ML20005C067
| ML20005C067 | |
| 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: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| References | |
| ISSUANCES-OL, NUDOCS 8111180349 | |
| Download: ML20005C067 (21) | |
Text
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- DOCHETED l!SNRC UNITED STATES OF AMERICA
~9 NUCLEAR REGULATORY COMMISSION i BEFORE THE ATOMIC SAFETY AND LICENSING BOARD"- r F ,- ;
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i In the Matter of )
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PACIFIC GAS AND ELECTRIC COMPANY ) Docket Nos. 50-275 O.L.
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, (Diablo Canyon Nuclear Power ) 4 Plant, Units 1 and 2) ) ,.
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NOV171981d-u.s.unsa M RESPONSE OF JOINT INTERVENORS TO II ,
SECOND SET OF INTERROGATORIES OF NRC STAFF b
On Octooer 14, 1981, the NRu Staff propounded interrog-atories to Joint Intervenors. We hereby respond to those discovery requests. e Response 1 All components of the pressurizer heater system, inc?uding supports and interconnecting wiring should be required to meet the i
applicable safety-grade design criteria. PG&E has responded 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 Apendix A to 10CFR50.1/ This is further defined in PG&E's Answer to Interrogatory No. 41 as the 480 volt vital breakers 52-lG-72
& -lH-74, control switches, and cable between the vital bus and the 1! Applicant Pacific Gas & Electric Company's Answer to Joint Intervenor's Second Set of Interrogatories, page 1 & 2.
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breakers.2/ This implies then that all of the rest of the pressurizer heater system has not been designed to meet the safety-grade design criteria listed above. The remainder of the system, therefore, consists of *be heaters themselves and their associated controls, along with int.. connecting wiring and supports. See PG&E January 26, 1981 submittal to NRC on Full Power License Requirements and associated Figures II.E.3.1-1 & -2 for diagrams showing the components contained within the pressurizer heater system.2/
Response 2 See Response A.8 to NRC Staff's Request for Admissions.
Response 3 See Response 4 below.
Response 4 10CFR50. 55a (a) (1) requires that structures, systems, and components shall be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. There can be no disagreement Lhat all systems at Diablo Canyon must meet the applicable safe.ty-grade design criteria. There may be some disagreement on what is applicable. If that is the thrust of this interrogatory, see Response 9 for a discussion on Diablo Canyon's failure to comply with applicable criteria. For specific reference to applicable codes, standards, etc., the NRC Standard Review Plan identifies the 2/ Ibid 1/, pages 16 & 17.
E/ Philip A. Crane to Frank J. Miraglia, January 26, 1981, pages II.E010 through 19.
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accpetance criteria for safety related instrumentation and control equipment.1/ The pressurizer heaters and controls should be evaluated in accordance with these criteria. A copy of this Table is attached.
Response 5 See Response A.8 to NRC Staff's Request for Admissions.
Response 6 The proposed arrangement addresses only the reliability of power supply to the pressurizer heaters. The heaters and associated equipment, instrumentation, controls, and supports are still subject to failures introduced through incomplete attention and compliance with the referenced safety-grade criteria.
Response 7 All of the applicable design criteria must be met. See SRP Table 7.1 attached hereto for specific GDC refercences.
Response 8 See Responses 4 and 9. ,
Response 9 Contention 10 does not state that the pressurizer heaters and associated controls fail to comply with specific details in the General Design Criteria, but rather that thic equipment has not been classified as safety-grade and, therefore, has not been required to .
meet the safety-grade design criteria listed. There is obviously no way to evaluate that compliance since PG&E has not submitted any detailed information on how theses. components do or do not meet the 1/ NUREG 75/087, Section 7, Table 7-1.
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specific criteria. This Interrogatory is therefore premature until sufficient detailed information is available to evaluate compliance.
