research Laboratory at Aberdeen Proving Ground.
r
- c. October,1955, to July,1959 - Various assignments involving power plant engineering ard technical operations. Involved in one conventional power plant startup.
- d. July,1959, to May,1961 - Staff engineering on Division staff including assignments in the nuclear power field at Vallecitos and Dresden.
- e. May,1961, to April,1962 - Engaged in Humboldt Bay Unit No. 3 startup preparation.
- f. - April,1962 to November,1964 - Various assignments in power plant nuclear engineering and other technict.1 operations at Humboldt Bay.
f(November 1978) 13.1A-9 Amendment 73
Robert Patterson Sheet 2 of 4 November,1964, to January,1968 - Assignment to Potrero Power Plant O
g.
for startup of 220 MWe conventional unit. Various assignments in power plant engineering and other technical operations at Potrero. Al so reassigned to Humboldt Bay during refueling outages to participate as shift nuclear engineer.
- h. January,1968, to January,1969 - Special assignment for preparation of Company power plant operator's training program and related manual.
- i. January,1969, to July,1970 - Assignment to Company's General Office engaged in Diablo Canyon license preparation. Includes a six-month assignment to R. E. Ginna PWR Plant.
- j. July,1970, to Augusi.,1971 - Engaged at Humboldt Bay in Diablo Canyon startup preparation on Diablo Canyon Task Force.
- k. August,1971 - Assigned to Diablo Canyon as Supervisc of Operations.
- 5. Nuclear Experience
- a. Vallecitos - Assigned to Vallecitos on a part-time basis from March, 1960, to May, 1961. Observed various phases of plant operation including the initial startup of the AYBWR,
- b. Dresden - Assigned to the Dresden for three months, September,1959, to December,1959, and for three weeks, June,1960, to July,1960. These periods included the initial loading and low-level testing and the half-power to full-power testing.
O Amendment 73 13.1A-10 (November 1978)
Robert Patterson Sheet 3 of 4 ;
O c. Humboldt Bay - Participated in prestartup activities including preparation of training material, initial loading and low-level testing procedures. Conducted training of operating personnel for AEC licensed exam. Received an AEC Senior Operator's License. Participated in preoperational testing of equipment and systems. Directed initial loading and testing programs as shift nuclear engineer. Directed the preparation of reactor refueling procedures subsequent to initial fuel-ing and directed the performance of this work on shif t. Responsible for the theoretical analyses of reactor core nuclear and thermal-hydraulic. performance, plus evaluation of the performance of plant safeguard and other auxiliary equipment.
- d. R. E. Ginna - Assigned to Ginna for six months from June,1969, to December,1969. Conducted training program for operators taking the AEC Operator Licens examination. Participated in the preoperational testing program and review of tert results for acceptance of systems.
O Participated in initial loading, low-level physics testings, and power operation testing programs.
- e. Diablo Canyon - Participated in the preparation and review of licensing material for Units 1 and 2 including PSAR, FSAR, and Technical Specifications. Supervised operating staff in the preparation of equipment operating procedures and related material prior to the startup of the plant.
a i
- f. Formal Training Courses ;
1
- 1) Introduction to Nuclear Physics - University of California j Extension, fall Semester, 1956. 1
- 2) Nuclear Reactor Engineering - University of California Extension, !
Spring Semester,1957.
O (November 1978) . 13.1A-11 Amendment 73
Robert Patterson Sheet 4 of 4
- 3) Radiological Health - USPHS and California Health Department, 0
three weeks, Spring,1958.
- 4) Neutron Physics - University of California Extension, Fall Semester, 1958.
- 5) Nuclear Radiation Detection - University of California Extension, Spring Semester,1959.
- 6) Radiation Biology - University of California Extension, Spring Semester, 1961.
- 7) Reactor Survey Course - General Electric APED, Spring Semester, 1961 (sections on instrumentation, core design, and operation).
- 8) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Winter, 1971.
- 9) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three week course,1974) and Option II (one week course, 1978).
O Amendment 73 13.lA-12 (November 1978)
RESUME 3 ASSISTANT TO THE PLANT SUPERINTENDENT S
James D. Shiffer Sheet 1 of 4
- 1. Birthdate - !! arch 24, 1938
- 2. Citizenship - USA
- 3. Education
- a. B.S. Chemical Engineering, Stanford University,1960.
- b. 'H.S. Nuclear Engineering, Stanford University,1961.
- c. Registered Professional Engineer (M.E. and N.E.), California.
- 4. Employment History - Joined pG&E in September,1961.
O a. Suniner,1959 and Sununer,1960 - Employed by PG&E in sunrner engineer program. Assigned to Vallecitos Boiling Water Reactor and Central Chemical Laboratory.
- b. September,1961, to April,1962 - Engaged in Humboldt Bay Unit No. 3 startup preparation.
c, April,1962, to July,1969 - Various assignments in power plant nuclear engineering, chemical engineering, and other technical operations at Hun 1boldt Bay,
- d. July,1969, to July,1970 - Engaged at Humboldt Bay and Company's General Office in Diablo Canyon startup preparation. Includes a seven-month assignment in Rochester, New York during startup and initial testing of R. E. Ginna PWR Plant.
(November 1978) 13.l A-13 Amendment 73
James D. Shiffer Sheet 2 of 4 9
- e. July,1970, to August,1971 - Engaged in Diablo Canyon startup preparation on Diablo Canyon Task Force.
- f. August,1971, to October,1978 - Assigned to Diablo Canyon as Power Plant Engineer.
- g. November,1978 - Assigned to Diablo Canyon as Assistant to the Plant Superintendent.
- 5. Nuclear Experience
- a. Education - Master degree thesis research involving operation of the Stanford Swimming Pool Reactor, irradiation of foils to determine reactor parameters, flux wire counting and radiochemical work.
- b. Vallecitos - Assigned to Vallecitos for two summers. Participated in the startup of the AVBWR plant.
- c. Humboldt Bay - Participated in prestartup activities including preparation of training material, initial loading and low-level testing procedures and power testing procedures. Conducted training of operating personnel for AEC license examinations. Received an AEC Senior Operator's License. Participated in preoperational testing of equipment and systems. Directed initial loading and testing programs as shift nuclear engineer. Directed the preparation of all reactor refueling procedures subsequent initial fueling and directed the performance of this work on shift. Responsible for the theoretical analyses of reactor core nuclear and thermal-hydraulic performance, plus evaluation of the performance of plant safeguard and other auxiliary equipment. Provided technical advice and guidance for the chemical and radiation protection engineers and participated in the establishment and implementation of the chemical, radiochemical, and radiation protection programs at the plant.
l Amendment 73 13. l A-14 (November 1978)
( - - - - - - _ _ - - - _ _ - _ - - - - _ _ _ _ - _
James D. Shiffer Sheet 3 of 4
- d. R. E. Ginna - Assigned to Ginna for seven months from July,1969, to February, 1970. Conducted training program for operators taking the AEC Operator's License examination. Participated in the preparation I and review of procedures and programs for initial loading, low-level physics testing, power operation testing, and radiochemical control.
Participated in initial loading, low-level physics testing, and power operation testing programs.
l
- e. Diablo Canyon - Participated in the preparation and review of licensing l material for Units 1 and 2 including PSAR, FSAR, and Technical :
Speci fications. Supervised staff of engineers (including persons experienced in nuclear engineering, instrumentation, radiation protection, and chemical engineering) engaged in the preparation of equipment operating and testing procedures, equipment specifications and related material required prior to the startup of the plant.
- f. Formal Training Courses
- 1) Stanford University Nuclear Engineering Curriculum as required by AEC Scholarship Program.
- 2) Digital Computer Applications for Nuclear Reactor Calculations, UCLA Extension, Spring,1963.
- 3) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Wi nter, 1971.
- 4) In-Place Filter Testing Workshop, Harvard School of Public Health, Fall, 1971.
- 5) Refresher Training in Radiological Engineering, General Electric Vallecitos Nuclear Center, Summer, 1972.
(~]
%/
(November 1978) 13.1 A-15 Amendment 73 l
i James D. Shiffer Sheet 4 of 4 O
- 6) Short Course in Reactor Noise Analysis, University of Tennessee,
, Fall, 1976.
- 7) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three week course,1974) and Option II (one week course, 1978).
- O O
Amendment 73 13. l A-16 (November 1978)
RESUME 4 SUPERVISOR OF MAINTENANCE
() Donald A. Backens Sheet 1 of 2 i
- 1. Birthdate - February 26, 1923
. '2. Citizenship - USA
- 3. Education
- a. High School
- b. California Maritime' Academy 1 i
- 4. Employaent History -- Joined PG&E in March,1946 a.
March,1946, to May,'1962 - Assignments in mechanical maintenance of 4
various conventional power plants as a machinist.
b.
May, .1962, to August,1971 - Assigned to Humboldt Bay Power Plant as Mechanical' Maintenance Foreman. Includes an assignment at the San Onofre PWR Plant.
- c. August,1971 - Assigned to Diablo Canyon as Supervisor of Maintenance.
- 5. Nuclear Experience
- a. Humboldt Bay - Directed and supervised all significant mechanical maintenance since initial power operation of Unit 3 including all major modifications, nondestructive inservice inspections and quality assurance programs. '
i b.. San Onofre - Assigned to San Onofre for one month in 1970 to. observe maintenance procedures and practices during refueling.
()
.c. . Zion - Assigned to Zion Nuclear Station for three weeks in 1974 to observe. maintenance procedures and ' practices during operation. l (November 1978) 13.1A-17 Amendment 73
Donald A. Backens Sheet 2 of 2
- d. Diablo Canyon - Participated in the design review, quality assurance 9
program preparation, specification of spare parts and machine shop layout.
- e. Formal Training Courses
- 1) Radiation Protection Training Course, Humboldt Bay Power Plant and Diablo Canyon Power Plant.
- 2) Nondestructive Testing and Inservice Inspection Seminar, Southwest Research Institute, 1971.
- 3) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Winter, 1971.
- 4) Nondestructive Testing - On-site class taught by General Dynamics /Convair instructors, January,1972.
- 5) EPRI Workshop on Outage Planning and Maintenance Management, August, 1978.
- 6) Quality Assurance Seminar presented by General Atomic Company, San Diego, July,1976.
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Amendment 73 13. l A-18 (November 1978 I
1 1
RESUME 5 RELIEF SHIFT SUPERVISOR Harvey J. Maloney Sheet 1 of 3
- 1. Birthdate - September 15, 1921 d
- 2. Citizenship - USA
- 3. Education
- a. High School
- b. Three years of college in Industrial Engineering and Business Administration, California State Polytechnic College and College of the Redwoods.
- 4. Employment History - Joined PG&E in September,1945 J
l v) a. September,1945, to September,1958 - Assignment to conventional steam power plant operations. Promoted to Shift Foreman in February, 1955.
Participated in the startup of Hunters Point Unit 3, Moss Landing Units 1, 2, 3, 4, and 5, Morro Bay Units 1 and 2.
- b. September,1958, to September,1959 - Assigned to Vallecitos Nuclear Power Plant for licensing and operation as Shift Foreman.
- c. September,1959, to December,1960 - Assigned to Morro Bay Power Plant as Shift Foreman.
- d. January,1961 to April,1961 - Assigned to AVBWR for startup testing and operation.
O U
(November 1978) 13. l A-19 Amendment 73
Harvey J. Maloney Sheet 2 of 3
- e. May,1961, to January,1962 - Engaged in special assignments in O
nuclear power field inclLding Humboldt Bay startup preparation work 2
and assignments to Dresden Unit i for four weeks to observe overall plant operations.
l f. January,1962, to July,1970 - Assigned to Humboldt Bay Power Plant as Shift Foreman for initial startup and power operation of Unit 3.
- g. July,1970, to August,1971 - Engaged at Humboldt Bay in Diablo Canyon Power Plant startup preparation on Diablo Canyon Task Force. Included an assignment to R. E. Ginna PWR Plant to observe overall plant operation.
- h. August,1971 - Assigned to Diablo Canyon as Relief Shift Supervisor.
- 5. Nuclear Experience O
- a. Vallecitos - Assigned as Shift Foreman for operation of VBWR and AVBWR.
Received operator license and qualified as a Radiation Monitor.
- b. Dresden - Assigned to Dresden for four weeks in June,1961, to observe overall plant operations.
- c. Humboldt Bay - Participated in prestartup activities of Humboldt Bay Power Plant Unit 3 including preparation of training material core loading and startup testing of equipment and systems. Received an AEC Senior Operator's License. Directed numerous refueling operations and power operations as Shift Foreman,
- d. R. E. Ginna - Assigned to Ginna for four weeks in July,1970, to observe overall plant operations.
O.
Amendment 73 13. l A-20 (November 1978)
Harvey J. Maloney Sheet 3 of 3 O e. Diablo Canyon - Participated in the preparation of equipment description and operating instructions, and other material related to the startup of Units 1 and 2.
I l
- f. Formal Training Courses i
- 1) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse l
APD, Wi nter,1971. )
- 2) Fundamentals of Nuclear Technology APED,1959. 1 i
- 3) Radiation Monitors Course, VBWR,1959.
- 4) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
- 5) Radiation Protection Training Course, Humboldt Bay Power Plant, 1962.
- 6) Humboldt Bay Power Plant Unit 3 Equipment Description and Operation {
Course, 1962.
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(November 1978) 13.lA-21 Amendment 73
RESUME 6 RELIEF SUPERVISOR Carl P. Schulze Sheet 1 of 3
- 1. Birthdate - December 3,1917 I
- 2. Citizenship - USA
- 3. Education - High School
- 4. Employment History - Joined PG&E in September,1948
- a. December,1936, to December,1940, and January,1942, to October, 1945 - U.S. Navy Service. Various shipboard assignments.
- b. September,1948, to November,1951 - Employed by PG&E with assignment to Avon Power Plant. Various assignments in operations.
- c. November,1951, to June,1958 - Assigned to Contra Costa Power Plant.
Participated in the initial startup of two conventional units.
- d. June,1958, to October,1960 - Promoted to Shift Foreman and assigned to Martinez Power Plant.
- e. October,1960, to March,1962 - Assigned to Vallecitos Nuclear Power Plant for licensing and operation as a Shift Foreman.
- f. March,1962, to July,1971 - Assigned to Contra Costa Power Plant.
Participated in the initial startup of two, 330 MWE conventional units.
Includes assignments to Vallecitos, June,1963, to December,1963, and May,1964, to August,1965, and a four week assignment to Dresden in August,1961, to observe overall plant operation.
- g. July,1971, to May,1972 - Assigned to Humboldt Bay Power P1 ant for refamiliarization in nuclear power and for licensing.
Amendment 73 13.lA-22 (November 1978)
Carl P. Schulze Sheet 2 of 3
- h. May,1972 - Assigned to Diablo Canyon as Shift Foreman.
- i. November,1978 - Promoted to Relief Shift Supervisor at Diablo Canyon.
S. Nuclear Experience
- a. Vallecitos - Assigned as a Shift Foreman for operation of VBWR and AVBWR. Received an AEC Operator's License and qualified as a radiation monitor.
- b. Dresden - Assigned to Dresden for four weeks in August,1961, to observe overall plant operation.
- c. Humboldt Bay - Assigned to Humboldt Bay for retraining in nuclear power plant operation. Received an AEC Operator's License.
- d. Zion Station - Assigned to Zion for one month to observe overall operations.
- e. Diablo Canyon - Participated in the preparation of equipment descriptions, operating procedures, and other material relating to the startup of Units 1 and 2. Participated in all preoperational and startup testing programs.
- f. Formal Training Courses
- 1) Radiation Monitors Course, VBWR,1960.
- 2) Humboldt Bay Power Plant Unit 3 Equipment Description and Operation Course, 1971.
- 3) Introduction to Nuclear Power, Humboldt Bay Power Plant,1971.
O (November 1978) 13.1A-23 Amendment 73
Carl P. Schulze Sheet 3 of 3
- 4) Radiation Protection Training Course, Humboldt Bay Power Plant, 0
1971.
- 5) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection, equipment description and operation.
Diablo Canyon Power Plant, 1974 - 1978.
