ML20040A969
| ML20040A969 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/11/1982 |
| From: | Bridenbaugh D, George Minor CALIFORNIA, STATE OF |
| To: | |
| Shared Package | |
| ML20040A967 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8201230018 | |
| Download: ML20040A969 (12) | |
Text
732.m CC7.T.5T0%W 1
EXHIBIT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY C0lo!ISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
)
In the Matter of
)
Nos. 50-275 0.L.
PACIFIC CAS AND ELECTRIC COMPANY
)
Docket 50- 323 0.L.
)
)
(Diablo Canyon Nuclear Power
)
Plant, Unit Nos.1 and 2) g
)
'b
'd AFFIDAVIT OF DALE G. BRIDEt*B AUGh AND GREGORY C. MINOR j
JAt!151982 > ~
d: $$:
0 STATE OF CALIFORNIA
)
ECh
)
ss' O
COUNTY OF SANTA CLARA
)
cui g nu r DALE G. BRIDENBAUGH and GREGORY C. MINOR depose and say I
under oath as follows :
A statement of my qual'-
is Dale G. Bridenbaugh.
My name ifications and experience has previously been provided to this Board as part of my testimony on Contention 10 and in Attach-ment A to that testimony.
A statement of my qual-My name is Gre gory C. Minor.
ifications and experience has previously been provided to this l
Board as part of my testimony on Contention 1 and in Attachment B 4
e 5,
8201230018 820114 PDR ADOCK 05000275 C
imm to that testimohy.
This affidavit relates to Joint Intervenors Contentions 10 and 12 as set forth in the ASLB Prehearing Conference Order of February 13, 1981 and specifically responds to the " State-ment of Material Facts as to Which There Is No Genuine Issue" contained in the December 21, 1981 NRC Staff Motion for. Summary Disposition of Contentions 10 and 12.,
We attest that:
1.
Classification of the pressurizer heaters at Diablo Canyon as " components important to safety" has not been establish-ed.
While the NRC Staff claims that they have been, Affidavits by the Applicant's personnel state that:
"There are no requirements for the pressurizer heaters and associated c' ontTols to be classi-fied as ' components important to s afety. '",1/ and:
"....the pressurizer heaters and ass oc.iated controls are not classified as ' components im-por tan t to s a fe ty. ' " 2/
It.therefore appears that the NRC Staff considers the pressur izer heaters as " components important to safety" but the Appli-cant has not treated them as such.
2.
The pressurizer heaters do perform critical functions identified in 10 CFR 100, Appendix A, Section III (c).
Physical 1/
Pacific Gas and Electric Company's Motion for Summary Dispo-sition, Affidavit of John B. Hoch, p. 2, paragraph 6..
2/
Ibid 1, paragraph 7.
e.
integrity of -the heaters is required to preserve the " integrity of the reactor coolant pressure boundary."
Post-accident de-cay heat removal via the natural circulation mode is a function, required and is normally achieved via and is specified in the Emergency Operati' g Procedures to be per. formed by the pressuri-n zer heater system.
3.
Pressurizer heaters are no.rmally utilized in con-trolling reactor pressure while bringing the plant to cold shut-down.
4.
Failure of the pressurizer heaters to operate would allow the reactor system to depressurize at essentially an un-controlled rate unless additional equipment is brought into operation in a mode not normally utilized and which has not been clearly defined in the: Emergency Dperating_ Procedures.
5.
The pressurizer heater system is the normal system utilized to control the primary reactor pressure.
The pres-surizer heater,s have been designated as "important to safety" by the NRC Staff,2/ they have been recommended to be upg'raded in numerous NRC studies and reports, and are recognized as being of importance to the reduction of challenges to the other safety sys te ms.
6.
Two manual transfer switches with associated safety-related protective devices h' ave been provided to connect the 3/
NRC Staff Motion for Summary Disposition, 12/21/81, p. 6.
3-
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pressurizer' heaters to on-site standby power supplies.
They a:
not, ilowever, operable from the control room as was recommendec~
by the TMI Action Plan.b/
In addition, the safety-related de-vices serve only to protect the on-site power system rather thz i
also protecting the pressure control function.
7
'Diablo Canyon has U-tube steam generators.
The Ap-plicant has not yet, howe ve r, demonstrated the adequacy of natt ral circulation through these steam generators at Diablo Canyor
~
under adver'se pressure control conditions.
8.
The fact of the high points of the coolant loops being normally covered with secondary coolant supplied by main or auxiliary feedwater systems does not, of itself, assure ade-quate cooling of the core.
Other systems must be operable, op-erator actions must not interfere with the system's necessary function, and conditions conducive to maintenance of natural circulation must be present.
This has not been demonstrated a-Diablo Canyon nor have the Emergency Operating Procedures been fully and adequately prepared.
s 9.
The condensation of steam in the coolant loops witl no loss of natural circulation has not been demonstrated at Di:
blo Canyon.
10.
If sufficient steam were present; reactor coolant conditions would change from single phase natural circulation 4/
NUREG-0737, Clarification of TMI Action Plan Requirements,
- p. 3-86.
4
to some two-phase ' mixture.
If adequate cooling is provided, it would achieve a two-phase boiling condensation condition.
11.
Loss of natural circulation could be blocked in U-tube steam generators if secondary cooling to the steam gener-ators is inadequate.
12.
Natural circulation tests performed at the LOFT' and Semiscale, facilities have not been. shown to be directly ap.-
plicable at Diablo Canycn through actual demonstrations at that plant.
13.