However, it is likely that non-compliances exist for the following reasons:
- a. GDC 20 requires, among other things, that the protection systen shall be design "to initiate the operation of systems important to safety." Standard Review Plan Table 7-1 extends the applicability of GDC 20 to all instrumentation and control functions important to safety.1! PG&E'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 of an operator to a location three
_ floors down in the Auxiliary Building and verbal confirmation that such action has been taken.S! This complex procedure does not meet the automatic initiation requirements of GDC 20.
- b. None of the pressurizer heater system, other than the breakers, switches and portion of the bus connection cables identified in Response 1, have been qualified in accordance with GDC 2 (seismic and environmental 1! NUREG 75/087, Sction 7, Table 7-1.
S! Philip A. Crane to Frank J. Miraglia, January 26, 1981, page II E-14.
qualification) CDC 22 (protection system -independence , .
" separation") o.1 GDC 3 (fire protection.
- c. Since +h :e components have not been classified as ,
important to safety, the requirement of GDC 1 (Quality standards and records). do not appear to have been applied. ,
Response 10 >
The proposed system does not provide adequate assurance that the nitigating systems will be available and that the plant operators will be able to respond to the needs for maintenance of natural circulation capability. The failure to assure this capability imposes . undue risk to public health and safety.
Response 11 PG&E must demonstrate that the heaters and associated controls have been classified as " components important to safety" and that they meet the applicable safety-3rade design criteria.
Response 12 <
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 Ilosgri Amendment to the FSAR as having be.cn seismically tested (See llosgri Seismic evaluation, VOL. III, Table 7-7 " Seismic Qualification Minimum Required Active Valves for llot Shutdown and/or Cold Shutdown.") The PORV's and Block Valves are not specifically identified in the FSAR Section 3.2 tables but e
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 othar details of the classification and qualification of these three types of valves. -
However, proper operation o'f power operated relief valves, associated 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. The re fore , these valves must also be classified as safety-grade components and required to meet all safety-grade design criteria. There is insufficient infortation to know if the existing valves and their associated equipment meet the necessary requirements to insure reliable performance of their safety function under worst case accident conditions.
Similarly, the associated control 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-related instrumentation and control equipment which should be applied to these components. A copy of this table is attached. -
Until details are provided on how the valves and components meet the above safety and acceptance criteria, there can be no assurance of their adequacy to perform properly in all off-normal and accident conditions.
1 I
Response 13 l See RespcTse 12.
Response 14 In r 'ition to the discussion in Response 12, there are conditions where the block valves and PORVs may ind.'vidua11y or coll'ectively cons titute a potential break in the reactor coolant pressure boundary. Failure to operate correctly, in either opening or closing, may cause or aggrevate a small LOCA. The valves can also play an important role in mitigating
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the effects of an ATh'S accident. They may also serve as a mechanism for control and/cv mitigation of accident conditions when called upon to operate in the bleed and feed mode (in conjunction with Safety Inj ection) . Components which have this large an impact on pressure boundary integrity, accidents, and safety should be classed as safety-grade.
Response 15 See discussion in Response to Admissions B.1, B.2, B.3, B.7, B.10, B.17 and Interrogatories 12 and 14.
Responsc 16 Under normal circumstance and assuming proper operation of equipment, there may be no concern regarding these valves.
The deficiency in question in contention 12 is the classification and qualification of the valves and their associated controls and ins trumentation, without which there can be no assurancr that the public health and safety will be protected in off-normal and accident conditions.
Response 17 See Resnonse 12. _7_
Response 18 The devices of interest are those used to energize, control or monitor the operation of the PORV and block valves, as set forth on FSAR Figure'3.2-07 (sheet 2 of 4) and Figure 7.3-21 (sheet 1 of 2).
Response 19 ,
Proper and reliable operation of safety-grade valve cannot be insured unless their associated control and ins tru-mentation is also qualified, designed, ins talled and tes ted according to the applicable safety and acceptance ' criteria.
Response 20 See Response 12.
