ML20005C063
| ML20005C063 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/04/1981 |
| From: | Reynolds J CENTER FOR LAW IN THE PUBLIC INTEREST, JOINT INTERVENORS - DIABLO CANYON |
| To: | PACIFIC GAS & ELECTRIC CO. |
| References | |
| ISSUANCES-OL, NUDOCS 8111180343 | |
| Download: ML20005C063 (25) | |
Text
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RELATED CORRESPONDECE l _ _
00CMETED-USNRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 8 NOV -9 . Pl2 :02' BEFORE THE ATOMIC SAFETY AND LICENSING BOARD. . . . ,
)
In the Matter of )
)
PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 O.L.
) 50-3?3 0.L.
N (Diablo Canyon Nuclear Power ) fre Plant, Units 1 and 2) )
! #f0!D9 pl NOV171981d
( u.s. m -
RESPONSE OF JOINT INTERVENORS TO APPLICANT '\g g/}
PACIFIC GAS AND ELECTRIC COMPANY'S THIRD SET OF INTERROGATORIES y
, Joint Intervenors hereby respond to Pacific Gas and Electric Company I s Third Set of Interrogatories dated October 15, 1981 as follows: .
l Response 1 The PORV's and Block Valves
- are not specifically identified ;
in the FSAR Section 3.2 tables but they are included in the Hosgri Seismic Evaluat. ion (Vol. III Table 7. 8, " Summary -
Seismic Qualification Valves Required for Normal Shutdown and/or Cold Shutdown.") There are few other details of the classification and qualification of these three types of alves.
- In contras t, Diablo Canyon Safety Valves are classified as safety-grade and subjected to the requirement of Design Class I, Code Class I as described in FSAR Tables 3.2-1, 3.2-2, 3.2-3, and 3.2-4. Similarly, they were identified in the Hosgri Amendment to the FSAR as having been seismically tested (See Hosgri Seismic evaluation, VOL. III, Tabic 7-7 " Seismic Qualification Minimum Required Active Valves for Hot Shutdown l and/or Cold Shutdown.") oy[#l I 8111180343 811104 PDR ADOCK 05000275 h ,
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s However, proper operation of power operated relief valves, associated block valves and the instruments and controls for
. these valves is essential to mitigate the consequ.ences o; accidents in that thei'r failure can cause or aggravate a LOCA. The re fo,re , these valves must also be classified as safety-grade components and required to meet all safety-grade design criteria. There is insuf5icient information to know if the existing valves and their associated equipment meet the necessary requirement to insure reliable performa'n ce of their safety function under worst case accident conditions.
Response 2
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See Resporce 1. The failure of control and/or -ins truments
- could lead to failure of the associated valves, thereby causing or aggravating a LOCA. Thus, the associated controls and ins tru-ments fhr these valves must comply with applicable codes, s tandards, and regulatory practices . The 4
NRC Standard Review Plan (NUREG 75/087 Section 7, Table 7-1) identifies the acceptance criteria for safety-related instru-
! mentation and control equipment which should be applied to these components. A copy of this table 'is attached.
Until adequate details are provided on how the valves l and components meet the above safety and acceptance criteria, 4
! there can be no assurance of their ability to perform properly in all off-normal and accident condi'tions.
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Response 3 In addition to the discussion in Responses 1 and 2, there are conditions where the block valves and PORVs may individually or collectively constitute a potential break in the reactor coolant pressure boundary. Failure to operate correctly, in either opening or closing, may cause or aggravate a small LOCA.
The valves can also play an important role in mitigating the effects of an ATWS accident. They may also serve as a mechanism .
for control and/or mitigation of accident conditions when called upon to operate in the bleed and feed mode (in conjunction with Safety Injection) . Components which have this large an impact on pressure boundary integrity, accidents, and s afe ty should be classed as safety-grade. Examples include the following:
( a.)
