ML20032C175
| ML20032C175 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/02/1981 |
| From: | Lanpher L CALIFORNIA, STATE OF, HILL, CHRISTOPHER & PHILLIPS |
| To: | PACIFIC GAS & ELECTRIC CO. |
| References | |
| ISSUANCES-OL, NUDOCS 8111090342 | |
| Download: ML20032C175 (33) | |
Text
,
4 RELATED CORRESPONDENCE I
UhITED STATES OF AMERICA 00LKETED NUCLEAR REGULATORY COMMISSION USHM EN BEFORE THE ATOMIC SAFETY AND LICENSING BOARD OFFICE cr 3ECRETAk :
00CKEilHG & SERVICE b^
In the Matter of
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)
Docket Nos. 50-275 0.L.
PACIFIC GAS AND ELECTRIC COMPANY
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50-323 0, 4
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f$M~I 1/
diV (Diablo Canyon Nuclear Power Plant,)
Full Power Proce(/((/
e' Unit Nos. 1 and 2)
)
NOVO G t
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19 8 I 0
RESPONSE OF GOVERNOR EDMUND G. BROWN, JR. ',
u., Mlg " C TO APPL 1.'% f PACIFIC GAS AND ELECTRIC COMPANY'"S-8 THIRD SET OF INTERROGATORIES lj'}
l Governor Brown hereby responds to Pacific Gas and Electric t
i Company's Third Set of Interrogatories dated October 15, 1981 as follows:
Interrogatory 1 State in detail each and every fact available to you which you believe shows that the proper operation of power operated relief valves and associated block valves is essential to mitigate the consequences of accidents at Diablo Canyon.
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
- 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 D503 Amendment to the FSAR as having been seismically tested (See J
(.)'
Hosgri Seismic evaluation, VOL. III, Table 7-7 " Seismic Qualification Minimum Required Active Valves for Hot Shutdown and/or Cold Shutdown.")
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l Hosgri Seismic Evaluation (Vol. III Table 7.8, "S umma ry -
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.
However, proper operation of power operated relief valves, associated block valves and the instruments and controls for these valves is essential to mitigate the consequences of accidents in that their 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 insufficient information to know if the existing valves and their associated equipment meet the necessary requirement to insure reliable performance of their safety function under worst case accident conditions.
Interrogatory 2 State in detail each and every fact available to you which~
you believe shows that the proper operation of the ins truments and controls for power operated relief valves and associated block valves is essential to mitigate the consequences of accidents at Diablo Canyon.
Restonse 2 See Response 1.
The failure of control and/or instruments could lead to failure of the associated valves, thereby causing or aggravating a LOCA.
Thus, the associateil controls and ins tru-ments for these valves mus t comply with applicable codes,
s tanda rds, and regulatory practices.
The _ _ _ _ _ _ - _ _ _ _ - _ - - _ _ - - _ _ _ _ _.
NRC Standard Review Plan (NUREG 75/087 Section 7, Table 7-1) identifies the' acceptance criteria for safety-related ins tru-mentation and control equipment which should be applied to these comp one nts.
A copy of this table is attached.
Until adequate details are provided on how the valves and components meet the above safety and acceptance criteria, I
there can be no assurance of their ability to perform properly in all off-normal and accident conditions.
Interrogatory 3 Describe in detail each and every way in which the failure of power operated relief valves, associated block valves or the instruments and controls for these valves can cause a LOCA at Diablo Canyon.
Response 3 In addition to the discussicn in Responses 1 and 2, there are conditions where the block valves and PORVs may individually or collectively constitute a potential break :n the reactor coolant pressure boundary.
Failure to opernte 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 niso serve as a mechanism for control and/or mitigation of accidera conditions wher called upon to operate in the bleed and feed mode ' n conjunction with Safety Injection).
Components which have this large an -..
s impact on pressure boundary integrity, accidents, and safety should be classed as safety-grade.
Examples ii.clude 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 ope ration.
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.
f'G6E response to Short Term Lessons Learned, February 29, 1980, page III-B-13.
(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 s cenario).