Response 21 Contention 12 does not s tate that the PORV and block valves and associated controls and instrumentation fail to comply with specific details in the General Design Criteria but rather that this equipment has not been classified as safety-grade and therefore has not been required to meet the safety grade design criteria lis ted. Until PGGE submits detailed information on how the components do or do not meet the specif1: criteria, the evaluation sugges ted by the interrogatory is no t possible . This interrogatory is therefore, premature until sufficient detail is availabic on which to i
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evaluate compliance. However, since some valves and components are c1carly not classed as safety grade, there is a high likelihood that deficiencies would be discovered.
Unless and until the valves and their associated controls and instrumentation are classified as safety-grade and details provided on their compliance with the GDC and other acceptance criteria, there is no assurance that the valves will actually meet the subject criteria. At this time there is insufficient detailed information available to complete the assessment.
Response 22 See Response 21.
Responss' 23 Refer to Responses to Admissions B.1, B.2, B.3, B.7, B.10, B.17 and Interrogatories 12 and 14.
Response 24 See Response 21.
Response 25 See Response 21.
Response 26 .
This interrogatory has been answered in earlier responses to Admissions and Interrogatories. See, for instance, Response 23.
_9_
Response 27 This interrogatory has been answered earlier but the main issue is that the full vange of environmental conditions be
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covered including where necessary, accidents which are not now considered design besis accidents. It is also essential that the results be directly applirable to the Diablo Canyon Site and its as-built configuration.
Response 28 .
In a previous filing (dated November 3, 1981) on the Diablo Canyon Docket, witnesses have been identified.
Response 29 For each ?' mission Statement not admitted by Joint Intervenors, th' specific cc cern has been addressed in the Response to the Admission Statement.
Response 30 L See Response 4.
Response 31 See Responce 4.
Response to Document Requests All documents referred to in the above responses are on the Diablo Canyon docket or are in the public domain or are attached
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hereto. Thus, no additional documents are provided.
DATED: November 4, 1981 Respectfully submitted, JOEL R. REYNOLOS, 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 JOEL R. REYNOLDS Attorneys for Joint Inter-venors SAN LUIS OBISPO MOTHERS FOR PEACE
. SCENIC SHORELINE PRESERVATION CONFERENCE, INC.
ECOLOGY ACTION CLUB SANDRA SILVER ELIZABETH APFELBERG JOHN J. FORSTER 4
.' ATTACllMENT 1 NUREG 75/087 yganeo%
- . t U.Se NUCLEAR REGULATORY COMMISSION i?s! STANDARD REVIEW PLAN W'%!..DOFFICE OF NUCLEAR REACTOR REGULATION TABLE 7-1 l
ACCEPTANCE CRITERIA FOR INSTRUMENTATION AND CONTROLS l
Table 7-1 centains the acceptance criteria for. the SRP sections of Chapter 7e These acceptance criteria include the applicable General Design Criteriae IEEE standards.
Regulatcry Gaides, and Branch Technical Positions (BTP) of the Instrumentation and contrcl Syste s Branch (ICSB)e The applicar.ility of these criteria to spqcific secticns of Chapter 7 is indicated by an X in the matrix listing of criteria and SAR sections
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ACCEPT ANCE CRITERI A FOR INSTRtJMENTAT10.4 AND CONTROL SYSTEMS - TABLE 7-1 l CRITE RI A TITLE APPLICABILITY REMARKS 7.1 7,2 7.3 7.4 7.5 7.6 7.7 i 1. 10 CFR Part 50
- a. 10 CFR 150.34 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
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- b. GDC 2 Design Bases for Protection Against Natural Phenomena X X X X_. X X
- c. GDC 3 Tire Protection X X X X X X
- d. GDC 4 Environmental and t11ssile 3 7d Design Bases X X X X X X
][ e. GDC 5 Sharing of Structures, Systems,
, and Components X X X X X X
- f. DGC 10 Reactor Design. 3 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
- 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 7 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 EE p. GDC 25 Protection System Requirements
? for Reactivity Control Mal functions X X X 4
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I B4 A X X X X X C7 I
L P3 X X X X X X X X X X X X X X P X A7 2 X X X X X X
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r l ee oy e rp e e m eo o el C T C PC RP P C RR CS R PO R R E TC C TR C TC A
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t TABLE 7-1 (CONTINUED)
CRITERIA TITLE APPLICABILITY REMARKS 7.1 7.2 7.3 7.4 - 7. 5 j[.j i 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. Guide 1.62.