A block valve ' failure to close when the PORV sticks open can create a small LOCA, one of the design basis events in the FSAR. In' the preceding example of a PORV stuck open, mitigation of the small LOCA may be accomplished by closing the associated block valve.
(b) There are sequences where failures of the block valves would prevent operation of the PORV's. Thus, block valve failure could prevent the use of PORV's as a means of overpressure protection during low temperature operation. The Applicant's response to NUREG-0578 (TMI Lessons Learned) refers to both block valves and PORV's in regard to low temperature over-pressurization protection. (PGSE response to Short Term Lessons Learned, February 29, 1980, page III-B-13.
e
l (c) Failure of a PORV to close, and the failure of the ,
block valve to be closed by the operator coupled with the failure of the emergency coolant systems and ,
auxiliary feedwater system functions could result in core damage (for example , see the TMI-2 accident scenario) .
(d) Although the normal procedures do not appear to call for use of the block valves or PORV's to shutdown
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the reactor and maintain it in a safe shutdown condition, there are conditions where they may be called upon to assist in maintaining the plant in a safe shutdown condition. The TMI-2 accident and post-accident mitigation is such an example.
ATWS is not a design basis event for Diablo Can'on y (e')
at this time .
Therefore , ATWS has not been protected against solely with safety grade equipment.
Response 4 .
In addition to the accident scenarios set forth in Responses 1, 2, and 3, during a small break LOCA where there is also a PORV/ block valve failure, there is a possibility of erroneous behavior of the pressurizer function, pressurizer level indication, and vessel level indication. Operator action and, thus, system behavior in the light of such possibly misicading information cannot be predicted with certainty. .
Response 5 Yes. See also response to Interrogatory 6.
Response 6 Yes. See also response to Interrogatory S.
Response 7 No applicable.
Response 8 Not applicabic.
Response 9:
Diablo Canyon safety valves are classified as safety-grade and subj ected -to the requirements of Design Class I, Code Class I as described in FSAR Tables 3.2-1, 3.2-2, 3.2-3, and 3.2-4. Similarly,'they were identified in the Hosgri Amendment to the FSAR as having been seismically tested (see Hosgr seismic evaluation, Vol. III, Tabic 7-7, Seismic Quali-fication Minimum Required Active Valves for Hot Shutdown and/
or. cold Shutdown.") The PORV's and block valves are not spe-cifically identified in the FSAR Section 3.2 tables but they are' included in the Hosgri Seismic Evaluation (Vol. III, Table 7.8, " Summary-Seismic Qualification Valves Required for Normal Shutdown and/or Cold Shutdown." There are few other details of the classification and qualification of these three types of valves.
Proper operation of power operated relief valves, associat-ed block valves and the instruments and controls for these valves is essential to mitigate the consequences of accidents.
In addition, their failure can cause or aggravate a LOCA.
Therefore, these valves must also be classified as safety-grade
i components and required to meet all safety-grade design cri-teria. There is insufficient information to'know if the ex-isting valves and their associated equipment meet the neces-sary performance requirements to insure reliability perform-ance of their safety function under worst case accident con-ditions.
Similarly, the associated con. trol and instruments for these valves must comply with applicable codes, standards, etc. The NRC Standard Review Plan (NUREG-75/087), Section 7, Table 7-1) identifies the acceptance criteria for safety-re-lated instrumentation and control equipment which should be applied to these components. A copy of this tabic i,s attached.
Until details are provided on how the valves and components meet the above safety and acceptance criteria, there can'be no as-surance-of their adequacy to perform properly in all off-normal and accident conditiona. ,
Response 10: ,
(a) and (b) The location and intended purpose of each such valve are set forth in general in the Diablo Canyon Final Safety Analysis Report. The Ap-plicant, as the designer of the plant, should be thoroughly familiar with the location and in-tended purpose of each such valve. Also see
" Applicant's-Answers to Joint Intervenors' Second Set of Interrogatories", dated October 26, 1981, including particularly answer Nos. 46, 49, and 50.
l (c) and (d) See' Response to Interrogatories Nos. 1, 2, and 3.