(d)
Although the normal procedures do not appear to call for use of the block valves or PORV's to shutdown s
4
the reactor and maintain it-in a safe shutdown condition, there are conditions where : hey may.be called upon to assist in maintaining the plant in.
a safe shutdown ccndition.
The TMI-2 accident and post-accident mitigation is such~an example.
.(e)
ATWS is not a design basis event for Diablo Canyon at this time.
Therefore, ATWS has not been protected agains t solely with safety grade equipment.
Interrogatory 4 Describe in-detail each and every way in which the failure of power operated relief valves, associated block valves or the ins truments and controls for these valves - can aggravate a LOCA at Diablo Canyon.
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 misleading information cannot be predicted with certainty.
Interrogatory 5 Does Governor Brown believe that all power operated relief valves and all associated block valves at Diablo Canyon mus t be classified as components important to s afe ty?
9 Response 5 Yes.
See also response to Interrogatory 6.
Interrogatory 6 Does Governor Brown believe that all. power operated relief valves and all associated block valves at Diablo Canyon must be required to meet all safety-grade design criteria?
Response 6 Yes.
See also response to Interrcgatory 5.
Interrogatory 7 If Governor Brown's answer to interrogatory number 5 is in the negative, please list with specificity which relief and block valves at Diablo Canyon should be classified as components important to safety.
Response 7 No applicable.
Interrogatory 8 If Governor Brown's answer to interrogatory number 6 is in the negative, please list with specificity which relief and block valves at Diablo Canyon should be required to meet all s afe ty-grade design criteria.
Response 8 Not applicable. t
Interrogatory 9:
Is it Governor Brown's belief that none of the power op-erated relief valves or associated block valves at Diablo Can-yon have met all safety-grade design criteri5.?
Response 9:
Diablo Canyon safety valves are classified as safety-grade and subj ected to the requirements aof 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, Table 7-7,. Seismic Quali-fication Minimum Required Active Valves for Hot Shutdown and/
l or cold Shutdown.")
The PORV's and bleck alves are not spe-i 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 classificatior. 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. _...
9 s
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 control and instruments for these valves must comply with appitcable codes, standards, etc.
The NRC St.andard 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 table is 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.
Interrogatory 10:
For each and every pow operated relief valve and/or as-sociated block valve at Diablo Canyon at Governor Brown be-lieves has not met all safety-grade design criteria state:
(a)
The location of each such valve.
(b)
T:e intended purpose of each such valve.
(c)
How each such valve's failure could cause a LOCA.
(d)
How each cuch valve's failure could aggravate a LOCA.
(e)
Each and every fact upon which you base your belief that l
the valve has not met all safety-grade design criteria.
l i
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.
(c) and (d)
See Response to Interrogatories Nos.
1, 2, and 3.
(e)
" See response to Interrogatory 9.
6 i. _.
In te rro gato ry 11 State each and every fact known to Governor Brown which would substantiate Joint Intervenors' allegation that the
" staff recognizes that pressurizer heaters and associated controls are necessary to maintain natural circulation at hot stand-by conditions."
Resuonse 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 Interrogatories for a lis t of emergency operating procedures that include the use of pressurizer heaters).
Further, plant safety may be affected by many things, not the least of which is the need to minimize the number of challenges to the total system integrity and 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 t o s a fe ty".
- Pressurizer heaters are needed for this capability.
In addition, the pressurizer heaters mus t also maintain physical integrity for the reactor coolant pressure boundary to be maintained.
- NUREG-0578, page A-2. -
4 Finally, the IE TMI-2 investigation team recommended the fo llowing:
"The pressuri:er heater syctem should be classified as safety grade which would i
assure emergency power availability ar.d i
protection from failures due to environ-o mental conditions."
(Recommendations of 1
TMI 2 IE Investi;ation Team, at 23; 3
emphasis added) 4 Thus, the inves tigation recommendation is virtually the I
l same as the ccatention.
Inte rro gatory 12 State each and every fact known to Governor Brown which would substantiate Joint Intervenors' allegation that pressurizer heaters and associated controls should be classified as components important to safe ty.
Response 12 l
See Response No. 11 concerning the need for. classification of the components as important to s afe ty.
Further, all com-l ponents of the pressurizer heater sys tem, including supports and interconnecting wiring should be required to meet the applicable safe ty-grade design criteria.