Stations X X X X X X X
- b. IEEE Std 303 Criterta 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 g' ;a Nuclear Power Generating Stations X X X X X X X a= I' 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 1 Electrical Equipment SRP Section 3.10.
for Nuclear Power Generating Stations X X X X X X
- g. lEEE Std 379 Guide for the Application of the See Reg. Guide 1.53.
(ANSI N41.2) Single 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 .
See Reg. Guide 1.75.
(ANSI N41.14) IE Equipment and Circuits X X X X X X X 5'
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TABLE 7-1 (CONTINUED)
APPLICABILITY REtiARKS 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 X X X
Systems
- b. RG1.7 Control of Combustible Gas Concentrations in Containment following a Loss-of-Coolant X X X Accident
- c. RG 1.11 Instrument Lines Penetrating X X X X Primary Reactor Containment X X
- d. RG l.22 Periodic Testing of Protection X X X X X System Actuation functions X X X X SRP Section 3.10 l
- c. RG 1.29 Seismic Design Classification X X X g,
j, f. RG 1.30 Quality Assurance Requirements for the Installation, inspec-
'u tion, and Testing of Instrumenta- X X X tion and Electric Equipment - 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 X Stations" Use in conjunction with
- h. RG 1.47 Bypassed and Inoperable Status X Position 3, RG 1.17.
X X X X Indicatio.i for Nuclear Power X Plant Safety Systems
- i. RG 1.53 Application of the Single-failure Criterion to Nuclear Power Plant X X X X Protection Systems X X J. RG 1.62 Manual Initiation of Protection X X X X X Actions o
s ,
DBL _E 7-1 (Cg4TINUED)
CRITE RI A TITLE APPLICABILITY REMARKS 7.1 7.2 7.3 7.4 7.5 7.6 7.7 i
- 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-Cocied Power Reactors X X X X X X X
- m. RG 1.70 Standard Fortnat 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 'a
- o. RG 1.78 Assumptions for Evaluating the Habitcbility of a Nucle."r Power Plant Control Room During a s Postulated Hazardous Chemical Release X X c'n y r- RG 1.89 Qualification of Class IE Equip-ment for Nuclear Power Plants X X X X X X SRP Section 3.11.
'7 O q. RG 1.96 Design of Main Steam Isolation Valve L(akage Control Systms for Boiling Water Reactor Nuclear Power Plants X X Instrumentation for Earthquakes X X~
- r. RG 1.12
- s. RG 1.45 Reactor Coolant Pressure Boundary Leakage Detection Systcyns X .X
- t. RG l.67 Installation of Overpressure Protection Devices ' X X
.all
.nd
TABL E 7-1 (CONTINllED)
AF PLICABILIT REtiARKS CRITERIA TITLE 7.1 7.2 7.3 7.4 7.5 7.6 7.7 y I
- v. RG 1.95 Protection of Nuclear Power Plant Control Room Operators Against Accidental Chlorine X X
Releases
- w. RG 1.97 Instrumentation for Light Hater Cooled Nuclear Power Plants to Assess Plant Conditions During X
and following an Accident X SRP Section 3.10.
- x. RG 1.100 Seismic Qualification of Electrical Equipment for X X X X X X Nuclear Power Plants Instrument Spans and Setpoints X X X X X T
- y. RG 1.105
- z. RG 1.118 Periodic lesting of Electric ; I X
Pcwer and Protection Systems X X X SRP Section 3.10.
aa. RG 1.120 fire Protection Guidelines for 7 Nuclear Power Plants X X X X X X X 9
5
- 5. Branch Technical Positions (BTP) ICSS DOR Responsibility.