(e) _ See response to Interrogatory 9.
Resconse 11 While it may be possible to maintain natural circulation at hot stand-by conditions without;the pressurizer heaters and associated controls, such operation may be difficult to control and is contrary to the normal plant operating procedures (see PGSE response No. 45 dated October 26, 1981 to Joint Intervenors Second Set of Interrogatorics for a list of emergency operating procedures that include the use of
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pressurizer heaters). Further, plant safety may be affected by many things, not the least of which is the need to minimize the numb ~er of challenges to the total system integrity cnd to optimize the operability and controllability of systems used in the mitigation or control of abnormal events . The NRR Lessons Learned Task Force found that " maintenance of natural circulation capability is important to safety".* Pressurizer heaters are needed for this capability. In addition, the i pressurizer heaters must also maintain physical integrity for the reactor coolant pressure boundary to be maintained.
- NUREG-0578, page A-2.
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Response 12 See Response No. 11 concerning the nee,d for classification of the components as important to safety. Further, all com-ponents of the pressurizer heater system, including supports and interconnecting wiring should be required to meet the applicable safety-grade design criteria. PGSE has responded
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that only that equipment associated with the capability of obtaining power from the on-site emergency power supply needs to meet GDC 10, 14, 15, 17 and 20 of Appendix A to 10CFR50.
This is further defined in PGSE's Answer to Interrogatory No. 41 as the 480 volt vital breakers 52-1G-72 6 -1H-74, control
- Applicant Pacific Gas 6 Electric Company's Answers to Joint Intervenors' Second Set of Interrogatories, page 1 6 2.
switches and cable between the vital bus and the br'eakers.
This implies then that all of the rest of the pressurizer heater systen has not been designed to meet the safety-grade design criteria listed above. The remainder of the system, therefore, consists of the heaters themselves and their associated controls, along with interconnecting wiring and supports. See PGGE January 26, 1981 submittal to.NRC on -
Full Power License Requirement and associated Figures II .E.3.1-1 6 -2 for diagrams showing the components contained within the pressurizer heater system.
Response 13 See " Applicant's Answers to Joint Intervenors' Second Set of Interrogatories" dated October 26, 1981, particularly Respons_e 34 where the applicant clearly acknowledges that for Diablo Canyon the pressurizer heaters and associated controls are not classified "important to s afe ty" .
Contention 10 does not state that the pressurizer heaters and associated controls fail to comply with "any" specific details in the General Design Criteria but rather that this
- Applicant Pacific Gas G Electric Company's Answers to Joint Intervenors' Second Set of Interrogatories, pages 16 517.
- Philip A. Crane to Frank J. Miraglia, Janua ry 26, 1981, pages II.E-10 through 19.
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equipment has not been classified as safety-grade and therefore not been required to meet the safety-grade design criteria listed. There is _ obviously no way to evaluate that compliance since PGSE has not submitted any detailed information on how these components do or do not meet the specific criteria. This Interrogatory is therefore premature until sufficient detailed
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information is available to evaluate compliance. However, it is likely that non-compliances exist for the following reasons:
- a. GDC 20 requires , among o ther things, that the protection system shall be designed "to initiate the operation of systems important to s a fe ty . "
Stan'ard d Review Plan Table 7-1 extends the applicability of GDC 20 to all instrumentation
- and control functions important to safety.
PGSE's January 26, 1981 response to Full Power License Requirements describes the manual procedure necessary for transferring the pressurizer heater power supply onto the ESF buses . This requires the dispatch o f an operator to a location three floors down in the Auxiliary Building and verbal confirmation that such action has been taken. This complex procedure does not meet the automatic initiation requirements of GDC 20.
- NUREG 75/087, Section 7, Tabic 7-1.
- Philip A. Crane to Frank J. Miraglia, January 26, 1981, page II E-14.