PGSE has responded i
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 PG6E's Answer to Interrogatory No. 41 as the 480 volt vital breakers 52-lG-72 4 -1H-74, ' control
- Applicant Pacific Gas 5 Electric Company's Answers to Joint
'Intervenors ' Second Set of Interrogatories, page 1 4 2.
-11
l t
switches and cable between the vital bus and the breakers.
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 sys tem, therefore, consists of the heaters themselves and their associated controls, along with interconnecting wiring and supports.
See PGSE January 26, 1981 submittal to NRC on Full Power License Requirement and associated Figures II.E.3.1-1 5 -2 for diagrams showing the. mponents contained within the nn pressurizer heater sys tem.
Interrogatory 13 Does Governor Brown believe that the pressurizer heaters and associated controls at Diablo Canyon do not meet any safe ty-grade design criteria?
Response 13 See " Applicant's Answers to Joint Intervenors' Second Set of Interrogatories" dated October 26, 1981, particularly Response 34 where the applicant clearly acknowledges that for Diablo Canyon the pressurizer heaters and associated controls are not classified "important to s afe ty".
l 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 5 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, c
l pages II.E-10 through 19..
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 information 10 available to evaluate compliance.
However, it i
is likely that non-compliances exis t for the foll, wing reasons :
1 a.
GDC 20 requires, among o ther things, that the protection system shall be designed "to initiate the operation of systems important t o s a fe ty. "
Standard Review Plan Table 7-1 extends the applicability of GDC 20 to all ins trumentation and control functions important to safety.
PG6E'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.
This complex procedure does not meet the automatic initiation requirements of GDC 20.
NUREG 75/087, Section 7, Table 7-1.
- Philip A. Crane to Frank J. Miraglia, January 26, 1981, page II E-14. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _
b 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 impo_ tant to s a fe ty, the requirement of GDC 1 (Quality 1
' standards and records) does not appear to have been applied.1/
Interrogatory 14 For each applicable safety-grade design criteria that Governor Brown believes. the pressurizer heaters and associated controls at Diablo Canyon do not meet, state:
(a)
The specific criteria.
l (b)
All facts upon which you base your allegation that each such criteria is not met.
Response 14 See Response 13.
-1/
We note that the classification of pressurizer heaters and-associated controls is currently the subject of Union of
' Concerned Scientists Contention 3 in the ongoing TMI re-s tart hearings (NRC docket 50-289).
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Interrogatory 15 State in detail each and every fact known to Governor Brown which would substantiate Joint Intervenors' allegation that the Applicant's proposal.to connect two out of four of the heater groups to the present on-site emergency power supplies does not provide an " equivalent or acceptable level of protection."
Response 15 The proposed arrangement addresses only the reliability of power supply to the pressurizer heaters.
The heaters and associated controls are. still subject to failures introduced through incomplete attention and lack of compliance with-the applicable safety-grade criteria (See Responses 11, 12, 13 and 14).
l.
Interrogatory 16:
Identify each and every document upon which you base any answer to the preceding interrogatories and for each such document:
(a)
State the name, author, and date of the document, (b)
Identify the interrogatory answer to which the document relates.
(c)
Identify the specific page(s) of the document which re.'ates to the answer.
Response 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 3ocuments' description included sufficient information to identify the documents, including the identification of the specific page(s) of the document which relate to each inter-rogatory response.
Interrogatory 17:
Identify each and every document or exhibit Governor Brown may introduce into evidence as re pects any per. ding con-tentions in these proceedings.
Response 17:
The term "any a ading" as related to contentions is am-biguous.
Likewise, the term "these" lacks the necessary spe-c.fic basis.
Accordingly, we cannot identify any additional documents or exhibits as set forth in this request.
However, _ _ _ _ _ _ -
assuming that this request is limited to the subj ects identi-j fied as " Contention 10" and Contention:12" in the current Diablo Canyon P
- power license proceeding, the documents or exhibits relied upra which Governor Brown may introduce into evidence are 'dentified 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 Sttff and PGSE repsonses to Governor Brown and Joint Intervenor interro-gatories.
All parties have access to the documents provided during discovery.