- 6) Simulator Trdining - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three week course,1974) and Option 11 (one week course,1978).
O O
Amendment 73 13.lA-24 (November 1978)
~ - - . - - . - - . - - . - . . _ - . . . _ . . _ _ _ . - - - _ _ _
i i- RESUME 7
- SHIFT FOREMAN Donald C. Bashaw Sheet 1 of 2
.1. Birthdate - December 1,1922
, 2. Citizenship - USA
- 3. Education - High School
- 4. Employment History - Joined PG&E in September,1950
- a. September,1949, to October,1945 - U.S. Army Air Corps Service.
- b. September,1950, to February,1962 - Employed by PG&E and assigned to 1
Moss Landing Power Plant. Various assignments in plant operations including participation in the initial startup of four conventional units.
- c. February,1962, to June,1967 - Assigned to Humboldt Boy Power Plant for initial startup and power operation of Unit 3.
- d. June,1967, to June,1971 - Promoted to Shift Foreman and assigned to Potrero Power Plant.
- e. June,1971, to October,1971 - Assigned to Humboldt Bay Power Plant for refamiliarization with nuclear plant operations.
- f. October,1971 - Assigned to Diablo Canyon as Shift Foreman.
- 5. Nuclear Experience
- a. Humboldt Bay - Participated in all startup activities including the preoperational test program, initial core loading and power escalation testing as well as routine power operation. Received an AEC Operator's License.
(November 1978) 13.1A-25 Amendment 73
I Donald C. Bashaw Sheet 2 of 2
- b. R. E. Ginna - Assigned to Ginna for four weeks in July,1972, to Ol observe overall plant operation.
- c. Diablo Canyon - Participated in the prepara1 ion of the equipment Description and Operation Manual, Operating Procedures, and other material related to the startup of Units 1 and 2. Participated in all pre-operational testing programs.
- d. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
i 1
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant, 1962.
- 3) Humboldt Bay Equipment Description and Operation Course, Humboldt Bay Power Plant,1962.
- 4) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Wi nter , 197] .
- 5) Diablo Canyon License Preparation Course - Consisted of reactor theory, radiation protection, equipment description and operation.
Diablo Canyon Power Plant, 1974 - 1978.
- 6) Simulat Training - Westinghouse Nuclear Training Center, Zion, 1111noi- Option III (three week course,1974) and Option 11 (one week cour.;e,1978).
O Amendment 73 13. l A-26 (November 1978)
1 RESUME 8 i SHIFT FOREMAN Oliver A. Cole -
Sheet 'l of 3
- 1. Birthdate - December. 29, 1925 .j
- 2. Citizenship - USA '
- 3. Education High School 1
- 4. Employment History - Joined PG&E in March,1953
- a. June,1943, to June,1949 - U.S. Maritime Service, attended Mare Island -
Naval Shipyard Machinist Apprentice School and U.S. Maritime Training School.- Secured Marine Engineer's License. 3
- b. March,1953, to February,1962 - Employed by PG&E in power plant operations. Participated in the initial startup of two conventional O units at Contra Costa Powr Plant.
i
- c. February,1962, to July,1970 - Assigned to Humboldt Bay Power Plant for initial startup and powei operation of Unit 3 as a Senior Control Operator.
- d. . July,1970, to August,1971 - Assigned to Humboldt Bay Power Plant as Shift Foreman. Also engaged in Diablo Canyon startup preparation as member of Diablo Canyon Task Force. Includes an assignment at the R. E. Ginna PWR Plant.
- e. August,1971 - Assigned to Diablo Canyon as Shift Foreman. Includes a six-month training assignment to Morro Bay Power Plant as Shift Foreman. ,
O .
(November 1978) 13.lA-27 Amendment 73 -
l
Oliver A. Cole Sheet 2 of 3
- 5. Nuclear Experience O
- a. Humboldt Bay - Assigned to Humboldt Bay Power P1 ant Unit 3 as a Senior Control Operator. Participated in the preoperational test program, the initial loading, and low level and power operation test programs.
Received AEC Senior Operator's License. Directed power operation and refueling activities as Shif t Foreman.
- b. R. E. Ginna - Assigned to Ginna for four weeks in August,1970, to observe overall plant operations. 1
- c. Diablo Canyon - Participated in the preparation of equipment description l and operating instructions, and other material related to the startup of Units 1 and 2. f i
- d. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
O)
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant,
]
1962.
- 3) Humboldt Bay Equipment Description and Operation Course, Humboldt Bay Power Plant,1962.
- 4) Several PG&E sponsered courses including Steam Power Plant Fundamentals, Fundamental of Electricity, and Nuclear Power Plant Fundamentals.
- 5) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment. Westinghouse APD, Winter, 1971.
O Amendment 73 13.1A-28 (November 1978)
l Oliver A. Cole
]
Sheet 3 of 3
- 6) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection, equipment description and operation.
Diablo Canyon Power Plant, 1974-1978.
l
- 7) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illi noi s. Option III (three-week course,1974) and Option II (one-week course, 1978). ;
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O (November 1978) 13.1A-29 Amendment 73
RESUME 9 SHIFT FOREMAN William J. Dilbeck Sheet 1 of 3 i
- 1. Birthdate - July 4,1927 1
- 2. Citizenship - USA l 3. Education - High School l
- 4. Employment History - Joined PG&E in April,1952 I a. June,1945, to July,1949 - U.S. Navy Service. Various shipboard assignments in operations.
- b. April,1952, to April,1955 - Employed by PG&E with assignment to Moss Landing Power Plant. Participated in initial startup of two conventional units.
O
- c. April,1955, to April,1962 - Assigned to Morro Bay Power Plant as Auxiliary Operator and subsequently as Control Operator. Participated in the initial startup of four conventional units.
l
- d. April,1962, to August,1971 - Assigned to Humboldt Bay Power Plant l Unit No. 3. Participated in the initial startup as a Control Operator, subsequently promoted to Senior Control Operator,
- e. August,1971 - Promoted to Shift Foreman and assigned to Diablo Canyon to participate in startup preparation. Includes an assignment at the Point Beach PWR and a six-month training assignment at Morro Bay Power Plant as Shift Foreman.
l 9'
Amendment 73 13.lA-30 (November 1978)
i William J. Dilbeck Sheet 2 of 3 ,
- 5. Nuclear. Experience
- a. Humboldt Bay - Participatedin startup activities including the l preoperational test program, the initial' core loading and power i escalation testing. Received AEC Senior Operator's License, i i
l b. Point Beach - Assigned to Point Beach for one month to observe i overall plant operation.
l
- c. Diablo Canyon - Participated in preparation of equipment description and operation material and other material relating to the startup of Units 1 and 2. Participated in all preoperational and startup testing programs.
- d. Formal Training
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
I
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant, 1
1962.
i
- 3) Humboldt Bay Power Plant Unit 3 Equipment Description and Operation, 1962.
- 4) Fireman and Basic Engineering School, U.S. Navy,1945. )
- 5) Diablo Canyon License Preparation - Consisted of reactor theory, ;
radiation protection, equipment description and operation. I Diablo Canyon Power Plant, 1974-1978.
i
- 6) Simulator Training - Westinghouse Nuclear Training Center, j
Zion, Illinois. Option III (three-week course,1974) and I
Option 11 (one-week course, 1978). l (November 1978) 13.1A-31 Amendment 73 l
4 1
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O 1
1 1
1 1
O Amendment 73 13.lA-32 (November 1978)
)
RESUME 10
. SHIFT FOREMAN,
. - Ronald L. Ewing Sheet 1 of 2-l Birthdate . November 3,1941 1.
4
- 2. Citizenship - USA 2
l l 3. Education - High School .
l 4. Employment History - Joined PG&E in November,1966.
) a. October,1961 to July,1964 - U.S. Army.
1
! b. November 1966 to June,1973 - Employed by PG&E at Humboldt Bay i Power Plant. Promoted to Control Operator (Reactor operator) in f August, 1971.
L i
l' c. June,1973 - Promoted to. Senior Control .0perator and assigned to l Diablo Canyon.
- d. December,1976'- Promoted to Shift Foreman at Diablo Canyon.
- 5. Nuclear Experience L
- a. Humboldt Bay - Advanced through all operating classifications at
! the Plant to the position of Control Operator. Participated in 4
startups, shutdowns, scram recoveries, power operation, refueling i operation, and-special' tests. Received an AEC Reactor Operator's License in January,1970.
I b. Diablo Canyon l- Participated in the initial training programs and startup testing of the plant.
- O (November 1978)I 13.1A-33 Amendment 73 l
Ronald L. Ewing Sheet 2 of 2 O
- c. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1969.
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant, 1969.
- 3) Humboldt Bay Equipment Description and Operation Course, Humboldt Bay Power Plant,1969.
- 4) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection, equipment description and opera-tion. Diablo Canyon Power Plant, 1974-1978.
- 5) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course, 1974) and Option II (one-week course,1978).
Amendment 73 13.1A-34 (November 1978)
I t _ _ - - _ _ - - - - _ _ - - - - - - - - - - - - - - - - . -- - - - - - - - - - - - - _ - - . - - - -
i RESUME 11 SHIFT FOREMAN Oscar E. Sundquist h Sheet 1 of 3
~
l 1. Birthdate - August 20, 1924 l
- 2. Citizenship - USA !
- 3. Education f
l a. High School 4
4
- b. Three years of college in Mechanical Engineering, North Dakota State College, Michigan State College, and the University of Washington.
4
- 4. Employnent History'- Joined PG&E'in January,1952 i
^
! a. February,1943, to March,1947 - U.S. Army Air Force.
- b. January,1952, to February,1962 - Employed by PG&E at Moss Landing Power Plant. Participated in the initial startup of two conventional l j units.
1 l
- c. February,1962, to February,1973 - Assigned to Humboldt Bay Power Plant for initial startup and power operation of Unit 3. Promoted to Shift Foreman in March,1970.
l d. February,1973 - Assigned to Diablo Canyon as Shift Foreman.
4 i
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(November.1978) 13.1A-35 Amendment 73
car E. Sundquist Sheet 2 of 3 O
- 5. Nuclear Experience
- a. Humboldt Bay - Assigned to Humboldt Bay Power Plant Unit 3 as a Control Operator. Participated in the preoperational test program, the initial loading, and low level and power operation test programs. Subsequently received an AEC Senior Operator License and was promoted to Shift Fore-man. Directed power operation and refueling activities as a Shift Foreman,
- b. Connecticut Yankee - Assigned to Connecticut Yankee for four weeks in September,1970, to observe overall plant operations.
- c. Diablo Canyon - Participated in the preparation of operating procedures and other material related to the startup of Units 1 and 2. Partici-1 pated in all pre-operational and startup test programs.
- d. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant, 1962.
- 3) Humboldt Bay Equipment Description and Operation Course, Humboldt Bay Power Plant,1962.
- 4) Several PG&E sponsored courses including Steam Power Plant Fundamentals, and Nuclear Power Plant Fundamentals.
- 5) Diab'lo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment. Westinghouse APD, Winter, 1971.
O Amendment 73 13.lA-36 (November 1978)
Oscar E. Sundquist
_. Sheet 3 of 3
%./
- 6) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection, equipment description and operation. Diablo Canyon Power Plant, 1974-1978.
- 7) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course,1974) and Option II (one-week course,1978).
O O
(November 1978) 13.1A-37 Amendment 73
RESUME 12 SHIFT FOREMAN William F. White Sheet 1 of 3
- 1. Birthdate - June 20, 1920
- 2. (.itizenship - USA
- 3. iducation
- a. High School
- b. Two Years College
- 4. Employment History - Joined PG&E in September,1945
- a. January,1942, to September,1945 - U. S. Navy Service. Various
, shipboard assignments as Machinist Mate.
- b. September,1945, to June,1970 - Employed by PG&E in power plant operations, participated in the startup of cotwentional units at Hunters Point and Potrero Power Plants. Promoted to Shift Foreman Ja nuary , 1967.
- c. July,1970, to August,1971 - Assigned to Humboldt Bay Power Plant as Shift Foreman. Also, participated in Diablo Canyon startup preparation work.
- d. August,1971 - Assigned to Diablo Canyon as Shift Foreman. Includes an assignment at the R. E. Ginna PWR Plant.
- 5. Nuclear Experience
- a. Humboldt Bay - Assigned to Humboldt Bay Power Plant as Shift Foreman.
Received AEC Operator's License. Participated in power operation and the 1971 refueling of the unit.
Amendment 73 13. l A-38 (November 1978)
I
William F. White i
Sheet 2 of 3 O l 1
- b. R. E. Ginna - Assigned to Ginna for one month in 1972 to observe overall plant operation. i l
4 I
- c. Diablo Canyon - Participated in the preparation of the Equipment Description and Operating Manual and other material related to the startup of Units 1 and 2. Participated in all preoperational and startup test programs.
I
- d. Formal Training Courses '
- 1) Numerous PG&E sponsored courses related to the operation of Power Plants and Electrical Systems.
< 2) Principles of Nuclear Reactor Operation, University of California ,
1 Extension, 1963.
- O 3) Industrial Atomic Energy Uses, Hazards, and Controls, IBEW union ;
course,1963.
- 4) Radiological Health and Basic Science Review U.S. Department of Health Education and Welfare,1970.
l 5) Introduction to Nuclear Power, Humboldt Bay Power Plant,1970.
- 6) Radiation Protection Training Course, Humboldt Bay Power Plant,
- 1970.
- 7) Humboldt Bay, Equipment Description and Operation Course, Humboldt Bay Power Plant,1970.
- 8) Diablo Canyon Design Lecture Series -- A series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD,1971, 1-(November 1978) 13.1A-39 Amendment 73
William F. White Sheet 3 of 3
- 9) Disblo Canyon License Preparation - Consisted of reactor 0
theory, radiation protection, equipment description and operation. Diablo Canyon Power Plant, 1974-1978,
- 10) Simulator Training - Westin9 house Nuclear Training Center, Zion, Illinois. Option III (three-week course, 1974) and Option II (one-week course,1978).
O 4
O Amendment 73 13.lA-40 (November 1978)
}
RESUME 14 TRAINING C0ORDINATOR (j Tim J. Martin Sheet 1 of 2
- 1. Birthdate - March 23, 1947
- 2. Citizenship - USA
- 3. Education - High School
- 4. Employment History - Joined PG&E in October,1965.
- a. October,1965 to April,1967 - Employed by PGandE in the General Construction Department.
- b. April,1967 to September,1969 - Transferred to Morro Bay Power Plant and assigned to the Operations Group.
- c. September,1969 to June,1973 - Assigned to Humboldt Bay Power Plant. Promoted to Control Operator (Reactor Operator) in June, 1972.
I
- d. June,1973 - Promoted to Senior Control Operator and assigned to Diablo Canyon.
- c. October,1976 - Promoted to Shift Foreman at Diablo Canyon.
- f. September,1978 - Appointed Training Coordinator at Diablo Canyon.
1 b
G (November 1978) 13.1A-41 Amendment 73
Tim J. Martin Sheet 2 of 2 O
- 5. Nuclear Experience
- a. Humboldt Bay - Advanced through all operating classifications at the Plant to the position of Control Operator. Participated in startups, shutdowns, scram recoveries, power operation, refuelings and special tests. Received an AEC Reactor Operator's License in July,1971 (0P-2973).
- b. Diablo Canyon - Participated in the initial training programs and l startup testing of the plant.
- c. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1970.
- 2) Radiation Protection Training Course, Humboldt Bay Power Plant, 1970.
- 3) Humboldt Bay Equipment Description and Operations Course, Humboldt Bay Power Plant,1970.
- 4) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection equipment description and opera-tion. Diablo Canyon Power Plant, 1974-1978.
- 5) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course, 1974) and Option II (one-week course,1978).
O Amendment 73 13.1A-42 (November 1978)
RESUME 15 ASSISTANT TRAINING COORDINATOR Stephen R. Fridley i Sheet 1 of 2
! 1. Birthdate - June 24, 1948 i 2. Citizenship - USA
- 3. Education - High school and 3 years of college.
- 4. Employment History - Joined PG8E in August,1970.