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- b. None of the pressurizer heater system, other than the breakers, switches and portion of the bus connection cables identified in Response 1, has been qualified in accordance with EGC 2 (seismic and environmental qualification) GDC 22 (protection system independence,
" separation") on GDC 3 (fire protection).
- c. Since these components have not been classified as important to safe ty, the requirement of GDC 1 (Quality standards and records) does not appear to have been applied.1/
Response 14 See Response 13.
Response 15 The proposed arrangement addresses only the reliability of power supply to the pressurizer heaters i The heaters and associated controls are still subject to failures introduced through incomplete attention and lack of compliance with the applicable s afe ty-grade criteria (See Responses 11, 12, 13 and 14) .
1/ We note that the classification of pres'surizer heaters and associated controls is currently the subject of Union of Concerned Scientis ts Contention 3 in the ongo.ing TMI re-start hearings (NRC docket 50-289). ,
I e
Resnonse 16:
Documents which were utilized as the base of answers to Interrogatories 1 to 15 herein were identified at the-point of reference in the specific; interrogatory responses. The documents' description included sufficient information to identify the documents, including the identification of the specif_ic page(s) of the document which relate to each inter-rogatory response.
-Response 17:
The term "any pending" as related to contentions is 'm-a biguous. Likewise, the term "thesc" lack's the necessary spe-cific basis. Accordingly, we cannot-identify any additional However, documents or exhibits.as set forth in this request.
assumtng that this request is limited to the subjects identi-fied as " Contention 10" and " Contention 12" in the current Diablo Canyon full power _ license proceeding, the documents or exhibits relied upon which Joint Intervenors may introduce into
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evidence are identified in the foregoing Responses 1 through-
- 16. Additional documents and exhibits may be identified during the ongoing document discovery and as a result of NRC Staff and I
PG&E responses to Governor Brown and Joint Intervenor interro-
! gatories. All parties have access to the documents provided i
! during discovery. Further, such documents and exhibits will gneerally be referenced in the testimony of Joint Intervenors, witnesses which will be submitted to all parties in this full-power proceeding prior to any hearings.
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4 gysponse 18:
Assuming that this request is limited to the subject iden-tified as " Contention 10" and " Contention 12" in the current Diablo Canyon full power license proceeding, the identification of wit-nesses Joint Intervenors may call to testify was set forth in "JJint Intervenors' Identification of Witnesses for Full Power Proceeding" dated November 3, 1981. At that time, the following potential witness for the subject two contentions was identified:
Robert Pollard. Joint Intervenors sill identify other witnesses in the future once the decision is made to preSent other witnesses. At this time Joint Intervenors do not plan to subpoena any witnesses on
" Contention 10" and " Contention 12". Further information in response to this interrogatory will be supplied when it becomes available to Joint Intervenors' counsel.
DATED" November 4, 1981 Respectfully submitted, JOEL R. REYNOLDS, ESQ.
JOHN R. PHILLIPS, ESQ.
Center for Law in the Public Interest 10951 W. Pico Boulevard Los Angeles, CA 90064 (213)470-3000 DAVID S. FLEISCHAKER, ESQ.
P. O. Box 1178 Oklahoma City, OK 73101 By _
[yDEL R. RETNGLDS Attorneys for Joint Inter-venors SAN LUIS OBISPO MOTHERS FOR PEACE SCENIC SHORELINE PRESERVATION CONFERENCE, INC.
l ECOLOGY ACTION CLUB SANDRA SILVER ELIZABETH ~APFELBERG t JOHN J. FORSTER 1
i . u. u. n . . . i NUREG 75/087 enneou f U.S. NUC1. EAR REGULATORY COMMISSION f* g ~g Wpi
! STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION TABLE 7-1 l
ACCEPTANCE CRITERIA FOR INSTRUMENTATION AND CONTROLS These l Table 7-1 contains the acceptance criteria for the SRP sections of Chapter 7.
acceptance criteria include the applicable General Design Criteria. IEEE standards.