Further, such documents and exhibits will generally be referenced in the testimony of Governor Brown's witnesses which will be submitted to all parties in'this full-power proceeding prior t'o any hearings.
Interrogatory 18:
Identify each and every witness Governor Brown may call or subpoena to the hearing on the matters presently pending.
For 4
each such witness, state:
(a)
The name, occupation, address and telephone number of each such person and_whether that person may appear for you as a voluntary witness or as subpoenaed witness.
(b)
The field or science in which each such person is suf-l ficiently scooled to enable him to express opinion evi-l dence in this matter, if any.
l (c)
Whether such witness will base his opinion:
l (i) in whole or in part upon facts acquired personally by that person in the course of an investigation or examination as to the facts; or (ii) solely upon information provided that person by others.
(d)
The qualifications of each such person that would quali-fy that person, if possible, as an expert witness.
}
i '
(e)
If any such witness has made a personal investigation or examination relating to any of the facts or ba'ses set forth in the answers to preceding interrogatories, state the date(s) and nature of each such investigation or examination.
(f)
Each and every fact, and each and every document, photo-graph, report, item, or other tangible obj ect supplied or made available to each such person for purposes of formulating his opinicas in this matter.
(g)
Whether each such perscn has rendered written reports, regarding facts, bases, or opinions as respects your answers to the preceding interrogatories or that per-son's contemplated testimony.
If so, state:
(i) the date(s) of each such report; and (ii) the name and address of the custodian of each such report.
Response 18:
Assuming that this request is limited to the subjects i-dentified as " Contention 10" and " Contention 12" in the cur-rent Diablo Cany)n full power license proceeding, the identifi-cation of witnesses Governor Brown may call to testify was set f arth in " Governor Brown Identification of Witnesses for Full Power Proceeding" dated October 16, 1981.
At that time, the following three potential witnesses for the subject two conten-tions'were identified:
Dale G. Bridenbaugh, Richard B. Hubbard, and Gregory C. Minor.
The Governor will identify other witnesses in the future once the decision is made to present other wit-nesses.
At this time the Governor does not plan to subpoena any witnesses on " Contention 10" and " Contention 12".
(a) and (b)
Resumes for Bridenbaugh, Hubbard, and Minor which set forth the occupation, address, tele-phone number, and educational background were supplied on October 16. J
(c)
The bases for the potential witnesses' opin-ions are currently being deveolped.
It appears that the facts on which the witnesses will base their opinions will be a combination of facts concerning the Diablo Canyon b uclear Station acquired personally as well as information pro-vided the witnesses by others including the spe-cific documents provided by the NRC Staff and PGSE in response to discovery requests.
(d)
See October 16, 1981 filing.
(e)
At this time, the potential witnesses have con-
~
ducted the general personal investigations and examinations to develop the facts or bases set forth in the foregoing interrogatory responses.
The specific investigations and examinations des-cribed herein were conducted during October, 1981.
However, the bases for the -otential witnesses' testimony included the relevant experience in the nuclear industry summarized in the October 16, 1981 filing.
(f)
See Responses to Interrogatories 1 to 16.
(g )
The potential witnesses have rendered no written reports regarding facts, bases, or opinions other than set forth herein.
Further, Governor Brown - -
obj ects to this interrogatory in that any summary of the potential witnesses' testi-t many would be privileged as trial prepatory material.
See Kansas Gas and Electric Co.
j (Wolf Creek Nuclear Generating Station, Unit I
1), ALAB-327, 3 NRC 408 (1976).
t Respectfully submitted, Byron S. Georgiou Legal Affairs Secretary Covernor's Office State Capitol Sacramento, California 95814 h,% 1 C.- - -
b.e L
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s Herbert H. Brown Lawrence Coe Lanpher.
HILL, CHRISTOPHER AND PHILLIPS, P.C.
1900 M Ftreet, N.W.
Washington, D.C.
20036 Attorneys for Governor Edmund G.
Brown, Jr., of the State of California Dated:
November 2, 1981.