August,1970 - July,1974 - Employed by PGandE at liumboldt Bay a.
Power Plant.in the Operations Group.
e
- b. June 1973 - Promoted to Assistant Control Operator - Humboldt Bay Power Plant.
O c. July,1974 - Promoted to Control Operator (Reactor Operator) and assigned to Diablo Canyon Power Plant.
- d. July,1978 - Promoted to the position of Assistant Training Coordinator at Diablo Canyon Power Plant. l S. Nuclear Experience
- a. Humboldt Bay - Advanced through all operating classifications at the Plant to the position of Assistant Control Operator. Partici-pated in startups, shutdowns, scram recoveries, power operation, refuelings and special tests. Received an AEC Reactor Operator's License .in July,1973 (0P-3346). ;
l
- b. Diablo Canyon - Participating in the initial training programs and startup testing of the plant.
O (November 1978) 13.lA-43 Amendment 73
Stephen R. Fridley Sheet 2 of 2 O
- c. Formal Training Courses
- 1) Introduction to Nuclear Power, riumboldt Bay Power Plant,1972.
- 2) Radiation Protection Training Course, Humboldt Bay Power Pl a nt , 1972.
I
- 3) Humboldt Bay Equipment Description and Operations Course, Humboldt Bay Power Plant,1972.
- 4) Diablo Canyon License Preparation - Consisted of reactor theory, radiation protection, equipment description and operation.
Diablo Canyon Power Plant, 1974-1978.
- 5) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option II (one-week course) 1978.
O Amendment 73 13.lA-44 (November 1978)
RESUME 16 SENIOR NUCLEAR ENGINEER
( John M. Gisclon Sheet 1 of 3
- 1. Birthdate - August 16, 1938
- 2. Citizenship - USA
- 3. Education
- a. B.S. Mechanical Engineering, University of Nevada,1961
- b. Registered Professional Engineer (M.E. ), Nevada
- c. Registered Professional Engineer (Nuclear), California
- 4. Employment History - Joined PG&E in August,1970.
/~N G) a. Summer,1960 - Jenior Test Engineer - conventional power plant -
Kennecott Copper Corp. , McGill, Neveda.
l
- b. Summer,1961 - Engineering Trainee - Pacific Gas and Electric Company, Contra Costa Power Plant.
- c. September,1961, to July,1965 - Active duty U.S. Navy. Attended U.S. Naval School, Pre-Flight School. Commissioned as a Line Officer. Served in various surface shipboard assignments.
- d. October,1965, to September,1966 - Assigned by PG&E as Engineering Trainee to Pittsburg Power Plant. Various power plant engineering l a s si gnments.
- e. September, 1966, to November, 1968 - Various assignments in power plant nuclear engineering, testing, and technical operations at l Humboldt Bay Power Plant.
(November 1978) 13.1A-45 Amendment 73
John M. Gisclon Sheet 2 of 3
- f. November,1968, to August,1970 - Joined Westinghouse Electric 9
4 Corp. , (NRF - Bettis Atomic Power Laboratory) as Plant Engineer.
Various assignments in maintenance and modification of equipment and systems, and as design liaison for the liquid radwaste disposal system.
- g. August,1970, to August,1971 - Joined PG&E, engaged at Humboldt Bay in Diablo Canyon startup preparation as a member of the Diablo Canyon Task Force. Includes an assignment to the H. B. Robinson PWR Plant.
- h. August,1971 - Assigned to Diablo Canyon as Nuclear Engineer.
- i. May,1974 - Promoted to Senior Nuclear Engineer.
- 5. Nuclear Experience
- a. Humboldt Bay - Participated in refueling operations, and nomi plant operations, testing, and modification design as nuclear engi-neer. Prepared refueling procedures and directed this work as shift nuclear engineer. Responsible for the theoretical analysis of reactor core nuclear and thermal hydraulic performance plus evalua-tion of the performance of plant safeguard and other auxiliary equipment. Participated in revision of equipment description and operating procedures. Conducted visual fuel examination and fuel sipping programs. Conducted training programs for AEC license examination candidates.
1
- b. Westinghouse Electric Corp. (NRF) - Participated in various j maintenance and modification projects as cognizant engineer. Projects include steam generator tube removal, design liason for liquid rad- I waste disposal systems, and steam generator water level control. l
- c. H. B. Robinson - Assigned to Robinson for three months during 1970 and 1971. Participated in low-level physics ano power escalation test program as a member of the Westinghouse startup team.
Amendment 73 13.1A-46 (November 1978) l I
John M. Giscion Sheet 3 of 3
- d. Diablo Canyon - Participated in and supervised other Nuclear Engineers in the preparation and review of licensing material for Units 1 and 2 including the FSAR; Technical Specifications; equip-ment description and operating instructions; testing procedures; administrative procedures and operational quality assurance manual.
Participated in startup te., ting program. Supervised surveillance testing program,
- e. Formal Training Courses
- 1) University of Idaho - NRTS Graduate Education Program 1969-1970 Master of Nuclear Science Curriculum.
a) Reactor Physics for Engineers b) Nuclear Reactor Engineering
- 2) Diablo Canyon Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Winter, 1971.
- 3) Nondestructive Testing - On-site class taught by General Dynamics /Convair instructors, January,1972.
- 4) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course,1972) and Op?. ion II (one-week course,1978).
13,1A-47 Amendment 73 (November 1978)
RESUME 17 NUCLEAR ENGINEER Richard C. Anoba Sheet 1 of 3
- 1. Birthdate - February 2,1951
- 2. Citizenship - USA
- 3. Education
- a. B.S. Nuclear Engineering, University of California, Santa Barbara, 1973.
- b. M.S. Nuclear Engineering, Mass 6chutetts Institute of Technology, 1975.
- 4. Employment History - Joined PGandE in June,1978.
- a. August,1973 to August,1974 - Employed by the Atomic Energy Commission. Reactor Engineer responsible for the construction and operation of various test facilities in the Liquid Metal Fast Breeder Program.
- b. August,1974 to August,1975 - Employed by the Energy Research and Development Administration. Assigned to the Massachusetts Institute of Technology for Advanced Training engaged in the develop-ment of flow measurement techniques in an LMFBR blanket assembly using laser anemometry.
- c. August,1975 to August,1976 - Employed by ERDA. Assigned to the Hanford Engineering Development Laboratory. Worked as a Test Facility Engineer responsible for various test programs and design efforts in support of the construction of the Fast Flux Test Facility.
O Amendment 73 13.lA-48 (November 1978)
Richard C. Anoba Sheet 2 of 3
- u. August,1976 to June,1978 - Employed by the Department of Energy.
Assigned to the FFTF Project Office. Worked as a Plant Systems Engineer responsible for the construction, startup and operation of various systems in the FFTF.
- e. June,1978 - Assigned to Diablo Canyon Power Plant as a Nuclear Engineer.
- 5. Nuclear Experience
- a. Education - Master's degree thesis research involving the development flow measurement techniques in an LMFBR blanket assembly using :
laser anemometry.
- b. Atomic Energy Commission - Participated in the design of a sodium p removal facility at the Liquid Metal Engineering Center and the development of a low chloride sodium fire extinguishant.
- c. Energy Research and Development Administration - Involved in a variety of projects associated with FFTF, including the development of an air tight electrical penetration for highly irradiated FFTF sodium cells, design of a sodium test loop, development of a method for corrosion product removal from the water cooling jacket of an electromagnetic pump, and a test program to increase the heat transfer capability of the dump heat exchanger.
- d. Department of Energy - Project Engineer involved with the design, construction and startup of the FFTF heating and ventilation systems, radioactive gas processing systems and auxiliary cooling systems.
Participated in a special task force to verify the seismic design of various installed equipment in FFTF. Participated in the integrated leak rate test for FFTF.'
O (November 1978) 13.1 A-49 Amendment 73
Richard C. Anoba Sheet 3 of 3
- e. Diablo Canyon - Nuclear Engineer engaged in procedure preparation 9
and startup testing of various plant systems and equipment.
- f. Formal Training Courses
- 1) Massachusetts Institute of Technology Curriculum as required by the Atomic Energy Commission Intern Program.
- 2) Extension courses on nuclear design and fuel management at Catholic University, Washington, D.C. ,1973.
- 3) Three-day sodium loop operations training course at the Hanford Engineering Development Laboratory,1975.
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e i Amendment 73 13. l A-50 (November 1978) l l
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1 RESUME 18 l
- NUCLEAR ENGINEER William B'. Kaefer l
Sheet 1 of 2 '
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- 1. Birthdate - May 7,-1943 4
- 2. Citizenship - USA
! i j 3. Education - B.S. Mechanical Engineering, Purdue University,1966 l i
- 4. . Employment HistoryL- Joined PG&E in January,1971
- a. ' July,1966, to December,1970 - Active duty U.S. Navy. Commissioned from Officer Candidate School as line officer. Served in various '
shipboard assignments in submarine service.
. b. January,1971, to August,1971 - Employed by PG8E with various training assignments at Humboldt Bay Power Plant. Participated in the startup preparations for Diablo: Canyon as member of Diablo Canyon Task Force.
- c. August,1971 - Assigned to Diablo Canyon as Nuclear Engineer.
Includes an assignment to the Point Beach PWR Plant.
- 5. Nuclear Experience
- a. U.S. Navy
- 1) July,1967, to January,1968. Operator training on the Navy S5G prototype reactor (GE-PWR) at the Naval Reactor Facility, Idaho Falls, Idaho. Qualified as Engineering Watch Officer.
- 2) . March,1970, to December,1970. Assigned to the engineering department' of new construction, nuclear submarine (W-PWR), Mare - F Island Naval Shipyard. Experience in primary and secondary
. systems construction, testing, and initial core loading.
(November 1978) 13.1A-51 Amendment 73
William B. Kaefer Sheet 2 of 2 O
- b. Humboldt Bay - Participated in 1971 refueling outage as shift nuclear engineer. Experience included fuel sipping and core rearrangement.
- c. Point Beach - Assigned to Point Beach for one month during October, 1971, to observe power operation of Unit 1, review physics testing procedures, and assist in startup preparations for Unit 2.
- d. Diablo Canyon - Participated in the preparation and review of licensing material for Units 1 and 2 including the FSAR, the equip-ment description and operating manual, test procedures, and the administrative procedures and operational quality assurance manual.
Participated in the preparation of PG&E's Apprentice Control Technician training program. Participated in startup test program. ,
Responsible for SNM accountability as special Nuclear Material I l
Custodian. i
- e. Formal Training Oll
- 1) Graduate of six months officer course of Navy Nuclear Power School, Mare Island Naval Shipyard, Vallejo, California. January, 1967, to July,1967.
- 2) Graduate of six months operational reactor training at S5G prototype, NRF, Idaho Falls, Idaho. July,1967, to January,1968.
- 3) Simulator Training - Westinghouse Nuclear Training Center, Zion, Illinois. Option III (three-week course,1974) and Option II (one-week course,1978).
O Amendment 73 13.lA-52 (November 1978)
RESUME 19 NUCLEAR ENGINEER
. David B. Miklush Sheet 1 of 2
- 1. Birthdate - January 28, 1950
- 2. Citizenship - USA
- 3. Education
- a. B.S. Mechanical Engineering, UCLA,1972
- b. Registered Professional Engineer (M.E. ), California '
- 4. Employment History - Joined PGandE in February,1978.
- a. September,1972 to April,1976 - Employed by General Atomic Company - Participated in the Technical Graduate Program at General O- Atomic with three 6-month assignments in manufacturing design engi-neering, and site startup at Fort St. Vrain Nuclear Power Plant.
Permanently assigned at Fort St. Vrain from August 1974 to April 1976 in construction and operations.
- b. April,1976 to February,1978 - Employed by General Electric '
Company - Responsible Design Engineer for the BWR refueling, fuel handling, and auxiliary. service bridges. Assignment consisted of verification of vendor hardware designs and initial design of the fuel grapple for BWR 6.
- c. February,1978 to present - Employed by PGandE - Nuclear Engineer assigned to the Diablo Canyon Nuclear Power Plant.
O (November 1978) 13.1 A-53 . Amendment 73
David B. Miklush Sheet 2 of 2
- 5. Nuclear Experience
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- a. Fort St. Vrain - Participated in initial core loading; shift operations engineer during low power physics tests to 2% power.
- b. General Electric - Design of nuclear fuel handling and servicing equipment.
- c. Diablo Canyon - Nuclear Engineer engaged in procedure preparation and startup testing of various plant systems and equipment.
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Amendment 73 13.1A-54 (November 1978) l
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RESUME 20 !
NUCLEAR ENGINEER James A. Sexton Sheet 1 of 3- l 1
- 1. Birthdate - June 25, 1939
- 2. Citizenship - USA i
- 3. Education - B.S. Engineering, California State University at Humboldt, i
- 1972.
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- 4. Employment History - Joined PGandE in 1958.
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- a. 1958 to 1960 - Assigned to various conventional oil and gas fired power !
plants as operator. l
- b. 1962 to 1970 - Assigned to Humboldt Bay Power Plant as operator. ;
Progressed from Auxiliary Operator to Control Operator. Received l AEC Operator's License.
- c. 1970 to 1972 - Assigned to Humboldt Bay Power Plant as Assistant Engineer.
- d. 1973 (9 months) - Employed by Stone and Webster Engineering as Advisory Engineer on construction of James A. Fitzpatrick BWR. . !
- e. 1973 to 1976 - Employed by Burns and Roe as Senior Plant Test and Operations Engineer and Operations Test Supervisor for startup activitiu at various conventional power plants and at Three Mile i Island PWR and Washington Public Power Supply BWR.
- f. May,1977 to November,1977 - Employed by PGandE General Construction Department. Assigned to Diablo Canyon Power Plant as Startup Group Supervisor. Group responsibilities included reactor coolant, 4 condensate, feedwater, and ventilation systems.
(November'1978) 13.1A-55 Amendment 73
? . , , . . . .-
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James A. Sexton Sheet 2 of 3
- g. November,1977 to present - Transferred to PGandE Steam Department.
O Assigned to Diablo Canyon as Nuclear Engineer.
- 5. Nuclear Experience
- a. Humboldt Bay - Served as a plant operator for eight years. Started as Auxiliary Operator responsible for operation of plant auxiliary systems; including high and low voltage power distribution systems, reactor refueling operations, steam and feedwater systems operation, turbine and generator support systems operation. Progressed to Control Operator responsible for reactor and turbine generator operations. Performed all control room operations; including reactor and turbine generator power changes, reactor control rod timing,
, turbine generator trip tests and reactor refueling. Served as Assistant Engineer for two years. Responsible for taking and evaluating plant efficiency data, condenser cooling water and condenser efficiency tests, personnel radiation records, radioactive waste water inventory study.
Received AEC Operator's License.
- b. James A. Fitzpatrick Nuclear Power Plant - Directed initial nuclear vessel cold hydrostatic test; assigned startup responsibilities on reactor cleanup and radiation waste facilities.
- c. Three Mile Island Nuclear Power Plant - Responsible for originating startup procedures on nuclear plant safety and instrument systems.
- d. Washington Public Pouer Supply System BWR, Unit 2 - Directed the preparation of nuclear systems descriptions and plant startup schedule. Directed the construction testing on nuclear plant systems providing technical guidance and inspection to the constructors for Burns and Roe Incorporated.
O Amendment 73 13.lA-56 (November 1978)
James A. Sexton Sheet 3 of 3-
- e. Diablo Canyon Nuclear Powr Plant - Participated in the startup activities of nuclear systems as a group supervisor. Responsible for the origination of surveillance test procedures as Nuclear Engineer.
- f. Formal Training Courses
- 1) Participated in all formal training courses leading to AEC Operator's License at Humboldt Bay.
- 2) Seven-day training program on Wssi.inghouse PWR Simulator at Zion, Illinoi s, October,1978.
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i (November 1978) 13. l A-57 Amendment 73 l
RESUME 21 NUCLEAR ENGINEER Lawrence F. Womack,11 Sheet 1 of 2
- 1. Birthdate - October 25, 1953
- 2. Citizenship - USA
- 3. Education
- a. B.S. Physics, Stanford University,1975.