Regulatory Guides, and Branch Technical Posittens (BTP) of the Instrumentation and CentrolSystemsBranch(ICSS). The applicability of these criteria to specific sections of Chapter 7 is indicated by an X in the matrix listing of criteria and SAR sections. The BTP listed in Table 7-1 are contained in Appendix 7-A to the Chapter 7 SRP section.
r USNRC STAND ARD REV.EW PLAN
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ACCEPTANCE CRITERI A FOR INSTRU.*iENTATION A4D CONTROL SYSTEMS - TABLE 7-1 l
CRITERIA TITLE APPLICABILITY REM /RKS 7.1 7.2 7.3 7.4 7.5 7.6 7.7
- 1. 10 CFR Part 50 '
- a. 10 CFR 550.34 Contents of Application:
Technical Information X X X X X X X
- b. 10 CFR 550.36 Technical Specifications X X X X X X
- c. 10 CFR 550.55a Codes and Standards .X X X X X X X
- 2. General Design Criteria (GDC), Appendix A to 10 CFR Part 50
- a. GDC 1 Quality Standards and Records X X X X X X
- b. GDC 2 Design Bases for Protection Against Natural '^.cnomena X X X X X X
', 7a Design Bases X X X X X X
][ c. GDC 5 Sharing of Structures, Systems, and Components X X X X X X
- f. DGC 10 Reactor Design 1 X X X X X
- g. GDC 12 Suppression of Reactor Power Oscillations X X X X
- h. GDC 13 Instrumentation and Control X X X X X X X
- i. GDC 15 Reactor Coolant System Design X X X X X l
- j. GDC 19 Control Room X X X X X 'X X
- k. GDC 20 Protection System Functions X X X X X X
- 1. GDC 21 Protection Systems Reliability -
and Testability X X X X X X
- m. GDC 22 Protection System Independence X X X X X X
- n. GDC 23 Protection System Failure Modes X X X X X X
- o. GDC 24 Separation of Protection and Control Systems X X X X X X X RF p. GDC 25 Protection System Requirements 5 for Reactivity Control Mal functions X X X
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I B4 X X X X A X C7 I
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-. PP3 X X X X X X X X X X X X X X J A7 2 X X X X X X
) 7 D
E X U X X X X X X X X X 1
X X X X X X X X X N
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3 4 5 7 8 0 1 3 4 6 0 4 R 6 7 8 9 3 4 4 4 4 4 5 5 5 5 5 E 2 2 2 2 3 3 3 3 T C C C C C C C C C C C I C C C C C C C C D D D D D D D D D D D R D D D D D D D D G G G G G G GG G G G C G G GG G G G G c d e f g h 1
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1 TAN E 7-1 (CONTINUED) 3 l CRITERIA TITLE APPLICABILITY REM %RKS i
l 7.1 7.2 7.3 7.4 7.5 7.6 7.7
- 3. Institute of Electrical and Electronics Engineers (IEEE)
Standards:
- a. IEEE Std. 279 Criteria for Protection Systems See 10 CFR 550.55a(h)
(ANSI N42.7) for Nuclear Power Generating and Reg. Guidr 1.62.
Stations X X X X X X X
- b. IEEE Std 303 Criteria for Class IE Electric See Reg. Guide 1.32.
Systems for Nuclear Power Generating Stations X X X X
- c. IEEE Std 317 Electric Penetration Assemblies See Reg. Guide 1.63.
in Containment Structures for SRP Section 3.11.
Nuclear Power Generating Stations X X X X X X X
- d. IEEE Std. 336 Installation, Inspection and See Reg. Guide 1.30.
(ANSI N45.2.4) Testing Requirements for Instru-mentation and Electric Equipment During the Construction of yd -w Nuclear Power Generating Stations X X X X X X X 4.1" e. IEEE Std 338 Criteria for the Periodic Testing See Reg. Guide 1.118.
of Nuclear Power Generating Station Protection Systems X X X X X X
- f. IEEE Std 344 Guide for Seismic Qualification See Reg. Guide 1.100 (ANSI N41.7) of Class I Electrical Equipment SRP Section 3.10.
for Nuclear Power Generating Stations X X X X X X
- g. IEEE Std 379 Guide for the Application of the See Reg. Guide 1.53.