NUREG 75/087 c rea 3
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U.S. NUCLEAR REGULATORY COMMISSION MI 1, + " 3 STANDARD REVIEW PLAN
%..D OFFICE OF NUCLEAR REACTOR REGULATION TABLE 7-1 ACCEPTANCE CRITERIA FOR INSTRUMENTATICN AND CCNTRCLS l
Thesa l
Table 7-1 contairs the acceptance criteria for the SRP sect'ans of Chapter 7.
acceptance criteria include the applicable General Cesign Criteria. IEEE standards.
Regulatory Guides, and Branch Technical Posittens (BTP) of the Instr.: mentation and Control Systems Br:nch (ICSB). The applicability of these criteria to specific secticns of Chapter 7 is indicated of an X in t.e matrix listing of criteria and SAR sections. The BIP listed in Table 7-1 are conteined in Appencix 7. A to the Chapter 7 SRP secticn, f
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CRITERIA TITLE ffPLICABILITY REMARKS 7.1 1,2 7,3 7.4 7.5 M 7.7 l
1.
10 CfR 650.34 Contents of Application:
Technical Information X
'i X
X X
X X
l h.
10 CFR 550.36 Technical Specifications X
'X X
X X
X c.
10 CFR 650.55a Codes and Standards
.X X
X X
X X
X 2.
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GDC 1 Quality Standards and Records X
X X
X X
X b.
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X X
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X c.
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X X
X
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X X
X X
X f.
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X X
X X
X g.
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1 X
X h.
CDC 13 Instrumentation and Contrul X
X X
X X
X X
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X X
X X
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_X X
X X
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X X
X X
X l.
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X X
X X
X m.
_ Protection System Independence X
X X
X X
X n.
GDC 23 Protection System failure n> des X
X X
X X
X o.
GDC 24 Separation of Protection and Control Systems X
X X
X X
X X
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GDC 25 Protection System Requirements 5
for Reactivity ContrJ1 Malfunctions X
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ac n
e em l
nm t
nm o
e s
no e
oe A
oe r
o D
m aC a
m 1
on cb i
gO a
t r
me l
Ct Ct e
C e
l t
e l
a o
Et 1
Ca aa L
A 1
ep l
n l
C s
t s
t s
t t
t t og n
t e
yy Ra y
na o
f y n
f y n
f y a
f n
sn on o
s t c C
t on C
l y
oS e
oS e
oS W
o e
ye Ci C
y m
S n t
S i n d
i io l
c m
n va es v
ti r
a n
gg n
gl n
gp g
g n
n ra y
i ct o
u e
nn i
na i
nu n
n i
gi or r
d id nm t n ie t
ea t
d g
ii a
i v a
in i
i a
na tt a
e cu bt c
t r c
i r
tl t
t o t
t a l
t t
it ce n
so ad ms a
oe a
s e
so n
sm n
se o
s n
pn an i
ee oy e
rp e
e m
eo o
ee o
el o
e o
i o ee r
PO R
R E
TC C
TR C
TC C
l C
PC RP P
C A
I 1
3 4
6 0
4 5
6 7
R 6
7 8
9 3
4 5
7 8
0 E
2 2
2 2
3 3
3 3
3 4
4 4
4 4
5 5
5 5
5 T
I C
C C
C C
C C
C C
C C
C C
C C
C C
C C
R D
D D
D D
D D
D D
D D
D D
D D
D D
D D
C G
G G
G G
G G
G G
G G
C G
G G
G G
G G
q r
s t
u v
w.
a b
c d
e f
9 h
1 x
y z
a b
c d
e f
9 h
1 yh0 2
g'"
=
t TABLE 7-1 (CONTINUED)
CRITERIA TITLE APPLICABILITY REKARKS
]L, L 7.2 7.3 7.4 7.5 7.6 7.7 j
3.
Institute of Electrical and Electronics Engineers (IEEE)
Standards:
a.
IEEE Std. 279 Criteria for Protection Systems See 10 CFR ISO.5Sa(h)
( ANSI i442.7) for Nuclear Power Gen? rating and Reg. Guide 1.62.
Stations X
X X
X X
X X
b.
IEEE Std 308 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.
Testing Requirements for Instru-rentation and Electric Equipment During the Construction of gda Nuclear Power Generating Stations X
X X
X X
X X
- o. lF 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 ! Electrical Equipment SRP Section 3.10.
for Nuclear Power Generating Stations X
X X
X X
I g.