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- b. M.S. Mechanical Engineering, Stanford University,1976. I 1
l 4 Employment History - Joined PGandE in March,1978. l
- a. July,1972 to June,1976 - Employed by the Nuclear Physics Group, Stanford University Physics Department, on a half-time basis as a laboratory assistant.
- b. July,1976 to February,1978 - Employed by Westinghouse Hanford Company in the Fast Flux Test Facility (FFTF) Operations Department in preparation for startup of the FFTF. Includes a nine-week assignment at EBR-II at the Idaho National Engineering Laboratory.
- c. March,1978 - Employed by Pacific Gas and Electric Company. Assigned to Diablo Canyon Power Plant as a Nuclear Engineer.
- 5. Nuclear Experience i
- a. Education - Master's degree course work involving nuclear reactor theory, reactor design, and transport theory.
O Amendment 73 13.lA-58 (November 1978)
Lawrence F. Womack, II Sheet 2 of 2
- b. FFTF - Participated in operator's training course. Responsible for pre-sodium fill testing and operation of plant systems as shift operations engineer. Prepared operating procedures and training
, material for use by the Operations Department. Received an Operation's Engineer operating license for the FFTF.
- c. EBR-II - Assigned to EBR-II for nine weeks in April,1977 to June, I 1977. Participated in accelerated operator training program and shift reactor operations. Completed written and oral testing on ;
plant reactor and systems before departure.
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- d. Diablo Canyon - Nuclear Engineer engaged in procedure preparation and startup testing of various plant systems and equipment.
- e. Formal Training Courses
- 1) Stanford University Mechanical Engineering Curriculum in reactor theory, design, and transport theory.
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- 2) FFTF Operator Training Course, Fall 1976.
- 3) Industrial Thermometry: Fundamentals, Calibration and Time Response, short course at the University of Tennessee Department of Nuclear Engineering, Fall 1978.
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1 O l (November 1978)- 13.1A-59 Amendment 73 i
1 RESUME 23 l SENIOR CHEMICAL AND RADIATION PROTECTION ENGINEER Jerome V. Boots Sheet 1 of 2
- 1. Birthdate - June 29, 1936 l l
- 2. Citizenship - USA
- 3. Education - B.S. Chemistry, Universi:y of California,1960
- 4. Employment History - Joined PG&E in June,1957
- a. June,1957, to September,1965 - Assigned to PG&E Department of Engineering Research, various assignments involving analytical chemistry and radiochemistry. I
- b. September,1955, to July,1970 - Various assignments in chemical engineering and radiation protection at Humboldt Bay Power Plant.
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- c. July,1970, to August,1971 - Engaged at Humboldt Bay in Diablo Canyon startup preparation as member of Diablo Canyon Task Force.
Includes assignments to R. E. Ginna PWR Plant and Connecticut Yankee PWR Plant.
- d. August,1971 - Assigned to Diablo Canyon as Chemical and Radiation Protection Engineer.
- e. January,1975 - Promoted to Senior Chemical and Radiation Protection Engineer.
- 5. Nuclear Experience
- a. Department of Engineering Research - Also assigned to Hazelton Nuclear Science Corporation on an intermittent basis during 1962 through 1955 to receive on-the-job radiochemistry training. Assignments included Amendment 73 13.1A-60 (November 1978)
$ Jerome V. Boots Sheet 2 of 2 setting up low-level counting facility for processing Humboldt Bay environmental radiation surveillance samples and performing radiochemical work.
- b. Humboldt Bay - Assigned to Humboldt Bay as chemical engineer.
Supervised the chemical and radiochemical programs during plant operation. This included plant ' equipment performance evaluation,
) chemical control, radioactive waste management and training of technicians and operators. Assumed additional responsibilities of radiation protection engineer in May,1968. Supervised plant's radia-tion protection program during normal operation and refueling outages.
- c. Connecticut Yankee and Ginna Power Plants - Assigned to Connecticut Yankee for three weeks and Ginna for one week to observe rad protection and chemistry programs in 1970.
- d. Diablo Canyon - Participated in the preparation of radiation protection training material, equipment specifications, radiation control standards and procedures, emergency plans, chemical procedures, and licensing material. Supervised radiation protection and chemistry personnel in conduct of laboratory operations and training required prior to and during plant startup.
- e. Formal Training Courses
- 1) Nuclear and Radiochemistry, University of California Berkeley, 1958,
- 2) . Nuclear Physics, University of California Berkeley,1959.
- 3) Nuclear Radiation Detection, University of California Berkeley, 1962.
, 4) Refresher course in Radiological Engineering, General Electric Vallecitos Nuclear Center,1972.
L (Novembcr 1978) 13.1A-61 Amendment 73
RESUME 24 CHEMICAL AND RADIATION PROTECTION ENGINEER William A. O'haLa Sheet 1 of 2
- 1. Birthdate - June 9,1943
- 2. Citizenship - USA 3
- 3. Education - B.S. Chemistry, California State Polytechnic College,1965.
- 4. Employment History - Joined PG&E in January,1970.
- a. October,1965, to March,1966 - Employed by Hazelton Nuclear Science Corporation, Palo Alto, California, as chemist. Assigned to low-level radiochemical analysis group.
- b. March,1966, to October,1969 - Active duty U.S. Navy. Commissioned from Officer Candidate School as line officer. Served in various l
assignments aboard surface ships and as research chemist at U.S. Naval Radiological Defense Laboratory, San Francisco, California.
- c. January,1970, to October,1972 - Joined PG&E with assignment to Humboldt Bay Power Plant for training in chemical engineering and radiation protection. Included an assignment to San Onofre Nuclear Generating Station.
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- d. November,1972 - Assigned to Diablo Canyon as Chemical and Radiation j Protection Engineer.
- 5. Nuclear Experience
- a. Hazelton Nuclear Science Corporation - Assigned to low,-level radio-chemical analysis of various substances with particular emphasis on tracer analysis of sea water.
O Amendment 73 13.1A-62 (November 1978)
.__ .. .. __ _ ___ _ ..._._ _ _ _ - - . . _ _ ____ _ _ . _ . _ . . ~
William 'A. O'Hara Sheet 2 of 2 O b. Humboldt Bay Power Plant - Assigned to Humbolck Bay Power Plant for training as Chemical and Radiation Protection Engineer including the areas of plant equipment performance evaluation, chemical control, radiochemistry, radioactive waste management, radiation protection supervision, and training of operators and technicians. Assisted the Chemical. and Radiation Protection Engineer and assumed his duties during his absence. Participated in three refueling outages.
4
- c. San Onofre Nuclear Generating Station - Assigned to San Onofre for three weeks to observe chemistry and radiation protection programs.
- d. Formal Training Courses
- 1) Short course in Liquid Scintillation Counting - Series of lectures given by the International Conference on Organic Scintillators and Liquid Scintillation Counting at the University-O of' California Medical Center, San Francisco.
- 2) Diablo Canyon Design Lecture Series - Series of lectures given by designers of Diablo Canyon systems and equipment, Westinghouse APD, Winter,1971.
- 3) Refresher Course in Radiological Engineering, General Electric Corporation Vallecitos Nuclear Center,1972.
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'(November 1978) 13. l A-63 Amendment 73
RESUME 25 RADIATION AND PROCESS MONITOR FOREMAN George Paul M. Lyon Sheet 1 of 2
- 1. Birthdate - April 12, 1949
- 2. Citizenship - USA 4
- 3. Education - B.S. Chemical Engineering, Brigham Young Uni' sity, 1976.
- 4. Employment History - Joined PGandE in January,1977.
- a. June to August 1970,1971,1972 and 1973 - Employed by San Diego Gas and Electric Company as a summer student engineer. Assigned to the central laboratory and engineering groups.
- b. September 1973 to May,1974 - Employed by Brigham Young University.
Assigned to various positions including assistant Bio Chem stock-1 room mnager, Organic laboratory teaching assistant, and chemical research assistant.
- c. January,1977 - Joined PGandE with assignment to the Geysers Power Project for the installation of the Hydrogen Sulfide abatement system.
- d. August,1977 - Assigned to Diablo Canyon as Radiation and Process Monitor Foreman. )
O Amendment 73 13.lA-64 (November 1978)
j l
George Paul M. Lyon Sheet 2 of 2 s 1
- 5. Nuclear Experience l
- a. Diablo Canyon Power Plant - Assigned to Diablo Canyon Power Plant j for training as Radiation and Process Monitor Foreman including j the areas of plant equipment, chemical control, radiochemistry, l radioactive waste management, radiation protection supervision, i laboratory supervision, and training of technicians. Assist the i Chemical and Radiation Protection Engineer and partially assume i his duties during his absence. j
- b. Formal Training Courses
- 1) Lecture series on Radiation Protection,1977.
- 2) Lecture series on Introduction to Nuclear Power,1978.
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(November 1978) 13.lA-65 Amendment 73
RESUME 26 SENIOR INSTRUMENT AND CONTROL SUPERVISOR Kenneth C. Doss Sheet 1 of 3
- 1. Birthdate - August 16, 1932
- 2. Citizenship - USA
- 3. Education - A.S. Electronics, Cuesta College,1969
- 4. Employment History - Joined PG&E in March,1952
- a. March,1952, to May,1955 - Assigned to Electric Transmission and Distribution Department as member of line crew,
- b. May,1955, to May,1961 - Transferred to Steam-Electric Generation Department with assignment to Morro Bay Power Plant as Instrument Repairman. Participated in the startup of Morro Bay Power Plant Units 1 and 2.
- c. May,1961, to July,1970 - Additional assignments at Morro Bay Power Plant as Test Engineer and subsequently as Instrument Maintenance Foreman. Participated in startup of Morro Bay Units 3 and 4 including prestartup check of plant control systems. Special assignment of revising the Company Apprentice Instrument Repairman training program.
- d. July,1970, to August,1971 - Engaged at Humboldt Bay in Diablo Canyon startup preparation work as a member of the Diablo Canyon Task Force.
l Special assignment to task force for revising the Company Apprentice Control Technician training program.
- e. August,1971 - Assigned to Diablo Canyon as Instrument and Control l Supervisor.
O I Amendment 73 13. l A-66 (November 1978) l
Kenneth C. Doss Sheet 2 of 3
- f. September,1977 - Promoted to Senior Instrument and Control 4 Supervisor.
- 5. Nuclear Experience
- a. Humboldt Bay - Training assignment at Humboldt Bay to observe various instrumentation procedures and overall plant operation.
Participated in the 1971 refueling operation, as a member of the Diablo Canyon Task Force.
- b. Diablo Canyon - Participated in the preparation of training material for operators and technicians, including description and instructions for control systems, nuclear instrumentation and com-puters. Also participated in specifying test equipment and spare parts supplies for all instrument and control systems. Established program and procedures for test equipment calibration, preventative maintenance, and surveillance testing. J l
- c. Formal Training Courses
- 1) P-250 Computer Maintenance School, Westinghouse APD, sixteen weeks, 1970.
- 2) P-2000 Computer Maintenance School, Westinghouse APD, two weeks, 1970. .
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- 3) Nuclear Analog Control Systems Training School, Westinghouse APD, '
two weeks,1972.
- 4) Instrumentation and Control Training School, Westinghouse APD, twelve weeks, 1972, p 5) Introduction to Nuclear Power, Humboldt Bay Power Plant and Diablo Canyon Power Plant.
(November 1978) 13.lA-67 Amendment 73
Kenneth C. Doss Sheet 3 of 3
- 6) Radiation Protection Course, Humboldt Bay Power Plant and 0
Diablo Canyon Power Plant.
- 7) Eberline Radiation Pratection Equipment Maintenance, one week, 1974.
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- 8) Atomic International Vibration and Loose Parts Monitor Maintenance, two days,1976.
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Amendment 73 13.lA-68 (November 1978)
i RESUME 28 i INSTRUMENT MAINTENANCE FOREMAN )
Mark W. Stephens Sheet 1 of 2
- 1. Birthdate - November 14, 1939 1
- 2. Citizenship - USA
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- 3. Education - High School
- 4. Employment History - Joined PGandE in July,1959 l
- a. U.S. Coast Guard, 1959 - 1959.
- b. July,1959 to August,1960 - Assigned to PGandE Potrero Power Plant and Hunters Point Power Plant as a Mechanical Maintenance Department Helper.
- c. August,1960 to May,1962 - Assigned to Humboldt Bay Power Plant as an Apprentice Machinist.
- d. May,1962 to April,1967 - Assigned to Humboldt Bay Power Plant as an Apprentice Control Technician. Promoted to Control Technician, July, 1964.
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.i e. April,1967 to October,1971 - Assigned to Potrero Power Plant and Hunters Point Power Plant as a Control Technician,
- f. October,1971 - Assignment to Diablo Canyon Power Plant as a Control Technician.
- g. April, 1976 - Promoted to Instrument Maintenance Foreman.
O (November 1978) 13.1A-69 Amendment 73
._ ._. - - _ _ .J
fiark W. Stephen Sheet 2 of 2 O
- 5. Nuclear Experience
- a. Humboldt Bay - Assigned to protection systems testinn, control and indication systems maintenance and calibration. ,
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- b. Diablo Canyon - Assigned to construction inspection, specifying l control and protection systems spare parts, calibration and maintenance of control and protection systems, and preparation of surveillance test and maintenance procedures.
- c. Formal Training Courses
- 1) Introduction to Nuclear Power, Humboldt Bay Power Plant,1962.
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- 2) Radiation Protection Course, Humboldt Bay and Diablo Canyon i Power Plants.
- 3) P-250 Computer Maintenance School, Westinghouse APD, sixteen weeks, 1970.
- 4) P-2000 Computer Maintenance School, Westinghouse APD, two weeks, 1970.
- 5) Instrumentation and Controls Training School, Westinghouse APD, twelve weeks, 1972.
O; Amendment 73 13.1A-70 (November 1978) ]
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'(November 1978) 13.1A-71 Amendment 73 1
RESUME 30 MECHANICAL FOREMAN Doyle L. Abney Sheet 1 of 2
- 1. Birthdate - April 6,1937
- 2. Citizenship - USA
- 3. Education
- a. High School
- b. Machinist Trade School
- c. USAF Apprentice Machinist Program
- 4. Employment History - Joined PGandE in May,1968.
- a. January,1955 to January,1962 - Military service with USAF.
- b. January,1962 to March,1964 - Machinist employed by Elders Industries.
- c. March,1964 to May,1968 - Machinist employed by Bakersfield Machine Co.
- d. May,1968 to January,1977 - Employed by PGandE as Machinist at various conventional oil and gas fired power plants.
- e. January,1977 - Transferred to Diablo Canyon Power Plant as Mechnical Foreman.
O (November 1978) 13.1 A-72 Amendment 73
Doyle L. Abney Sheet 2 of 2 e_s O
- 5. Nuclear Experience
- a. Diablo Canyon - Engaged in startup preparation work, including supervising initial implemantation of preventative maintenance program and turbine reblade job.
- b. Humboldt Bay - Assigned to Humboldt Bay Unit 3 for five weeks in 1969 to participate in turbine overhaul,
- c. Assigned to Trojan Nuclear Power Plant for one week in May, 1978 to observe maintenance practices.
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- 6. Formal .t 'ning Courses l
- a. Radiation Protection Training Course - Diablo Canyon Power Plant.
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Amendment 73 13.lA-73 (November 1978) i
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O (November 1978) 13.lA-74 Amendment 73 4
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RESUME 31 MECHANICAL FOREMAN William Robert Ryan
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Sheet 1 of 2
- 1. Birthdate - October 3, 1939
- 2. Citizenship - USA
- 3. Education - High School
- 4. Employment History - Joined PGandE in May,1959
- a. May,1959 to September,1959 - Employed by PGandE in power plant operations.
- b. September,1959 to April,1965 - Employed by PGandE in power plant maintenance. Successfully completed apprentice machinist program.
A V c. April, 1965 to March, 1974 - Assigned to Moss Landing Power Plant as a Traveling Machinist. Work assignments included overhauls at Morro Bay, Potrero, Humboldt Bay, and the Geysers. Participated in the Moss Landing Unit No. 6 and No. 7 startup.