(ANSI N41.2) % ingle failure Criterion to Nuclear Power Generating Station Protection Systems X X X X X X X
- h. IEEE Std 384 Criteria for separation of Class SeeReg.buide1.57 (ANSI N41.14) IE Equipment and Circuits X X X X X X X
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TABLE 7-1 (CONTINUED)
APPLICABILITY RLMARKS CRITERIA TITLE 7 .1 7.2 7.3 7.4 7.5 7.6 7.7 f ,
-- 4. Regulatory Guides (RG)
- a. RG 1.6 Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution X i X X Systems
- b. RG 1.7 Control of Combustible Gas Concentrations in Containment following a Loss-oi-Coolant X X Accident X Instrument Lines Penetrating
- c. RG 1.11 Primary Reactor Containment X X X X X X
- d. RG 1.22 Periodic Testing of Protection X X X System Actuation Functions X X X Seismic Design Classification X X X X X X SRP Section 3.10 l g, c. RG 1.29 j ,
- f. RG 1.30 Quality Assurance Requirements for the Installation. Inspec-
-a
's tion, and Testing of Instrumenta-tion and Electric Equipnent X X X X X X X j;
- g. RG 1.32 Use of IEEE Std 308 " Criteria for Class IE Electric Systems for Nuclear Power Generating X X X Stations" X Use in conjunction with
X X X X Indicatio.1 for Nuclear Power X X Plant Safety Systems
- 1. RG 1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems X X X X X X J. RG 1.62 Pbnual Initiation of Protection Actions X X X X X o
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I TABLE 7-1 (CONTINJED)
CRITERIA TITLE APPLICABILITY REMARKS 7.1 7.2 7.3 ;4 7_. 7.5 7.6 7.7
- k. RG 1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plant X X X X X X X
- 1. RG 1.68 Preoperational and Initial Startup Test Programs for ,
Water-Cooled Power Reactors X X X X X X X
- m. RG 1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 2 X X X X X X X
- n. RG 1.75 Physical Independence of Electric Systems X X X X
- o. RG 1.78 Assumptions for Evaluating the
. Habitability of a Nuclear Power Plant Control Room During a sa Postulated Hazardous Chemical i Release X X
-a p. RG 1.89 Qualification of Class IE Equip-
., ment for Nuclear Power Plants X X X X X X SRP Section 3.11.
$' q. RG 1.96 Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants X X
- r. RG 1.12 Instrumentation for Earthquakes X X
- s. RG 1.45 Reactor Coolant Pressure Boundary Leakage Detection Systems X .X
- t. RG 1.67 Installation of Overpressure Protection Devices X X
- u. RG 1.80 Pre-operational Testing of -
Instrument Air X X X X. SRP Section 9.
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TABLE 7-1 (CONTINUED) .
TITLE APPLICABILITY REMARKS CRITERI A 7.1 7.2 7.3 7.4 7.5 7.6 7.7 gp _ _ _ _ . ___
-d
Releases X
- w. RG 1.97 Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and following an Accident X X Seismic Qualification of SRP Section 3.10.
- x. RG 1.100 Electrical Equipment for X X X X Nuclear Power Plants X X
- y. RG 1.105 Instrument Spans and Setpoints X X X X X X
- z. RG 1.llB Periodic Testing of Electric Power and Protection Systems X X X X X X Fire Protection Guidelines for SRP Section 3.10.
[" aa. RG 1.120 X Nuclear Power Plants X X X X X X
-a 7a 3$ 5. Branch Technical Positions (BTP) ICSB
- a. BTP ICSB 1 Backfitting of the Protection and D0R Responsibility.