IEEE Std 379 Guide for the Application of the See Reg. Guide 1.53.
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.
IE Equipment and Circuits X
X X
X X
X X
E
?
C
1 TABLE 71 (CONTINUED)
CR11ERIA TITLE APPLICABIL11Y RLMARKS 7.1 7.2 '7.3 7.4 j[.j5 7.6 7.7 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.
RG 1.7 Control of Combustible Gas Concentrations in Contairment following a Loss-of-Coolant Accident X
X X
c.
RG 1.11 Instrument Lines Penetrating Primary Reactor Contairment X
X X
X X
X d.
RG 1.22 Periodic Testing of Protection System Actuation functions X
X X
X X
X l
e.
RG 1.29 Seismic Design Classification X
X X
X X
X SRP Section 3.10 ga jn f.
RG 1.30 Quality Assurance Requirements for the Installation..Inspec-i
-a tion, and Testing of Instrunenta-tion and Electric Eauipnent -
X X
X X
X X
X l
j; l
9 RG 1.32 Use of IEEE Std 308" Criteria for Class IE Electric Systems for Nuclear Power Genert. ting Stations
- X X
X X
Use in conjunction with l
h.
RG 1.47 Bypassed and Inoperable Status Indicatio.i for Nuclear Power X
X X
X X
X Position 3, RG 1.17.
Plant Safety Systems i.
RG 1.53 App 1! cation of the Single-Failure Criterion to Nuclear Power Plant Protection Systems X
X X
X X
X J.
RG 1.62 obnual Initiation of Protection Actions X
X X
X X
o l
l l
1
+
1.
1 TABLE 7-1 (CONTINUED)
CRITERIA TITLE APPLICABILITY RIMARKS 7.1 7.2 7.3 7.4 7.5 7.6 7.7 k.
RG 1.63 Electric Penetration Assemblies in Containment Struct es for I
Water-Cooled Nuclear Power Plant X
X X
X X
X X
1.
RG 1.68 Preoperational and Initial Startup Test Programr for Water-Cooled Power Reactors X
X X
X X
X X
m.
RG 1.70 Standard Fonr.at and Content l
of Safety Analysis Reports for Nuclear Power Plants, Rev. 2 X
X X
X X
X X
n.
RG 1.75 Physical Independ(nce of Electric l
Systuns X
X X
X l
O.
RG 1.78 Assumptions for Evaluating the Habitability of a Nuclear Power Plant Lontrol Roan During a ro Postulated Hazardous Chemical i
Release X
X cn RG 1.89 Qualification of Class IE Equip-
-4 ment for Nuclear Power Plants X
X X
X X
X SRP Section 3.11.
5 q.
RG 1.96 Design of Main Steam Isolation l
Valve leakage Control Systems l
for Boiling Water Reactor Nuclear Power Plants X
X
,[
r.
RG - 1.12 Instrumentation for farthquakes X
X i
s.
RG i.45 Reactor Coolant Pressure Boundary Leakage Detection Systesas X
X i
t.
RG 1.67 Installation of Overpressure l
Protection Devices X
X u.
RG 1.80 Pre-operational Testing of Instrument Air X
X X
X.
SRP Section 9.
I
-R!
?
a i
4
TABLE 7-1 (CONTINUED)
J CRITERIA TITLE APPLICABILITY
'RIMARKS i
7.1 7.2 7.3 7.4 7.5 7.6 7.7 22 l
,2 v.
RG 1.95 Protection of Nuclear Power Plant Control Room Operators Against Accidental Chlorine i
Releases X
X
~
w.
RG 1.97 Instrumentation for Light IJater Cooled Nuclear Power Plants to j
Assess Plant Conditions Durirg and following an Accident X
X x.
RG 1.100 Seismic Qualification of SRP Sec tion 3.10.
Electrical Equipsent for Nuclear Power Plant; X
X X
X X
X y.
RG l.105 Instrunent Spans and Setpoints X
X X
X X
I(
z.
RG 1.118 Periodic Testing of Electric Power and Frotection Systems X
X X
X X
X F"
aa.