- d. January,1971 to March,1974 - Various assignments as temporary Mechanical Foreman.
- e. March,1974 - Assigned to Diablo Canyon as Mechanical Foreman.
- 5. Nuclear Experience
- a. Humboldt Bay - Temporary assignment as Traveling Machinist.
- b. Surry - One-week assignment to Surry kuclear Station to observe maintenance procedures and practices.
O Amendment 73 13.1A-75 (November 1978)
William Robert Ryan Sheet 2 of 2 O
- c. Formal Training Courses
- 1) Diesel Maintenance - One-week class taught by Alco Diesel, October,1976.
- 2) Radiation Protection Training Course - Diablo Canyon Power Plant.
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(November 1978) 13.lA-76 Amendment 73
RESUME 32 ELECTRICAL FOREMAN Gerald M. Zocher Sheet 1 of 2
- 1. Birthdate - flovember 28, 1928
- 2. Citizenship - USA
- 3. Education - High School
- 4. Employment History - Joined PGandE Company in September,1950.
- a. September, 1946 to July, 1949 - U.S. Air Force; served as crew chief constructing and maintaining communications facilities.
- b. September,1950 to July,1961 - Assignments in mechanical and electrical maintenance of various conventional power plants as an p electrician.
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- c. July, 1961 to February, 1964 - Assigned to Humboldt Bay as an rlectrician and six months as temporary Electrical Foreman.
- d. February,1964 to July,1973 - Assignments in electrical maintenance of various conventional power plants as an electrician.
- e. August,1973 - Assigned to Diablo Canyon as Electrical Foreman.
- 5. Nuclear Experience
- a. Humboldt Bay - Performed electrical tests and maintenance of all electrical equipment related to Unit 3 since initial power operation.
- b. Zion - Assigned two weeks in April, 1974 to observe maintenance l
p practices and procedures, j V !
Amendment 73 13.1A-77 (flovember 1978) l
Gerald ti. Zocher Sheet 2 of 2 O
- c. Diablo Canyon - Participated in procurement of spare parts, writing procedures for maintsr,;nce of electrical equipment and the layout of electric shop and i 9 iated equipment.
- d. Formal Training Courses
- 1) Radiation Protection Training Course, Humboldt Bay Power Plant,1962.
- 2) Radiation Protection Training Course, Diablo Canyon Power Plant.
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(November 1978) 13.1A-78 Amendment 73
RESUME 33 MAINTENANCE ENGINEER Robert M. Nanninga Sheet 1 of 2
- 1. Birthdate - February 28, 1947
- 2. Citizenship - USA
- 3. Education
, a. B.S. Mechanical Engineering, University of New Mexico,1970.
- b. New Mexico EIT Certificate #983,1970.
- 4. Employment History - Joined PGandE in July,1970.
- a. July,1970 to October,1970 - Field Engineer, PGandE General p Construction Department, Diablo Canyon Plant - Mechanical Department Piping Inspector.
- b. October,1970 to February,1971 - Assigned to Design Drafting Department, San Francisco General Office. Wrote computer programs and checked instrumentation schematics for Diablo Canyon.
- c. February,1971 to April,1972 - Assigned to Design Drafting Department, San Francisco General Office. Designed pipe layouts and check drawings for Turbine Building and Containment Structure, Diablo Canyon Plant,
- d. April,1972 to July,1972 - Assigned to General Construction flechanical Piping Group at Diablo Canyon. Inspected pipe installation in Unit 1 Turbine Building.
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Amendment 73 13.1A-79 (November 1978)
Robert M. Nanninga Sheet 2 of 2 0;
- e. July,1972 to December,1972 - Assigned to Pismo Beach Railroad Siding Receival Yard. Inspected, performed surveillance of and supervised maintenance of equipment in storage. Coordinated ship-ment of material to Diablo Canyon jobsite.
- f. December,1972 to September,1978 - Assigned to Miscellaneous Mechanical Group. Coordinated equipnent installation with con-tractors, and prepared construction and repair procedures.
Experience primarily involved with conventional plant type mechani-cal equipment with the exception of nuclear fuel handling system.
- g. September,1978 - Assigned to Steam Generation Department as Maintenance Engineer.
- 5. Nuclear Experience
- a. Diablo Canyon - 1970 to 1978 - Field Engineer for General Construc-tion Department, Mechanical Group. On loan to Engineering and Design Draf ting Departments from October 1970 to April,1972, working on Diablo Canyon. Performed assignments in piping, equipment receival, miscellaneous mechanical equipment installation, maintenance and operational troubleshooting while assigned to jobsite.
- b. Formal Training Courses
- 1) Convair School for Nondestructive Testing, Diablo Canyon Power Plant. Certified as a level II NDT inspector in radiography, ultrasonics, magnetic particle and liquid penetrant testing, October,1970.
- 2) Radiation Protection Training Course, Diablo Canyon Power Plant, July, 1976.
O 13.1 A-80 Amendment 73 (November.1978)
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RESuttE 34 l
QUALITY CONTROL SUPERVISOR Juanito S. Diamonon ;
Sheet 1 of 3
- 1. Birthdate - November 23, 1944
- 2. Citizenship - USA
- 3. Education
- a. B.S. Electrical Engineering, Mapua Institute of Technology, lianila, 1965.
- b. Registered Professional Engineer (E.E.), California.
- 4. Employment History - Joined PGandE in June,1970.
,- a. November 1965 to May 1970 - Employed by Manila Electric Company
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initially as a Cadet Engineer (engineering trainee) for a 2-year training assignment in verious company departments like Operations (assigned to Tegen Power Plant Units 1 and 2,110 f@l. each), Load Dispatching, and Engineering. Participated in the startup of 165 fM Gardner Power Plant Unit 1 and 220 t61 Gardner Unit 2. Subsequently assigned to Electrical Engineering Department as a Design Engineer for conventional power plant and substation projects. Projects undertaken include 330 fM Snyder Power Plant Un t 1, 216 fiVA North i
Port and 133 MVA Malibay Transmission / Distribution Substations,
- b. June 1970 to August 1974 - Immigrated to the United States in June, 1970. Employed by PGandE as an Engineering Designer in Engineering Department's Design Drafting and was assigned to the Diablo Canyon Project. Worked with Responsible Engineeers in the design of the electrical and instrumentation portions of the Nuclear Steam Supply Systems, Condensate and Feedwater System, Turbine-Generator and Auxiliary Systems, and Onsite Electrical Distribution System.
Amendment 73 13.1A-81 (November 1978)
Juanito S. Diamonon Sheet 2 of 3
- c. October 1974 to March 1977 - Assigned to Diablo Canyon Power Plant as member of the onsite Quality Assurance Department's Operations group. As Quality Assurance Engineer, performed audits on pre-operational tests conducted by General Construction's Startup group to verify compliance with the FSAR, Quality Assurance Procedure PRC-10, NRC Regulatory Guide 1.68, and other quality requirements.
Also performed audits on the Plant Staff to verify implementation of the Quality Assurance Program for Operations.
- d. April 1977 - Assigned to the Plant Staff as Quality Control Super-visor.
- 5. Nuclear Experience
- a. Diablo Canyon - Participated in the design, preoperational testing, and quality assurance programs.
- b. Radiation Protection Training Course, Diablo Canyon Power Plant, 1974.
- 6. Formal Training Courses
- a. Quality Assurance Principles, Practices and Techniques for the Nuclear Power Industry, L. Marvin Johnson and Associates, Inc., May, 1975.
- b. Quality Assurance Audit Techniques for the Nuclear Power Industry, L. Marvin Johnson and Associatas, Inc., March, 1975.
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- c. Quality Assurance Management Workshop, L. Marvin Johnson and Asso- l ciates, Inc., July, 1975. l l
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(November 1978) 13.1A-82 Amendment 73 l
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p Jua_nito S. Diamonon Sheet 3 of 3 i
- d. Seminar on 10CFR50, Appendix B, Frank H. Squires Consultants, April, 1975.
- e. Short Course in Metallurgy, Fabrication and Nondestructive Exam-ination of Welded Code Pressure Vessels, California Polytechnic State University, July, 1976.
- f. Short Course in Power System Protection, California Polytechnic j State University, May,1976.
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9 Industrial Power System (838A and 838B), University of California Berkeley Extension, Fall,1971 and Spring,1972,
- h. Practical Approach to Power System Protective Relaying 843, University of California Berkeley Extension, Fall,1972.
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Amendment 73 13.lA-83 (flovember 1978)
RESUtiE 36 SECURITY SUPERVISOR Bryan A. Dettman Sheet 1 of 2
- 1. Birthdate - February 20, 1943
- 2. Citizenship - USA
- 3. Education
- a. B.A. Police Science & Administration, Sacramento State University,1971.
- b. M.A. Public Administration, Golden Gate University,1975.
- c. Advanced Certificate, State of California, Department of Justice, Commission of Peace Officer Standards and Training.
- d. State of California Lifetime Vocational Education Teaching Certificate.
- e. State of California, Department of Consumer Affairs, Certified " Laws of Arrest" and "Fireanns" Instructor.
- 4. Employment History - Joined PG&E in August, 1975.
- a. 1964 to 1971 - City of Walnut Creek, California, Police Department.
Employed as a police officer. During employment rose to final position as Director of Investigative Division. While employed, also served as a detective and as a sergeant.
- b. 1971 to 1975 - Bay Area Rapid Transit District, Police Services.
Employed as Police Lieutenant and Bureau Commander. While employed, was in charge of the following operations at various times.
- 1) Administrative Bureau Commander (set up liaison and agreements with 18 local law enforcement agencies and designed and implemented hiring program to staff entire Police Service Department.)
O (November 1978) 13.lA-84 Amendment 73
Bryan A. Dettman Sheet 2 of 2
- 2) Technical Services Bureau Commander (Bureau comprised of Detective Division, Revenue Protection Division - i.e.,
armored trucks and a large cash facility, Communications Division.)
- 3) Special Projects Coordinator - Responsible for obtaining Federal grants and for writing and attempting to get legislative passage of new State laws pertaining to the BART Police Services Department.
- 4) Evening Patrol Division Commander - Responsible for all police activities on the BART system (3 p.m. - 6 a.m.)
- c. August 1974 - Consultant to Bechtel Corporation on security at Honolulu International Airport.
- d. 1975 to Present - Pacific Gas and Electric Company - Currently Os responsible for security program at Diablo Canyon Power Plant. Wrote 1 and implemented the plant's Security Plan. ;
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During employment, was also responsible for auditing and upgrading security program at Company's Humboldt Bay Nuclear Power Plant.
Assisted in re-writing of Humboldt Bay Security Plan.
Additionally, during course of employment have visited and inspected security programs at seven other nuclear facilities.
O Amendment 73 13.lA-85 (November 1978)
RESUME 37 SECURITY SHIFT SUPERVISOR Larry C. Fisher Sheet 1 of 3
- 1. Birthdate - July 15, 1946
- 2. Citizenship - USA
- 3. Education
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- a. AA Police Science & Public Administration, Sacramento City College, 1974.
- b. BA Criminal Justice, California State University, Sacramento,1978.
- c. State of California, Department of Consumer Affairs, certified
" Exercising Power of Arrest" and " Firearms Training" instructor.
- d. USAF Technical School, " Electronics Intercept Operations / Analyst Specialist, Keesler AFB,1966.
- e. USAF Technical School, " Avionics Communications Repair Specialist, ,
Keesler AFB, 1975.
- 4. Employment History - Joined PGandE in May, 1978
- a. 1966 to 1970 - United States Air Force. Electronic Intercept Operator / Analyst Specialist. Collected and analyzed electro-magnetic intercept information to determine radar types by comparing parameters and plotting aircraf t relative bearings to determine location. Produced electronics intelligence, mission data, to be released to command and other consumers.
Assisted in updating the Electronic Order of Battle (E0B) data base. Obtained rank of sergeant and supervised 5 to 15 personnel.
- b. 1971 to 1973 - United States Air Force Reserve. Trained as and ;
functioned as a Security and Law Enforcement Officer. Held rank of Sergeant and supervised 3 to 5 personnel.
(November 1978) 13.1A-86 Amendment 73
4
> Larry C. Fishe_r_
Sheet 2 of 3 O c. 1975 to present - United States Air Force Reserve
- 1) Trained as and functioned as an Airborne Radio Communications Repair Technician. Worked on communications equipment assoc-iated with the C-130 (Hercules) and KC-135 (Tanker) aircraft.
- 2) Trained as and functioned as an Avionics Job Controller.
Dispatched avionics maintenance personnel to aircraft (s) for work discrepancies in such a manner that time and cost effi-j ciency were at a maximum.
- d. 1971 to 1975 - Hargrave Security Services, Sacramento, California During employment, worked as Security Officer at various industrial, i commercial, and private locations. Worked as Security Officer at
. Greyhound Bus Station, Sacramento, CA, and acted as liaison between
- Greyhound, Inc. and local law enforcement. Rose to final position as Operations Supervisor and Field Representative.
- 1) Operations Supervisor - Supervised 25 to 160 personnel; responsible for hiring, training, posting, scheduling, payroll, i and client billing. Maintained monthly and annual statistics. ,
- 2) Field Representative - Conducted security surveys (County Fairs, industrial and retail businesses, and individual residences.)
e, 1975 to 1977 - Vanguard Security Services, Sacramento, CA During employment, worked at Rancho Seco Nuclear Power Plant as a Security Officer. Rose to rank of Sergeant and supervised a shift consisting of 8 to 10 personnel.
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- f. July 1977 to May 1978. Sacramento Municipal Utility District, Sacramento, CA. Employed as a Watch Commander at Rancho Seco Nuclear Power Plant. During employment, rose to Special Agent position tempo- I rarily for six months. Responsibilities included supervision of five Watch Commanders and the contract security force. Served as liaison l between Security and Operations and the contract security force and O SMUD.
Amendment 73 13.lA-87 (November 1978)
l.arry C. Fisher Sheet 3 of 3
- g. tiay 1978 to present - Pacific Gas and Electric Company 9
Currently responsible for supervision of contract security force personnel - supervising the operation training, and development of procedures relating to operational emergencies, escorts, and security records. Wrote Contingency Plan.
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, 1 l (flovember 1978) 13 lA-88 Amendment 73
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RESUME 38 SECURITY SHIFT SUPERVISOR Larry G. Lunsford Sheet 1 of 2
- 1. Birthdate - August 27, 1945 )
- 2. Citizenship - USA
- 3. Education I
- a. AS degree - Police Science, Cuesta Community College - 1972.
- b. Advanced Certificate, State of California, Department of 1
Justice, Commission of Peace Officer Standards and Training. l i
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- c. Supervisory & Management Certificates, State of California, i l
Department of Justice, Commission of Peace Officer Standards i and Training.
- d. State of California Vocational Education Teaching Certificate.
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- e. State of California, Department of Consumer Affairs, certified j
" Laws of Arrest" and " Firearms" Instructor. !
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- f. Federal Bureau of Investigation, certified " Chemical Agent" instructor's certificate.
- g. Graduate - Federal Bureau of Investigation National Academy, Washington, D.C. ,1977.
1
- h. Federal Bureau of Investigation " Hostage Negotiater" certificate, 1975. l
- 1. "Special Tactics and Weapons" Commander for San Luis Obispo l I
Police Department, 1973-1978.
l j, State of California certificates in Terrorist Activities, Officer l Survival, and Civil Disorder.
- k. Federal Bureau of Investigation," Anti-Sniper" course certificate, Quantico, Virginia.
' Amendment 73 13.1A-89 (November 1978)
Larry G. Lunsford Sheet 2 of 2
- 1. State of California, Department of Justice, certificate in O
" Executive Protection."
- 4. Employment History - Joined PGandE in April, 1978.
- a. 1966 to 1978 - City of San Luis Obispo Police Department .
Employed as a Police Officer. During employment, rose to final position of Chief of Detectives, holding rank of lieutenant.
While emoloyed in capacity of Police Officer, special assignment includes:
- 1) Wrote, directeo, and acted in two police training films seen throughout the United States.