Emergency Power Systems of Nuclear X X X X X Reactors
- b. BTP ICSB 3 Isolation of Low Pressure Systems from the High Pressure Reactor .
Coolant System X X X
- c. BTP ICSB 4 (PSB) Fequirements on Motor-Operated Valves in the ECCS Accumulator X X Lines X
Technical Specifications X X
- e. BTP ICSB 9 Definition and Use of " Channel-Calibration" - Technical Specifications X X X X X
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TABLE 7-1 (CONTINUED)
TITLE APPLICABILITY REMARKS CRITERI A 7.1 7.2 7.3 7.4 7.5 7.6 7.7
- f. BTP ICSB 10 Electrical and Mechanical Equipnunt Seismic Qualification Program X X X X X Replaced by Reg. Guide 1.100
- g. BTP ICSB 12 Protection System Trip Point Changes for Operation with Reactor Coolant Pumps Out of Service X X X
- h. BTP ICSB 13 Design Criteria for Auxiliary Feedwater Systems X X
- i. BTP ICSB 14 Spurious Withdrawals of Single Control Rods in Pressurized Water Reactors X X X
- j. BTP ICSB 15 (PSB) Reactor Coolant Pump Breaker Quali fication X X ra k. BTP ICSB 16 Control Element Assembly (CEA)
' Interlocks in Combustion
- d Engineering Reactors X X 35 1. BTP ICSB 18 (PSB) App'.ication of the Single
Failure Criterion to Manually-Controlled Electrically-Operated X X X Valves X
- m. BTP ICSB 19 Acceptability of Design Criteria for Hydrogen Mixing and Drywell Vacuum Relief Systems X X X
- n. BTP ICSB 20 Design of Instrumentation and Controls Provided to Accomplish Changeover from Injection to Recirculation Mode X X X X
- o. BTP ICSB 21 Guidance for Application of Reg. .
Guide 1.47 X X X X X X
- p. .BTP ICSB 22 Guidance for Application of Req.
Guide .122 X X X X X X b
TABLE 7-1 (CONTINUED) y CRITERIA TITLE APPLICABILITY REMARKS 3 .
7.1 7.2 7.3 7.4 7.5 7.6 7.7 '
? _
- q. BTP ICSB 23 Qualification of Safety-Related Replaced by Reg. Guide 1.97.
Display Instrumentation for
- Post-Accident Condition Monitor- l ing and Safe Shutdown X X
- r. BTP ICSB 24 Testing of Reactor Trip Syston Replaced by Reg. Guide 1.118.
and Engineered Safety Feature Actuation Systm Sensor Response Times X X r X X
- s. BiP ICSB 25 Guidance for the Interpretation of General Design Criterion 37 for Testing the Operability of the Emergency Core Cooling System as a Whole X X X
- t. BTP ICSB 26 Requirments for Reactor Protec-tion System Anticipatory Trips X X Design C-iteria for Thermal Replaced by Reg. Guide 1.106
- u. BTP ICSB 27 7 Overload Protection for Motors of Motor-Operated Valves X
- [ X X X b
5 m
UNITED STATES OF A> ERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
PACIFIC GAS AND ELECTRIC COMPANY) Docke t No. 50-2 75 0. L.
) 50-323 0.L.
(Diablo' Canyon Nuclear Power )
P lan t , Unit Nos. 1 and 2) )
AFFIDAVIT OF DALE G. BRIDENBAUGH , RICHARD B. HUBBARD, AND GREGORY C. MINOR FOR JOINT INTERVENORS DALE G. BRIDENBAUGH , RICHARD B. HUBBARD, AND GREGORY C.
MINOR, being duly sworn, do say under oath that I, the undersigned have assis ted in preparing and reviewing responses of Joint Inter-venors to Pacific Gas and Electric Company's Third Set of Interrog-atories Nos. 1-18. Said answers are true and correct to the bes t of my knowledge and belie f.