RG 1.120 fire Protectior. Guidelines for SRP Section 3.10.
2a74 Nuclear Power Plants X
X X
X X
X X
3$
5.
Branch Tecnnical Positions (BTP) ICSB 1
4.
BTP ICSB 1 Backfitting of the Protectior and DOR Responsibility.
Emergency Power Systems of Nuclear Reactors X
X X
X X
Isolation of Low Pressure Systems b.
BTP ICSB 3 from the liigh Pressure Reactor Coolant System X
X X
c.
BTP ICSB 4 (PSB)
Requirenents on Motor-Operated Valves in the ECCS Accunulator Lines X
X X
d.
BIP ICSB 5 Scram Breaker Test Requirements -
1 chnical Specifications X
X e.
BIP ICSB 9 Definition and Use of " Channel-Calibration" - Technical Specifications X
X X
X X
l
~
-.--.q
~
\\
TABLE 7-1 (CONTINUED)
CRITERIA TITLE APPLICABILITY REMARKS 7.1 7.2 j[t3 7.4 7.5 7.6 7.7 f.
BIP ICSB 10 Electrical and Mechanical Equipnent Seismic Qualification Prograia X
X X
X Replaced by Reg. Guide 1.100 BTP ICSB 12 Protection System Trip Point g.
Changes for Operation with Reactor Coolant Pumps Out of Service X
X
,X h.
BIP ICSB 13 Design Criteria for Auxiliary Feedwater Systems X
X i.
BTf' ICSB 14 Spurious Withdrawals of Single Control Rods in Pressurized Water Reac; ors X
.I X
j.
BTP ICSB 15 (PSB)
Reactor Coolant Pump Breaker Quali fication X
X ra k.
BIP ICSB 16 Control Element Assembly (CEA)
Interlocks in Conbustion e
- e Engineering Reactors X
X 3[
1.
BTP ICSB 1B (PSB)
Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves X
X X
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 C$angeover 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 7s md a
4
TABLC 7-1 (CONTINUED) y CR!llRIA TITLE APPLICABILITY REMARKS,.
7.1 M 7.3 7.4 7.5 7.6 7.7 5
q.
BTP ICSB 23 Qualification of Safety-Related Replaced by Reg. Guide 1.97.
~
Display Instrumentation for
~
Post-Accident Condition Monitor-ing and Safe Shutdown I
X r.
blP ICSB 24 Testing of Reactor Trip Systen Replaced by Reg. Guide 1.118 and Engineered Safety feature Actuation Systen 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 Systen as a imole X
X X
t.
BIP ICSB 26 Requirunents for Reactor Protec-tion Systen Anticipatory Irips X
X H
u.
BIP ICSB 27 Design Criteria for Thennal Replaced by Reg. Guide 1.106 Overload Protection for Motort-
- [
of Motor-Operated Valves X
X X
X h
UNITED STATES OF AMERICA NUCLEAR REGULATORY. COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COMPANY)
Docket No. 50-2 75 0. L.
)
50-323 0.L.
(Diablo Canyon Nuclear Power
)
Plant, Unit Nos. 1 and 2)
)
AFFIDAVIT OF_
DALE G.
BRIDENBAUGH, RICHARD B. HUBBARD, AND GREGORY C. MINOR FOR GOVERNOR EDMUND G.
B ROWN, J R._
DALE G. BRIDENBAUGH, RICHARD B. HUBBARD, AND GREGOPY C MINOR, being duly sworn, do say under oath that I, the undersigned have assisted in preparing and reviewing responses of Governor Edmund G. Brown, Jr. to Pacific Gas and Electric Company's Third Set of Interrogatories Nos. 1-18.
Said answers are true and correct to the bes t of my knowledge and belief.
b Dale G.' Bridenbau'gh
/
l 2
Richard B.
Hubbard bd['
0etober 30, 1981
- 4/
Subscribed and sworn to before me this 8d day of[g[4e/
1981.
-<- -<=oa...oq l Q_. 4 0FFICIAL SEAL y
] fm)sa:.