- 2) Authored an article for a professional magazine dealing with officer survival.
- 3) Was consultant to 16 law enforcement agencies throughout the ,
United States dealing with Special Weapons and Tactics. l l
- 4) Served as Watch Commander on all three shif ts. l April 1978 to present - Employed by Pacific Gas and Electric Company in the capacity of Security Shift Supervisor. Specifically in charge of security training.
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(November 1978) 13.lA-90 Amendment 73 i
RESUME 39 SECURITY SHIFT SUPERVISOR O- Vernon F. Sciocchetti Sheet 1 of 2
- 1. Birthdate - July 31, 1923
- 2. Citizenship - USA
- 3. Education
- a. U.S. Naval School of Hospital Administration, Bethesda, Marlyland,1953. <
- b. Standard First Aid & Personal Safety Instructor, San Luis Obispo, CA, 1978.
- c. State of California, Departnent of Consumer Affairs, certified as " Laws of Arrest" and " Firearms" instructor.
- 4. Employment History - Joined Pacific Gas and Electric Company March,1978.
- a. 1941 to 1960 - U.S. Navy. During employment rose to rank of Chief Petty Of ficer. Held various supervisory positions performing medical administrative functions while serving the Fourteenth Naval District Headquarters, Pearl Harbor, during the last seven years of ,
service. !
- b. 1960 to 1976 - Various positions in the fields of hospital supply and administration and building contractor.
- c. 1976 to 1978 - Pinkerton's Incorporated. During employment, rose to rank and position of Training Seargeant at Diablo Canyon Power Plant.
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Amendment 73 13. l A-91 (November 1978) i l
Vernon F. Sciocchetti Sheet 2 of 2 O
- d. March 1978 to present - Pacific Gas and Electric Company, Diablo Canyon Power Plant. Responsible for the supervision of the contract security force and development of procedures for officer performance ratings, package and material control, lock and key control, and security during fuel loading. Provided classroom instruction in Report Writing and Standard First Aid and Personal Safety.
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O (November 1978) 13.lA-92 Anendment 73 1
RESUM'. 40 SHIFT SECURITY SUPERVISOR Robert R. Smith Sheet 1 of 2
- 1. Birthdate -' August 10, 1921 i
- 2. Citizenship - USA
- 3. Education
- a. High school.
1
- b. U. S. Navy instructors school.
- c. Military police training.
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- d. Military and private security force weapons courses in rifle and handguns. '
- e. Computer operations courses at Lockheed,
- f. State of California, Department of Consumer Affairs certified " Exercising Powers of Arrest" and " Firearms Training" instructor.
- 4. Employment History - Joined PGandE in April,1978.
f
- a. 1940 to 1961 - United States Navy. Enlisted as a Seaman Recruit and progressed up through the ranks through competitive examinations to the rating of Chief Petty Officer in the field of Foundry Practices.
Supervised up to 145 men at one time and taught school for 1-1/2 years.
- b. 1962 to 1974 - Lockheed Missiles and Space Company, Sunnyvale, CA.
Worked for 2 years in the Industrial Security Depa: oment as a Secu' Officer with a secret security clearance granted by the U. S. Navy 6..a U. S. Air Force. For 10 years, assignment involved operation of large O scale IBM computers.
Amendment 73 13.lA-93 (November 1978)
/
Robert R. Smit Sheet 2 of 2
- c. 1975 to 1978 - Pinkertons, Inc. Employed as a Security Guard at the i
Diablo Canyon Nuclear Power Plant. I was a Security Guard for 2 years and was then promoted to the position of Captain. This position put me in charge of the Contract Security Force at Diablo Canyon for a period l of approximately one year, after which I joined the PGandE Security Departrent.
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- d. April 1978 to present - Pacific Gas and Electric Company, Diablo Canyon Power Plant. Employed as Security Shift Supervisor engaged in preparation of security procedures.
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(November 1978) 13.lA-94 Amendment 73
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RESUME 41 SECURITY SHIFT SUPERVISOR Ronald G. Todaro Sheet 1 of 2 l
- 1. Birthdate - April 26, 1949
- 2. Citizenship - USA
- 3. Education !
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- a. Completed most of the work required for a BS - Administration of Criminal Justice, San Jose State University,
- b. AA Police Science, Chabot College,1969.
- c. Intermediate Certificate, State of California, Department of Justice, j Commission of Peace Officer Standards and Training. )
- d. Bomb and Explosive Training; Alcohol, Tobacco, and Firearms Division, U. S. Treasury Department.
- e. State of California, Department of Consumer Affairs, certified " Laws j of Arrest and Firearms" instructor.
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- f. Hostage Negotiation Training, San Jose State University. l a
- 4. Employment History - Joined PGandE in March,1978.
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- a. 1967 to 1972 - City of Berkeley, California, Police Department. Employed 4
as an aide and trainee,1967 to 1970. Employed as a police officer, 1970 to 1972.
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I Amendment 73 13.lA-95 (November 1978)
Ronald G. Todaro Sheet 2 of 2
- b. 1972 to 1978 - Bay Area Rapid Transit District, Police Services.
Employed as a police of ficer.
Performed in the following capacities:
- 1) Patrolman - Performed all law enforcement functions in the BART District, conducted initial investigation of incidents, testified in court as required by investigations.
- 2) Field Training Officer - Supervised, trained, and evaluated new officers.
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- 3) Detective - Conducted follow-up investigation on najor cases, prepared cases for court, obtained criminal complaints from District Attorney's Officers and maintained liaison with other law enforcement agencies.
- c. 1978 to present - Pacific Gas and Electric Company, Diablo Canyon Power Plant. Responsible for the supervision of security force personnel -
supervising the operation, training, and development of procedures relating to alarm systems, closed circuit TV survaillance systems, and access control systems, and procedure development for all phases of the security system.
O 13.1A-96 Amendment 73 (November 1978)
13.2 TRAINING PROGRAM O I3.2.1 PROGRN1 DESCRIPTION The experfence obtained by the Company since 1956 in training personnel for operation and maintenance of nuclear power plants has enabled it to clearly define the training requirements for each position in the plant i
organization and to evaluate the various means of obtaining this training.
Based upon this experience, the Company has chosen to utilize a combination of formal classroom training and on-the-job experience at operating nuclear plants to achieve its training goals. A brief summary of each of the
- various training activities is given in Table 13.2-1. The extent to which each individual has parcicipated in these activities, or will participate prior to initial loading, has been individually determined based upon the persons position in the plant organization and his previous experience.
i Training for Plant Supervisorial Personnel i
The training programs for the initial appointees to supervisorial positions in the Diablo Canyon operating organization are summarized in the resumes.
1 In addition, the schedule of formal nuclear training designed to prepare candidates for NRC Operator and Senior Operstor License examinations is shown in Figure 13.2-1 in relation to the st hedule for preoperational testing and initial fuel loading.
As indicated in the resumes included in Appendix A, the majority of the supervisorial personnel have had several years of nuclear power plant experience at llumboldt Bay Power Plant or at other nuclear facilities.
Thus, the initial training for this group was largely concentrated in two major areas: 1) becoming familiar with the differences between the PWR and BWR concepts, and 2) study of the design and operation of the Diablo Canyon plant itself.
O (November 1978) 13.2-1 Amendment 73
The first formal involvement by a member of the plant staff in the Diablo Canyon project occurred in early 1968, when the Power Plant Engineer was assigned to the Company's General Office for approximately two months to assist in the preparation of the Unit 1 PSAR. In the spring of 1969, both the Power Plant Engineer and the Supervisor of Operations were engaged in similar work on the Unit 2 PSAR for approximately six months, and also assisted in the conceptual design of several plant systems. These two individuals were then assigned to the R. E. Ginna plant in the latter part of 1969 to participate in its startup testing program. Since that time, other key plant supervisory personnel have been sent to various PWR plants which were in operation or in the startup testing program. The majority of these assignements took place in the period from 1970 to 1972.
In July 1970, the second major step in the early supervisorial training activities occurred with the organization of the Diablo Canyon Task Force at llumboldt Bay. This was the first time that the najority of the supervisorial staff was assembled as a group. Initially, the Task Force consisted of 14 individuals on a full or part time basis. During the period that the group was at Humboldt Bay, work was begun on the various Task Force assignments, including preparation of training material, operating manuals, licensing material and technical specifications, and performing an operational review of the plant design.
In August 1971 the Task Force was transferred to the site, along with several supervisors who had not previously been on the Task Force, in order to obtain maximum participation of plant staff personnel in on-site activities, including observation of equipment installation and review and comment on the system and equipment preoperational and startup test procedures prepared by the General Construction Department.
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i Amendment 73 13.2-2 (November 1978)
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4 The secon.1 basic phase of the training program began with the arrival
, on-site of selected plant operators approximately 5 years prior to initial core loading. At this time, the formal nuclear training courses required to prepare individuals for the operator license examinations began. These courses were conducted by members of the plant supervisorial staff.
> Depending upon a particular individual's experience and qualifications, supervisors were either instructors or participants as appropriate during different portions of this program. It should be noted that, in most cases, the supervisors participating in these courses have coupleted similar training at Humboldt Bay Power Plant.
In addition to participation in the formal training programs, the supervisors will be actively engaged in preoperational testing and checkout of systens and eqaipment, hot functional testing, initial loading and low level testing, and in the power escalation program leading to commercial operation, as these activities take place.
Training for Plant Physical Force Personnel The training assignments for ' physical force personnel are discussed below.
- in addition, the schedule of formal nuclear training for NRC operator license examination candidates is shown in Figure 13.2-1 in relation to the schedule for preoperational testing and initial fuel loading.
All physical force personnel will be trained in radiation protection, quality assurance and security procedures and practices to an extent commensurate with their duties. In addition, the Radiation and Process Monitors and Control Technicians.may participate in the nuclear technology and plant design seminars. The Radiation and Process Monitors will also be trained in chemical and radio-chemical techniques.
(November 1978) 13.2-3 Amendment 73
It should be noted that the assignments discussed above are the special I assignments designed to qualify personnel for work at Diablo Canyon. In addition, most physical force personnel receive additional training of a more general nature under standard Company training programs, as determined by their particular classification. For example, Control Technicians and ,
maintenance personnel undergo apprenticeship programs lasting f rom 24 to 48 months before qualifying for journeyman status. Sinflarly, the Auxiliary I 1
Operators must successf ully complete a 30-month program on general power j plant technology in addition to their formal nuclear training, before they )
are eligible for promotion to Assistant Control Operator. In addition ,
some of the successful bidders for positions in the initial Diablo Canyon l organization have had previous nuclear experience at llumbold t Bay.
l The first physical force personnel assigned to the site were the Control Technicians, three of whom arrived in late 1971. They were placed on loan to the General Construction instrument staff and participated in the installation and checkout of plant equipment. An Apprentice Control Technician, who completed his apprenticeship in 1973, was transferred to the site in December 1972, and was also placed on loan to General Construction.
l The Senior Control Operators and Control Operators for Unit I were assigned to the site approximately 5 years prior to initial core loading, and began their formal training at that time. The remaining operators were assigned as the preoperational testing work load dictates. All operators will receive. extensive on-the-job training in the operation of plant controls during the preoperational and startup testing programs.
I Other physical force personnel were assigned to the site as dictated by the work load and in time to complete any training required prior to work l
assignment.
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Amendment 73 13.2-4 (Novembe r 1978) l
Other Training In addition to the previously mentioned training, plant personnel are required to routinely participate in a program of lectures, demonstrations, written assignments, and drills designed to familiarize them with fire protection procedures, security procedures, medical and first aid techniques, radiation protection principles, and their actions in the event <
of a plant emergency. The extent of the training that a particular individual receives is dependent upon the responsibilities of his or her position on the plant staff. i l
l Coordination and Administration l The Plant Superintendent is responsible for the overall training program for plant personnel, and as such he is responsible for coordinating and j administering the training program. As appropriate the Plant Superintendent delegates responsibility and authority for conducting any or l
all parts of the program to various supervisors on the Plant staff.
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i The Supervisor of Operations, the Power Plant Engineer, and the Supervisor j of ttaintenance are responsible to the Plant Superintendent for maintaining ;
ll an adequate number of trained personnel in their areas of responsibility. ,
9 The Training Coordinator and his assistant are responsible for coordinating t
the various programs and for maintenance of adequate training records, i
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- (November 1978) 13.2-5 Amendment 73
13.2.2 RETRAINING PROGRAM An operator retraining program meeting; the requirements of Appendix A to e
10 CFR 55 will be placed in ef fect wihin three months after issuance of the Unit 1 operating license.
Program Description Schedule The requalification program shall be conducted on an annual basis and upon its completion shall be promptly followed by successive programs.
Training Sessions The requalification program may include preplanned lectures, self study assignments and audio-visual presentations. A minimum of six preplanned lectures shall be presented during each annual program. The following subject material will be covered during each period.
- a. Theory and principles of operation
- b. General and specific plant operating characteristics
- c. Plant instrumentation and control systems
- d. Plant protection systems
- c. Engince ed safety systems
- f. Normal, abnormal and emergency operating procedures
- g. Radiation control and safety
- h. Technical specifications
- 1. Applicable portions of Title 10, Chapter 1, Code of Federal Regulations The lectures and other training sessions shall be spaced throughout the I year to the extent practical taking into account heavy vacation periods and intrequent operations such as refuelings.
l Amendment 73 13.2-6 Oll (November 1978)
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J If, as determined by the Supervisor of Operations, an individual has
. Justifiable cause such as sickness or emergency shif t coverage, he may be excused from required assigned lectures up to a maximun of 20% of the total lecture time.
Examinations The requalification program shall include annual written examinations to be administered to the licensed operators and senior operators. The annual examinations, similar to NRC licensing examinations, shall cover the material as outlined above.
4 Periodic written and/or aral examinations shall be administered by the Training Coordinator or his designee in order to provide further reference l
1 points toward determination of the licensed operator's requalification i progress. 4 i
The program shall also include an annual oral examination. This examination will verify each operator's knowledge of operating and emergency procedures. A written check list will be utilized to insure uniform coverage.
On-T he-Job-Training I
4 i Each licensed operator and each licensed senior operator either manipulates the controls or directs the activities of individuals during the plant control manipulations during the term of his license. For licensed i
operators and senior operators, these manipulations shall consist of at i least ten (10) reactivity control manipulations. These manipulations shall 4
involve reactor startups, reactor shutdowns and other control evolutions 4 Such OS*
O (Noverber 1978) 13.2-7 Amendment 73
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- a. Reactor startup to the point of adding heat
- b. Orderly reactor shutdown
- c. Manual control of S/G level during startup or shutdown
- d. Operation of the EHC syst,a in manual during critical reactor operations. j 4
- c. Boration during power operations j
- f. Dilution
- g. Operation of the manipulator crane over the core during refueling.
- h. Any significant (> 10%) power change during manual rod control j ope ra t ion.
- 1. tianual rod control prior to and during generator synchronization. )
1 J. Turbine stop valve exercising or testing j 1
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The various reactivity manipulations performed er directed by each licensed l Individual shall be recorded and a variety of manipulations shall be required. The Training Coordinator shall periodically review and approve the recorded manipulations.
Each licensed operator and senior operator will be required to be familiar with all design changes, procedure changes and facility license changes to an extent commensurate with his duties. An Adninistrative Procedure has been written to implement tt.is requirement. The procedure required that design change memoranda, procedure changes, and license changes he reviewed by each operator. The Supervisor of Operations or his delegate is responsible for assuring that all operators have complied with these instructions. lie will maintain a naster list as a record of compliance.
Each licensed operator and senior operator will be required to review the contents of abnormal and emergency procedures once in each two year period.
A record of this review, similar to that described above, will be naintained by the Supervisor of Operations or his delegate.
In cases of substantial changes, the above topics may be covered in one of the preplanned lectures.
O Amendnent 73 13.2-8 (November 1978)
j Simulator training nay be used by licensed personnel to satisfy the
! On-The-Job-Training requirements involving reactivity control manipulations
- and the requirements to demonstrate understanding of selected plant apparatus, nechanians and applicable operating procedures.