Dale G. Bridenbaugh /
77
!$fu it 6L8$ W Richard B. Hubbard
/
October 30, 1981 / d@t/ b C Subscribed and sworn to before me this k[ day od3/vfgr , 1981. ~ o - se -o - o.=o - o - o ,
OFFICIAL SEAL I ,
- 7. ; -
(/ - ,
ij CARLO F. CARALU
__ ' b d& I ,
m
%=/
.5' < Notary Futi:c Ca'ifornia Prinopal Office in santa c: ara county .
NOTARY PUBLIC ur commission expus oct. 5. e4 g My commission expires : /d/S <f/
. _. . _ . _ __.m_ .. . - __
i .
1 UNITED STATES OF AMERICA j NUCLEAR REGULATORY COMMISSION t
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD t
)
In the Matter of )
l }
PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 O.L.
) 50-323 0.L.
(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
)
)
i I
CERTIPICATE OF SERVICE a
I hereby certify that on this 4th day of November, 1981, I have served copies of the foregoing JOINT INTERVENORS' RESPONSE TO NRC STAFF'S REQUEST FOR ADMISSIONS, RESPONSE OF JOINT l INTERVENORS TO SECOND SET OF INTERROGATORIES OF NRC STAFF, and i
RESPONSE OF JOINT INTERVENORF TO APPLICANT PACIFIC GAS AND l
ELECTRIC COMPANY'S THIRD SET OF INTERROGATORIES, mailing them i through the U. S. mails, first class, postage prepaid.
i Admin. Judge John F. Wolf, Docket & Service Branch l
Chairman Office of the Secretary Atomic Safety & Licensing U.S. Nuclear Regclatory Board Commission U. S. Nuclear Regulatory Washington, D.C. 20555 Commission
, Washington, D.C. 20555 William Olmstead, Esq.
Marc R. Staenberg, Esq.
- Glenn O. Bright Edward G. Ketchen, Esq.
Atomic Safety & Licensing Office of the Executive Legal Board Director - BETil 042 U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 vt-re ,+ ere--',p -
-r.*.ar = y w,-gy,m,p.y-,y-+,e---ew,c.,w-, .uw-.-gy---c ee-#--- -,,r- - - , -,-+----r -w1.,--r .o rm +e,.e-r---,,,--,-,,ye---e-
_ __ - - _ = _ _ - _- . . . - ..
O Dr. Jerry R. Kline Nancy Culver Atomic Safety & Licensing 192 Luneta Board San Luis Obispo, CA 93401 U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 Mr. Fredrick Eissler Malcolm H. Furbrush, Esq.
Scenic Shoreline Preservation Vice President and General Conference, Inc. Counsel 4 4623 More Mesa Drive Philip A. Crane, Esq.
Santa Barbara, CA 93105 Pacific Gas & Electric Company P. O. Box 7442 Sandra A. Silver San Francisco, CA 94106
- Gordon Silver 1760 Alisal Street Arthur C. Gehr, Esq.
San Luis Obispo, CA 93401 Snell & Wilmer
' 3100 Valley Center David S. Fleischaker, Esq. Phoenix, AZ 85073 i P. O. Box 1178
) Oklahoma City, OK 73101 Carl Neiburger Telegram Tribune Bruce Norton, Esq. P. O. Box 112 3216 N. Third Street San Luis Obispo, CA 93402 >
Suite 202 i Phoenix, AZ 85012 Byron Georgiou, Esq.
Legal Affairs Secretary to Janice E.-Kerr, Esq. the Governor Lawrence Q. Garcia, Esq. State Capitol Building
- J. Calvin Simpson, Esq. Sacramento, CA 95814 California Public Utilities Commission Lawrence Coe Lanpher, Esq.
5246 State Building Hill, Christopher & Phillips 350 McAllister Street 1900 M. Street, N.W.
- l San Francisco, CA 94102 Washington, D.C. 20036 MHB Technical Associates 1723 Hamilton Avenue Suite K San Jose, CA 95725 l
2 JO R. R8Y pLD$, ESQ.
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