.Q canto v. cAnnu 4,gj.
i 4
notsrv puc::c cas.tarnia
~
w pr.wpai onice in NOTARY ~PUBLIC santa crara county
?
u amrntsson notes oct. s,1934 g
j r
My commission expires :
Mk,/ [Y b """ ~' - an
=
_..,. _. ~, _ _
_,.. _ _ _. _.. _ _ ~.. _.... _ _, _ _..
/..
UNITED STATES OF RMERICA NUCLEAR REGULATORY COi01ISSION 00tKETED USHFa BEFORE THE ATOMIC SAFETY AND LICENSIh3 BOARP NW -4 P4 :47 I
0FFICE OF SECRETART In the Matter of
)
00CKETING & SERvlCL
)
BRANCH PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 O.L.
)
50-323 0.L.
(Diablo Canyon Nuclear Power
)
Plant, Unit Nos. 1 and 2)
)
)
f f
CERTIFICATE OF SERVICE I hereby certify that copies of " GOVERNOR BROWN'S RESPONSE TO NRC STAFF'S REQUEST FOR ADMISSIONS," " RESPONSE GF GOVERNOR EDMUND G. BROWN, JR. TO SECOND SET OF INTERROGATORIES OF NRC STAFF," and
" RESPONSE OF GOVER'IOR EDMUND G. BROWN, JU. TO APPLICANT PACIFIC GAS AND ELECTRIC COMPANY'S THIRD SET OF INTERROGATORIES" in the above-referenced natter have been served to the following on November 3, 1981 by U.S. Mail, first class.
Mr. Thomas Mcore, Chairman Atcmic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. W.
Reed Johnson Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. John H. Buck Atomic Safety and Licensinc Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20355 Chairman Atomic Safety and Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washingten, D.C.
20555 John F. Wolf, Esq., Chairman Atc=ic Safety and Licensing Board U.S. Nuclear Regulatory Commission Nashington, D.C.
20555 i
/.. Mr. Glenn O. Bright Atomic Safety and Licensing Soard U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Jerry R. Kline Atomic Safety and Licensing-Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 William J. Olmstead, Esq.
Edward G.
Ketchen, Esq.
Lucinda Low Swartz,_Esq.
Office of Executive Legal Director BETH 042 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 ATTENTION:- Docketing and Service Section Mrs. Elizabeth Apfelberg 1415 Cozadera San Luis Obispo, CA 93401 Janice E.
Kerr, Esq.
Public Utilities Commission 5246 State Building
--350 McAllister Street-San Francisco, California 94102 Mrs. Raye Fleming 1920 Mattie Road Shell Seach, California 93449 Mr. Frederick Eissler Scenic Shoreline Preservation Conference, Inc.
4623 More Mesa Drive Santa Barbara, California 93105 Mr. Gordon Silver Mrs. Sandra A.
Silver 1760 Alisal Street San Luis Obispo, California 93401 Jcel R.
Reynolds, Esq.
John R.
Phillips, Esq.
Center for Law in the Public Interest 10951 West Pico Boulevard Third Floor Los Angeles, California 90064 Bruce Norton, Esq.
Norton, Burke, Berry & Junck 3216 North Third Street, Suite 300 Phoenix, Arizona 85012 k
i + -
- Philip A. Crano, Jr., Eeq.
'F. Ronald _Laupheimar, Esq.
Richard F.
Locke, Esq.
Pacific Gas and Electric Company P.O. Box 7442 San Francisco, California 94106 David S. Fleischak'er, Esq.
P.O. Box 1178 Oklahoma City, Oklahoma 73101 Arthur C. Gehr, Esq.
Snell & Wilmer 3100 Valley Bank Center Phoenix, Arizona 85073 Mr. Richard B. Hubbard MHB. Technical Associates 1723 Hamilton Avenue, Suite K San Jose, California 95125 Mr. Carl Neiberger Telegram Tribune
~
P,0.
Box 112 San Luis Obispo, California 93402 Byron S. Georgiou, Esq.
Legal Affairs Secretary Governor's Office State Capitol Sacramento, California 95814
?
+.-
/
/
-./Jy/
/p r,
<,/
m Christopher B. Hanback HILL, CHRISTOPHER AND PHILLIPS, P.C.
1900 ;i Street, N.W.
Washington, D.C.
20036 November 3, 1981 i
.- -