Evaluations The licensed operator requalification program shall include personnel evaluations in the following areas:
- a. Annual written examination results. These examinations will be 1
graded by the Training Coordinator. Personnel scoring better than 80% on any section may be excused from the lectures covering that subject in the following year's program. Personnel scoring between 70%-80% on any section will be required to attend the appropriate l 1
lectures in the following year. Personnel scoring less than 70% in l any section of the test will be required to participate in remedial training in that area and be reexamined. Any personnel who receive an overall grade of less than 70% will be removed from shift and placed on an accelerated training program. This training will continue until they obtain a passing grade on re-examination.
- b. Periodic written examination results. These examinations may be open or closed book and will be graded by the Training Coordinator.
A passing grade of 80% is required on these periodic exams. Anyone receiving less than 80% will be required to participate in remedial training until he is successful on a re-examination.
- c. Annual oral examination results will be evaluated on a pass-fail basis by the Training Coordinator. Areas of failure will require
. remedial training and re-examination.
O (November 1978) 13.2-9 Anendment 73
f
- d. Responsa to actual abnormal or energency conditions will be evaluated by the Supervisor of Operations and the Training Coo rc'ina t o r. These situations will be graded on a pas,s-fall basi. i and areas of weakness will be accentuated in subsequent training 4
sessions.
1 Licensed Plant Staf f !! embers ,
i Licensed and senior licensed nembers of the plant staff who do not normally perform or direct activities involving reactor operation will nevertheless be required to participate in the requalification progran. In many cases the normal duties of these personnel will take the place of specific training.
Examples of this are such things as serving on the Plant Staff Review Committee and acting as an instructor for the scheduled lecture series.
Whenever it can be documented that a licensed plant sta f f member has undergone equivalent training as part of his normal job duties, he may be excused from that portion of the formal requalification program. Such personnel shall, however, participate in at least the following portions of the program.
i l 1. Take the annual written examination and participate in the lecture series as required based on the results.
- 2. Operate or direct the operation of the reactor through at least ten reactivity changes.
- 3. Systematically review design, procedure and facility license l
changes, i
- 4. Systematically review all abnormal and energency procedures on a yearly schedule.
- 5. Be subjected to a systematic oral denonstration exanination regarding the actions to be taken during sinulated abnormal and 1
emergency conditions. I 1
O Amendment 73 13.2-10 (November 1978) l l
[ l l
l i
Records
("
Records docunenting the licensed operator requalification program shall contain the following:
- a. The annual written examinations graded test papers and answer sheets will be kept at least until they have been audited by the NRC.
- b. Per[ odic exaninations, results and answers will be kept for a minimum of two years.
- c. Annual oral examinations and results will be kept until they have been audited by the NRC. l
- d. Records of the review of design changes, emergency procedures, procedure changes, license amendments, etc., will be maintained.
for at least two years.
- e. Records of attendance at lectures, audio-visual presentations, self I study programs and all other training sessions will be kept until they have been audited by the NRC.
At all tinies there will be a ninimum of one year's hard copy available at the plant. Records nore than one year old will be available at the plant on microfilm.
l l
4 (November 1978) 13.2-11 Amendment 73
i l
13.2.3 REPLACEMENT TRAINING Replacement training for operations personnel will follow the established Company procedures and programs. Each new Auxiliary Operator (nonlicensed position) undergoes a formal 30-month training progran. This progran consists of 75 lesson assignments, 12 review tests and three qualification exaninations. Each lesson assignment requires the t rai nee to study the applicabic sections of the Standard Training tlanual, Plant Equipment Desc ription fianual and Plant Operating Procedures Planual and then provide written answers to a series of questions. Each lesson assignment (except for those dealing with theoretical subjects) also includes a practical demonstration of the operating skills developed by the lesson. Each completed lesson, including the practical demonstration, is evaluated by the Shift Foreman. Records of each man's t raining are maintained by the Training Coordinator.
To insure uniform application of this program, the Shift Foremen are provided with a Supervisor's Guide for each lesson. This guide points out the key points to be considered in the written evaluation of each lesson.
A continuation of this basic program is provided for personnel above the level of Auxiliary Operator. This progran is identical to the one described above except that it covers more advanced topics. There are a total of 93 lesson assignments in the advanced progran, i
Those training programs have been in operation for a number of years and ;
have been effective in providing well trained replacement personnel in conventional plants.
1 O
Amendnent 73 13.2-12 (November 1978) !
l
13.2.4 RECORDS O
The Training Coordinator is responsible for the maintenance of general training files for all personnel. The files are maintained in accordance with an administrative procedure (see Section 13.5 of this chapter) which states that the files shall contain records of qualifications, experience, i training and retraining for each member of the plant organization. Each department head is also responsible for developing performance standards for personnel under his supervision in accordance with Company practice. l These performance standards are used in evaluating individual performance and the overall effectiveness of training programs.
1 l
I l
l O
l l
O (November 1978) 13.2-13 Amendment 73
TABLE 13.2-1 Sheet 1 of 7 SUPHARY OF ACTIVITIES EMPLOYED IN TRAINING PROGRAMS g FOR PERSONS IN INITIAL DIABLO CANYON OPERATING ORGANIZATION
- 1. Ilumboldt Bay Experience Many of the key individuals in the initial plant organization were members of the ilumboldt Bay staff prior to their transfers to Diablo Canyon and have extensive nuclear experience in their areas of responsibility.
Certain other individuals, who were not members of the ilumboldt Bay staf f, have been assigned there for appropriate periods to participate in ,
operations involving their areas of responsibility.
l
- 2. PWR Experience .
Key individuals were assigned to an operating PWR (or one in the process of preoperational and/or startup testing) to observe and/or participate in operations involving their areas of responsibility. The plants involved included R. E. Ginna, H. B. Robinson, Connecticut Yankee, Point Beach, and 7
San Onofre. Assignments ranged from three weeks to seven months, with most lasting approximately one month.
- 3. Participation in Company's Diablo Canyon Task Force Tids group consists of selected technical and operating supervisory personnel, and is responsible for the preparation and training material, operatit.3 manuals, licensing material and Technical Specifications, test procedures, and for performing an operational review of the plant design.
- 4. Design Lecture St.-i s This four week course ucs conducted in March 1971, at the Westinghouse l Atomic Power Division in Pittsburgh, Pennsylvania. Fifteen supervisors on the plant staff and one member of the Department of Steam Generation attended this course. The trainees were given a series of lectures lO (November 1978) Amendment 73
TABLE 13.2-1 Sheet 2 of 7 covering the function, design description, control and instrumentation, h normal and abnormal operation, and maintenance of all principal componenta of the Diablo Canyon Units 1 and 2 nuclear steam supply sys tems. These lectures were given by Westinghouse design engineers who were closely associated with the design of the plant. The lectures were supplemented by study of written information on pressurized water reactor technology provided by Westinghouse and trips to Westinghouse manufacturing facilities I where trainees were afforded an opportunity to witness actual f abrication 1
of components.
l
- 5. Nuclear Technology Course This course is taught by members of the plant staff and utilizes the Company's INTRODUCTION TO NUCLEAR POWER training manual as a text. The purpose of this course is to provide a general background in the field of nuclear power plant technology. The major topics which are included in the course are: l l
l
- a. Basic mathematics. i
- b. Basic atomic and nuclear physics.
- c. Introduction to nucicar reactors and nuclear power plant cycles,
- d. Light water reactor physics.
- e. Ilea t transfer considerations in light water reactors.
- f. Operating characteristics of light water reactors,
- g. Nuclear instrumentation.
- h. Chemical, radiochemical, and waste disposal considerations in water reactor operation.
- i. Reactor safeguards.
4 The treatment of these topics is of suf ficient depth to prepare an individual for applicable portions of the Senior Operator License exam-ination. The complete course, which is intended for license candidates, takes approximately 4 weeks and covers each of these topics in detail.
Abbreviated versions of the course covering subjects directly related to Amendment 73 (November 1978)
l TABLE 13.2-1 Sheet 3 of 7 i
, their duties and responsibilities are given to other personnel as appropriate.
d
- 6. Radiation Protection Training Course This course is taught by members of the plant staff and utilizes the Company's RADIATION PROTECTION TRAINING MANUAL, the Company's RADIATION CONTROL STANDARDS AND PROCEDURES, and other appropriate material as texts.
The STANDARDS are a compilation of technical statements of policy covering each aspect of a nuclear power plant's radiation protection program, and are based upon the requirements of 10 CFR 20 and other applicable regula-tions. The PROCEDURES provide practical information regarding the implementation of the STANDARDS and are based upon adaptations to nuclear power plant requirements of procedures and practices widely used throughout I
the atomic energy industry. - The TRAINING MANUAL is a general work which covers theory and other background material. The major topics covered in this course include:
- a. Basic radiation physics and biology.
- b. Sources of radioactivity in nuclear power plants.
- c. Radiation protection instrumentation.
- d. Fundamentals of shielding.
- e. Personnel exposure limits,
- f. Protective clothing and equipment,
- g. Control and transfer of radioactive materials.
- h. Decontamination practices.
- 1. Radiation monitoring techniques.
- j. Control of access.
- k. Records and reporting requirements.
The complete course for Radiation and Process Monitors requires approximately 4 weeks. A similar course designed for NRC license examination candidates required approximately I week. Personnel in other classifications receive g
shorter courses covering those subjects directly related to their duties and responsibilities.
(November 1978) Amendment 73
TABLE 13.2-1 Sheet 4 of 7
- 7. Plant Design and Operation Seminars This seminar course is conducted by supervisory personnel on the plant staf f and covers the design, description, and operation of each plant system plus related topics such as the Technical Specifications and the Site Emergency Plan. The course is primarily designed for operators and is expected to last npproximately 8 weeks. As appropriate, personnel in other classifications will receive shorter courses covering systems and equipment related to their areas of responsibility.
- 8. NRC Operator and Senior Operator License Examination Seminars These seminars are conducted by supervisory personnel on the plant staff for the benefit of license examination candidates. They consist of a review of appropriate items in activities 5, 6 and 7 above plus discussions of additional topics required to cover the items listed in 10 CPR 55.21-23.
1 The length of this program will be determined following an evaluation of the needs of the individuals involved, but based upon llumboldt Bay experi-ence, it is expected to last approximately four weeks.
- 9. P-250 and P-2000 Computer Maintenance Courses These courses, lasting a total of 16 weeks, were conducted in the f all of 1970 by Westinghouse Computer and Instrument Division personnel, and were designed to provide comprehensive coverage of the construction, operation, repair and maintenance of the P-250 and P-2000 computera. The course attended by Diablo Canyon personnel was held at the Company's Pittsburg Power Plant where a P-250 computer was available for use for the students.
O' Amendment 73 (Novenlmr l'378)
I
TABLE 13.2-1 Sheet 5 of 7
- 10. Instrumentation and Control Course This is a 12-week course intended for instrument maintenance supervisors and technicians, and is conducted by instructors from the Nuclear Instru-mentation and Control Department of Westinghouse at the department head-quarters in Baltimore. The general subjects covered include the design, maintenance and testing of the solid state rod control system, flux map-ping system, nuclear instrumentation system, radiation monitoring system, l and solid state protection system. The course combines both formal l classroom lectures on systems and modules and practical bench work with the equipment. Diablo Canyon personnel attended this course beginning in January 1972.
- 11. Process Controls Course This two-week course was conducted at the Portland General Electric Trojan site in June 1972 by Westinghouse Computer and Instrument Division personnel.
f The course material included lectures on both systems and modules for the various process control systems (feedwater control, steam dump, pressurizer level and pressure). In addition the various modules were available for hench work by the students, l
- 12. Refresher Course in Radiological Engineering This three-week course was conducted in 1972 by personnel from the health physics staff of the General Electric Company Vallecitos Nuclear Center.
It was designed to provide graduate level refresher training in radio-logical engineering topics such as internal and external radiation dosim-etry, radiation biology, atmospheric diffusion modules, instrumentation, I and environmental pathways. It consisted primarily of formal classroom lectures.
(November 1970) Amendment 73 I
1
TABLE 13.2-1 Sheet 6 of 7 13 Chemistry and Radiochemistry Seminars These seminars are conducted by supervisory personnel on the plant staff for training of Radiation and Process Monitors, and will take approximately four months. The subject matter for these seminars will include such topics as basic chemistry, laboratory techniques, radiochemical methods, and theory and use of counting room equipment. A vvriety of texts will be employed, including Company procedures manuals, vendor's instruction manuals, and standard chemistry and radiochemistry texts. In addition, the classroom work will be supplemented by actual laboratory training as appropriate.
1 i
- 14. Nondestructive Testing School l l
l l
This three-week course is presented at the site by instructors from l l
General Dynamics /Convair. The class consists of both formal lectures i and practical demonstrations, and persons completing it and successfully passing the examinations are qualified as ASNT level Il inspectors for l l
radiography, ultrasonic testing, magnetic particle testing, and liquid .
penetrant testing.
- 15. Operational Core Analysis Training This three-week course is presented in Pittsburgh, Pennsylvania by Westinghouse and is intended for Nuclear Engineers. It discusses the theory and operation of the Operational Core Analysis Computer Codes which will be used to monitor core thermal-hydraulic performance and fuel depletion.
O Ngendnw.nt 73 (November 1978)
TABLE 13.2-1 Sheet 7 of 7
- 16. Simulator Training Candidates for " Cold" NRC licenses will attend a 14-day training program at the Westinghouse reactor simulator at Zion, Illinois.
The course has been established so that the typical day is divided approximately equally into classroom work and " hands on" simulator time.
Emphasis during the first week will be on taking the trainee through a simulated operational cycle from a cold shutdown through plant heatup/
reactor startup/ turbine generator startup/ power operation / plant shutdown /
and plant cooldown. This first cycle will stress familiarization with control board and plant operation under normal operating conditions.
The second week's training will be a repeat of the training received the first week, but at an accelerated pace, and will incorporate the maximum number of minor and major malfunction situations in the time allotted.
The trainee will learn to identify specific malfunctions, analyze the hazards involved, and effeet proper corrective actions. The second week will also incorporate simulation and evaluation of normal and abnormal plant transients and the required operator action to effect recovery from transient and accident situations.
A refresher simulator training course of seven days duration will be provided shortly before initial loading.
17 In-Place Filter Testing Workshop This five-day workshop is conducted by the Harvard School of Public Health in Boston. The subject matter deals with the subjects of theory, design, and testing of HEPA and activated charcoal air filtration systems.
Approximately 50 percent of the time is devoted to classroom lectures and the'other 50 percent consists of laboratory work using DOP generators and detection equipment and other filter testing devices. The Power Plant Engineer attended this course in September 1971.
\
(November 1978) Amendment 73
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13.4 REVIEW AND AUDIT
]3.4.I REVIEW AND AUDIT - CONSTRUCTION PHASE i
i The independent review and audit program which will be employed during'the operation phase is not employed during construction. Rather, independent
- review and audit of construction activities is incorporated into the
. Quality Assurance Program During Design, Construction, and Preoperational
, Testing discussed in Section 17.1.
J l 13.4.2 REVIEW AND AUDIT - OPERATION PHASE Review and audit during the operation phase is accomplished both by senior members of the plant staf f and by independent review and audit groups as
- discussed below, and in Section 17.2.
i
' A
, 1. Plant Staff Review Committee '
In accordance with normal Company practice, a Plant Staff Review Committee will be extablished at the plant, and will meet on a regular basis, and on special occasions as required, to review overall plant operating and maintenance experience; training programs; adequacy of i procedures; and other pertinent items. Items requiring plant staff review include:
- a. Proposed tests of a significant nature involving safety related
- equipment.
i l b. Proposed design changes to safety related system and equipment.
] c. Operating, maintenance, and test procedures and changes thereto l f involving safety related aspects of safety related equipment.
4 I
i e
Amendment 73 13.4-1 (November 1978)
The members of the Committee include the Plant Superintendent, Supervisor cf Operations, Power Plant Engineer, Supervisor of
!!aintenance and Quality Control Supervisor. The Plant Su pe ri n t e nden t may request other plant supervisors or outside consultants to participate in the review meetings as appropriate. For example, the Chemical and Radiation Protection Engineer would ordinarily participate in the review of items involving radiation safety.
i Special Members of the Committee include:
l l
1
- 1. A Station Construction Supervisor during periods when start-up or )
other work directed by Station Constructton is in progress. The Station Construction Supervisor or his alternate may participate in discussion of all matt ers relating to such work, but will not take part in Committee decisions.