- a. BTP ICSB 1 Backfitting of the Protection and Emergency Power Systems of Nuclear X X X X X Reactors
- b. BTP ICSB 3 1 solation of Low Pressure Systems from the High Pressure Reactor X X X Coolant System
Lines
Technical Specifications X X
- e. BTP ICSB 9 Definition and Use of " Channel-Calibration" - Technical X X X X X
Specifications a
N-
TABLE 7-1 (CONTINUED)
APPLICABILITY REMARKS CRIT ERI A TITLE 7.1 7.2 7.3 7.4 7.5 j[,6_ ]67 __
1
- f. BTP ICSB 10 Electrical and Mechanical Equipment Seismic Qualification Replaced by Reg. Guide 1.100 X X X X X Program
- g. BTP ICSB 12 Protection System Trip Point Changes for Operation with Reactor Coolant Pumps Out of X X X Service
- i. BTP ICSB 14 Spurious Withdrawals of Single Control Rods in Pressurized X X X Water Reactors
- j. BTP ICSB 15 (PSB) Reactor Coolant Pump Breaker X X Qualification _
BTP ICSB 16 Control Elcment Asser.bly (CEA) ra k.
' Interlocks in Combustion X X
- d, Engineering Reactors 2, 1. BTP ICSB 18 (PSB) Application of the Single
"' Failure Criterion to Manually-Controlled Electrically-Operated X X X X Valves
- m. BTP ICSB 19 Acceptability of Design Criteria for Hydrogen Mixing and Drywell X X Vacuum Relief Systems X
- n. BTP ICSB 20 Design of Instrumentation and Controls Provided to Accomplish Changeover f-om Injection to X X X X Recirculation Mode
- o. BTP ICSB 21 Guidance for Application of Reg. ,~
X X X X X X Guide 1.47
- p. .BTP ICSB 22 Guidance for Application of Req.
X X X X X X Guide .122 0
TABLE 7-1 (CONTINUED)
TITLE APPLICABILITY REMARKS ,s .
vs CRITERIA 2 7.1 7.2 7.3 7.4 7.5 7.6 7.7 '
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- q. BTP ICSB 23 Qualification of Safety-Related Replated by Reg. Guide 1.97.
Display Instrumentation for ;
- Post-Accident Condition Monitor- l ing and Safe Shutdown X X Testing of Reactor Trip Syston Replaced by Reg. Guide 1.118.
- r. BTP IC5B 24 and Engineered Safety feature Actuation System Sensor Response Times X X X X X
- s. BTP ICSB 25 Guidance for the Interpretation of General Design Criterion 37 for Testing the Operability of the Emergency Core Cooling System X X as a Whole X
- t. BIP ICSB 26 Requirunents for Reactor Protec-tion System Anticipatory Trips X X Design Criteria for Thennal Replaced by Reg. Guide 1.106 u.
['y3 ,, BTP ICSB 27 Overload Protection for Motors X
of Motor-0perated Valves X X X
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0!e11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
PACIFIC GAS AND ELECTRIC COMPANY) Docket No. 50-275 0.L.
) 50-323 0.L.
(Diablo Canyon Nuclear Power )
Plant, Unit Nos. 1 and 2) )
AFFIDAVIT OF DALE G. BRIDENB AUGH , RICHARD B. HUBBARD, AND GREGORY C. MINOR FOR JOINT INTERVENORS DALE G. BRIDENBAUGH, RICHARD B. HUBBARD, AND GREGORY C.
MINOR, being duly sworn, do s ay under oath that I , the undersigned have assisted in preparing and reviewing responses of Joint Inte rvenors to NRC Staff's Second Set of Interrogatories No. 1-31.
Said answers are true and correct to the bes t of n:y knowledge and belief.
< l Dale G. Bridenbaugh AGM b $Y Richard B. Hubbard
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O ctobe r 30, 1981 -
4&tu Gregory C/ / Minor Subscribed and sworn to before me this hd day of hf/fler , 1981. .,m ,_,_om,,mome,g
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l T. OFFICIAL SEAL $
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CARLO F. CARALU (*
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w_,s. .,aos ou. s. m My commission expires : N/6 M 9
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