- 2. A Quality Assurance Engineer, who represents the Quality Assurance Department on-site and the Department of Engineering Research Supervisor who represents the Department of Engineering Research on-site, or their alternates may partici. ate in discussion of all matters relating to their areas of responsibility, but will remain l independent of Committee decisions.
During the period of plant start up, the General Construction Start-up Engineer will participate in plant staff review meetings.
Further in4ormation regarding the Committee is given in Chapte r 16.
O (November 1978) 13.4-2 Amendment 73
- 2. Independent Review and Audit Committees The Company employs a program of independent review and audit of nuclcar plant operations which has been in effect since the initial operation of the Company's Humboldt Bay Power Plant Unit No., 3 in 1963. This program, which will also be applied to the pre-operational testing, start up testing, and operation of Diablo Canyon, has been reviewed and appropriately modified so that it conforms to the requirements and recommendations of ANSI N18.7-1976, " Administrative Controls and Quality Assurance for Operational Phase of Nuclear Power Plants."
1 The charter of the General Of fice Nuclear Plant Review and Audit '
Committee (GONPRAC) is shown in Figure 13.4-1. This committee 1
satisfies the requirements of Section 4.3 of ANSI N18.7-1976. Further l
information regarding the CONPRAC is provided in Section 16.5. )
The Company also employs an advisory committee to assure the President r-that compliance with approved procedures and requirements in the operation of nuclear plants has been maintained. A primary function of the President's Nuclear Advisory Committee (PNAC) is to examine and report to the President on the activities of the CONPRAC. The Committee charter of the PNAC is shown in Figure 13.4-2.
\
Amendment 73 13.4-2a (November 1978)
85-13 Pacific Gas and Electric Company C0ttilTTEE CHARTER GENERAL OFFICE NUCLEAR PLA!!T REVIEW AND AUDIT COMMITTEE +
PURPOSE:
r This Committee is established to provide reviews of the audits, actions, and practices relating to operating nuclear power plants which may have '
a ocaring on nuclear safety and environmental matters, and to provide for separate audits relating to nuclear safety and environmental matters in-dependent of the personnel directly responsible for the function or ,
activity being audited.
REPORTS TO:
- Executive Vice President (Operations).
MEMBERSHIP:
- l. Members Engineering
- a. Chief Mechanical and Nuclear Engineer (Chairman)
- b. Chief, Engineering Quality' Control
- c. An Engineer O Steam Generation Department
- d. Manager (Vice Chairman) 1 I
- e. An Engineer Safety, llealth and Claims Department
- f. Manager i g
- g. A Health-Physicist !
Station Construction Department
- h. Ibnager Quality Assurance Department 1 Director Department of Engineering Research
- j. A Biologist
- 2. Executive Vice President (Operations) may appoint other members as may be necessary to meet the requirements of a following paragraph i headed " Qualifications".
)
- 3. Each member shall appoint an alternate to serve only when the member l cannot attend meetings or perform other required actions. The Committee shall maintain current records listing by name designated members and alternates.
APPOINTMENTS AND CHANGES:
- By Executive Vice President (Operations).
O, 4
l !
(November 1978) Amendment 73 !
Figure 13.4-1 (Sheet 1)
Peplating 12-1-76 Pace l
of 6
i 85-15 Pacific Gas arc Electric Company
_ COPEITTEE CHARTER l l
Committee Name: i GENERAL OFFICE NUCLEAR PLANT IU:NIEW AND AUDIT CO?Ct1TTEE l
QUALIFICATIONS:
- 1. At least four members shall have a minimum of a Bachelor's Degreee in Engineering or the physical Sciences and at least one member shall have a Bachelor's Det;ree in the Natural Sciences. These members shall each possess a minimum of three years of professional level experience in such areas as nuclear services, nuclear plant nperation, nuclear engineering, or power plant biological monitoring and evaluation. The Committee shall have the necessary overall nuclear background to determine when to engage the services of consultants and contractors for solving problems beyond the expertise of the Company organization.
- 2. Members and alternates shall collectively have the capability required to review the areas of:
- a. reactor operation I
- b. nuclear engineering j
- c. chemistry and radiochemistry '
- d. metallurgy
- e. instrumentation and control O f. radiological safety Q g.
h, mechanical and electrical engineering quality assurance
- 1. power plant .lological monitoring and evaluation.
FUNCTIONS:
1
- 1. The following shall be reviewed by the Committee: i !
l 1 l
- a. Written safety evaluations of changes in a facility, changes in procedures, and tests or experiments proposed and completed with-out prior NRC approval under the provisions of 10 CFR 50.59(b) to l verify that such changes, tests er experiments did not involve a l I
change in the technical specifications (+) or an unreviewed safety I question as defined in 10 CFR 50.59(c). i l
- b. Proposed changes in procedures, changes in a facility, and tests and experiments determined by a Plant Staff Review Committee to
{ involve a change in the technical specifications or an unreviewed
, saf ety quest ion as de'ined in 10 CFR 50.59(c), and any other
, proposed changes, tests or experiments which may be referred to l
{ the Caumlttee for its review by a Plant Superintendent or by a j i General Otfice department.
I
- c. Proposed changes in the technical specifications or license amend-
- ments relating to nuclear safety or the environment prior to im-
[sy i
plementation of the change or amendment. !
vi ,
l
- ) The term " technical specifications" as used herein refers to both health and safety technical r,pecifications and environmcntal technical specifica-tions appended to nuclear power plant f acility operation 11 cerises.
issued 6-1-77 Firure 13.4-1 (Sheet 2)
Leplating 12-1-76 Pace 2 of 6
'~ 85-i3 Pacific Gas and Electric Company COMMITTEE CHARTER
/ --.
( Corrrni t tes Name :
GENERAL OFFICE NUCLEAR PLANT REVIEW AND AUDIT COMMITTEE
- d. Violations, deviations, and abnormal occurrences such as:
(i) Violations of applicable codes, regulations, orders, technical specifications, license requirements, or internal procedures i or instructions having nuclear safety significance on facility i operation:
(ii) Significant operating abnormalities or deviations from normal l
or expected performance of plant safety-related structures, systems, or components: and I
- (iii) Reportable occurrences as defined in the plant technical specifications.
Review of events covered under this paragraph d shall include timely reporting to appropriate members of Company management on the results of any investigations made, and the recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.
,, e. Reports and meeting minutes of the Plant Staf f Review Committees.
~'
- f. Communications with the NRC involving nucicar safety or environmental matters relating to Company's nuclear power plants.
- g. Any other matter involving safe operation or environmental matters !
relating to Company's nuclear power plants which a Committee member i
deems appropriate for Committee consideration, or which is referred to the Committee by a Plant Superintendent, a General Office depart-ment, or the President's Nuclear Advisory Committee.
I
- 2. It is the function of the Committee to provide Company management with assurance that practices relating to operacing nuclear plants conform to laws and regulations, license provisions, and internal Company rules and policy related to nuclear safety or environmental matters. To accomplish this, the Committee shall:
i a. Review and evaluate audits conducted by those Company Departments i
having auditing responsibility.
- b. Perform audits itself employing either the entire Committee or parts thereof, as appropriate.
- c. Perform or otherwise arrange for accomplishing other auditing i
1 activities as may be directed by Executive Vice President (Operations).
(November 1978) Amendment 73 issued 6-1-77 Figure 13.4-1 (Sheet 3)
Feplating12-1-76 3 g, 6 Peoe
.~. . . _ _ _ _ _ _ _ _ _ _
b>-13 Pacific Gas and Electric Company COMMITTEE CHARTER O Com Mee Name:
d GENERAL OFFICE NltCLEAR PLANT REVIEW AND AUDIT CO:!MITTEE
- d. Recommend management approval for employing consultants to conduct audits whenever necessary,
- e. Report its findings, conclusions, and recocmendations to Cvmpany management.
PROCEDURE:
- 1. The Committee shall meet at least semi-annually and as required at the call of the Chainman, or the Vice-Chairman in the absence of the Chair-man. During the period commencing three months prior to initial fuel ,
loading and for one year after commercial operation of a nuclear unit I the Committee shall meet at.least quarterly.
I
- 2. A quorum for meetings or actions will consist of five members with at l 1 east one member from each of the three departments with two member- l ships. Either the Chairman or the Vice-Chairman shall be present at meetings or available for action. In addition, the member with capability in power plant biological monitoring and evaluation (or his alternate) shall be present when matters in his area of
,_ expertise are on the meeting agenda.
l
(
, *3. The Chairman shall appoint a secretary from the membership of the Coumittee who will keep written minutes of each meeting. The Chairman
! shall distribute copies of minutes within fourteen days following the j meeting to Committee members, to the Company Officers to whom the Committee members report, and the President 's Nuclear Advisory Committee. l Copics of the minutes shall be retained in Committee files.
l
- 4. The Chairman shall provide each member with an agenda in advance of
}' each meeting, and shall make arrangements with appropriate General Office departments for prior distribution of material to be reviewed by the Connittee at meetings.
- 5. Audits may be conducted at the discretion of the Committee by Committee members, either individually or as a group. Additionally, the Committee tay arrange for Company departments or sections, such as Quality Assurance, Steam Generation, Safety, llealth and Claims, etc., to conduct audits of special interest to the Committee. Reports of audits conducted by the
, Committ ee shall be distributed to the Company Of ficers to whom the Committee members report and the president's Nuclear Advisory Committee j uithin thirty days after completion and approval by the Committee.
i l 6. The Committee shall review auditing activities related to nuclear safety
,~ or environmental matters involving operating nuclear power plants. It shall also schedule its own audits and take other appropriate action as l
necessary to provide a comprehensive audit program. As a part of verify-Oi ,
ing compliance with applicable statutes, regulations, license provisions, and internal rules and policies, the auditing program will investigate I
such things as:
t ss vec 6-1-77 Figure 13.4-1 (Sheet 4)
'eD i at i ng 12-1-76 p3ae 4 og 6
l
B5-I3 Pacific Gas and Electric Company COMMITTEE CHARTER Conynittse Name:
GENERAL OFFIC'E NUCLEAR PLANT REVIEW AND AUDIT COMMITTEE 1
- a. implementation of' site emergency plans
- b. implementation of site security plans
- c. implementation of environmental monitoring programs
- d. corrective action following reportable occurrences
- e. an NRC inspection
- f. qualification and training of personnel, i
- 7. All audits conducted by the Committee shall be individually authorized by the Committee at scheduled meetings. The action authorizing an audit will designate the individuals making up the audit team and will designate the team leader. Committee audit teams may be made up of:
- a. Committee members.
- b. Other Company personnel made available on request by the Committee.
The Committee will request an engineer from the Quality Assurance Department for each audit team.
i c. Outside consultants when made available. l Committee audit teams may contain personnel from the department being )
audited, but such personnel shall not audit activities for which they l have a direct responsibility. l O 8. Audits by the Committee may be either announced or unannounced.
I Auditors shall discuss their findings in an exit interview with the person or persons being audited before leaving the site of the audit.
Audit reports, while still in draft form, will be furnished to the person or persons audited for comment before the reports are made I
final. Final reports will be accepted by the Committee before being distributed to management personnel. The Committee shall determine if necessary actions have been taken to correct deficiencies discovered during audits.
- 9. Reports of audits and other Committee activities will be distributed to Executive Vice Presidents, Senior Vice Presidents, the Vice President-Electric Operations, the President's Nuclear Advisory Committee, other affected management personnel, and others as directed. Copies of audit reports and related documents will be retained in the Committee's files.
i
- 10. Nothing herein shall eliminate the necessity or responsibility for audits by Company departments in the discharge of their regular duties.
- 11. The Committee or its Chairman may call on other Company personnel for consultation. The Committee may also use outside consultants when made available by Company management.
O (November 1978) Amendment 73 issued 6-1-77 Figure-13.4-1 (Sheet 5)
Replac.Ing 12-1-76 Pace 5 og 6 )
.. - . . ~ -_ .. .- . ...
00-1)
Pacific Gas and Electric Company COMMITTEE CHARTER Cownittee Name:
CENERAL OFFICE NUCLEAR PIRIT REVI9,' AND AUDIT C0!DfITTEE AUTHORITY:
- 1. The Committee is fact-finding and advisory in nature and is not an executive body. The Committtee does not authorize expenditures or issue direct orders.
- 2. The Chairman of the Committee is authorized to communicate directly with Company Officers, Department Heads, and Division Managers on matters relevant to the Conmittee's functions.
- 3. The Committee shall have the authority to inspect any Company records pertaining to plant safety or environmental matters.
C0!&lITTEE ESTABLISHED:
October 22, 1962, by N. R. Sutherland.
, . APPROVED:
J. D. WORTHI!!GTO1 Executive Vice President (Operations) i i
('
%/
l l
l
' lssued 6-1-77 Figure 13.4-1 (Sheet 6)
Replacing 12-1 Page 6 g, 6
85 -l !
Pacific Gas and Electric Company COMMITTEE CHARTER Committce Name:
'\") PRESIDENT'S NUCLEAR ADVISORY C0KMITTEE
, PURPOSE:
To advise the President of the results of reviews and audits conducted to determine compliance with approved procedures and requirements in the operation of nuclear fueled generating plants and make recommenda-
, tions as appropriate.
REPORTS TO:
President I
4 MEMBERSHIP:
1
- Manager, Safety, llealth and Claims Department (Chairman)
- Executive Vice President Assistant General Counsel l Director, Quality Assurance Department (Secretary) !
l l
APPOINTMENTS AND CHANCES:
(
By the President FUNCTIONS:
- 1. Examine and report on the activities of the General Office Nuclear Plant Review and Audit Committee. l
- 2. Perform or cause additional audits to be performed at its discretion or at the direction of the President.
- 3. Receive and review at its discretion any Company audit or related reports involving the operation of the Company's nuclear power plants.
- 4. At its discretion or at the direction of the President, review reports and other communications between regulatory agencies and the Company '
on operation of nuclear plants which relate significantly to nuclear '
safety.
- 5. At least semi-annually or at the direction of the President, prepare and submit reports of audits and other committee activities incluaing recommended changes in practicec relating to operating nuclear plants, to the President and other officers concerned.
) 6. At its discretion, investigate any phase of plant operation related to nuclear safety and per form such other tasks as may be assigned by the President.
- Revision
.(November 1978) Amendment 73 tssued 6-7-78 Figure 13.4-2 (Sheet 1) bepleting 2-7-74 Pace 1_ _o.f _ _ 2 _ _ _ _ _
t 85-14 Pacific Gas and Electric Company COMMITTEE CHARTER
( Conrnittee Name:
'~'
PRESIDEt:T'S NUCLEAR ADVISORY COMMITTEE PROCEDURE:
- l. Regular meetings will le held at least twice a year. Special meetings
! may be called at the discretion of the Chairman.
- 2. The Secretary will prepare and distribute minutes of committee meetings to the committee for approval. Approved minutes shall be distributed to the President, and to others as the committee deems necessary.
- 3. The Chairman may invite non-members to meetin'gs when their presence is desirable. Members may arrange for attendance of departmental associates when their participation is needed.
AUTHORITY:
- 1. The Committae is fact-finding and advisory in nature, and is not an executive body. The Committee does not authorize expenditures nor issue direct orders.
- 2. The Chairman of the Committee is authorized to communicate directly with Company Officers, Department Heads, and Division Managers on matters relevant to the Committee's functions.
COMMITTEE ESTABLISHED:
October 7, 1965, by the President under the original title " Nuclear Plant Audit Committee".
-' 'I APPROVED:
./9 7
. . +i i>w / - s r> s su
, JOIIN'F. BONNER
- President and Chief Executive Officer O l 4
" Revision Amendment 73 (November 1978) Figure 13.4-2__(Sheet 2)
Issued 6-7-78 Replacing 2-7-74 PaBe 2 of 2