The safety classification of PORV's and block valves and their associated instruments and controls is not clearly defined in the FSAR for Diablo Canyon, nor is it clear what the Applicant means by his use of the term "important to safety" in responses to interrogator [es on valve classification.
Thus, there is no assurdnce the valves are properly classified or qualified for their function.
14.
The PORV's and block valves are called upon in Emergency Operating Procedures to perform functions re.la'ted to' insuring the integrity of the reactor coolant pressure boundary (both for low temperature over-pressure conditions and operating and accident conditions).
However, contrary to our belief, the Applicant and Staff do not consider this as a safety function.
15.
The functions of the PORV and block valves include the following:
G e,
e e
a.
Maintain integrity of the primary
~pressure boundary.
b.
Provide pressure relief for. Low Temperature Overpressurization conditions.
~
c.
Reduce the number of challenges to the safety valves.
d.
Redu'ce the number of challenges to the ECCS.
Provide a blee'd capability 'during -
e.
the feed-and-bleed mode of operation to remove decay heat from the core (as, for example, was done during the TMI-2 accident).
Several of these functions are consistent with the functions in 10 CFR 100, Appendix A, Section III.C, which was used by the NRC to define criteria for " safety-related" classification.
- However,
\\
the Applicant contends PORV's and bkock valves are not relied upon for safety functions.
16.
The block valves are used to isolate a PORV and may also be used to provide throttling capability for back-up reactor coolant pressure control and for control of the bleed capability-in the bleed-and-feed mode of heat removal following an accident.
The Applicant does not consider these as safety functions; we disagree with this position.
17.
As the. accident at TMI-2 demonstrated, proper opera-tion of PORV's and block valves can be important in mitigating the e ffects of an accident.
They are also called upon in. the e,
\\
-ee e
.a.
Emergency Operating Procedures to provide a means for depres-surizing the reactor coolant so that back-up boration techniques may be applied.
The E0P's also assumIe the PORV's will automa-tically open in an ATWS event, an event which could lead to a major accident although not presently recognized as a design basis event.
Block valves are also used to mitigate and con-trol a small LOCA resulting from a f ailed PORV.
Despite these facts, the Applicant and Staff contend the block valves are not required to. mitigate the consequences cf a DBA.
18.
If a PORV failed it would cause a small LOCA.
If two or more failed due to a common-mode failure or systems inter-action, the effects would be more severe.
If the failure should occur simultaneously with a LOCA of other origin it would pro-duce confusing symptons and indicatEons to the operator, release additional contaminanted coolant to the containment and could re-sult in more severe consequences than a LOCA would otherwise pro-duce.
The Staff. contends that the simultaneous LOCA and failure We '
of a PORV would not significantly alter the consequences.
believe the impact could be significant.
19.
An unisolated stuck-open PORV was the fundamental cause of coolant loss leading to core damage in the TMI-2 acci-dent.
Thus, it is impossible to assure that stuck-open PORV's at Diablo Canyon could not lead to core damage.
Oaly under the most ideal conditions (i.e., ignoring systems interaction, e
common-mode faildres, operator error, and other system failures) can the Staff and Applicant assume no fuel damage will result from a stuck-open PORV.
We feel this is an unicasonable as-sumption.
20.
The pressure trip settings for the PORV's is slight-ly lower than tha't of the safety valves in order to reduce the number of challenges to the code safety valves.
We -consider this to be a safety-related function but Staff apparently dis-a grees.
21.
Several Emergency Operating procedures include descriptions of how the operators should go about searching for possible sources of coolant loss; specifically, they instruct operators, in several E0P's, to check for indications of leak-ing or open PORV's.
The operator would then take corrective ac-tiog such as closing the block valve.
If corrective action is not taken and there was continued leakage without make-up, the coolant pressure and level would drop and core cooling should
{
be automatically initiated.
If ECCS is not initiated.or if operator action precludes the continuted operation of a coolant source (as occurred at TMI-2), there is no assurance that proper core cooling will occur.
22.
Although the operator can isolate a stuck-open PORV by utilizing the block valve, he must first recognize the necessary
~
symptoms, properly diagnose the problem, and then take the proper.
e.
e_
action.
As has been shown by the experience at TMI, these steps can not be assured when the operators have only partial or misleading information or are predisposed to looking for a different causal event.
N s.
G e gory C. f ' nor
/
y January 11, 1982 Dale G. Bridenbaugh Subscribed and sworn before me this
//
day of d aecrq 1982.
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EXHIBIT 3 l
UNITED STATES OF AMERICA l
I NUCLEAR REGUIATORY COMMISSION 1
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i
)
In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COliPANY
)
Docket Nos. 50-275 0.L.
l
)
50-323 0 7..
(Diablo Canyon Nuclear Power
)
Plant, Units 1 and 2)
)
l
)
N sp
(
1
/
1 b
-4 i-PREPARED DIRECT TESTIMONY OF
?
JAtl i 51982, p s Y.:".f -7. h
- i$
DALE G. BRIDENBAUGH AND GREGORY C. MINOR s6 i'::.;h ON BEHALF OF GOVERNOR EDMUND G.
BROWN JR.
/, 'I.' L,'
s -
REGARDING CONTENTION 10 January 11, 1982 l
9
- e
c...
i 1
" PREPARED DIRECT TESTIMONY i
OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR REGARDING CONTENTION 10 i
I.
INTRODUCTION l
t-1.
My name is Dale G. Bridenbaugh.
I am a Professional Nuclear Engineer, licensed by the State of California, technical u
i co-founder and president of MHB Technical Associates, l
N l
consultant, technic consultants on energy and environment, with offices at l'23 Hamilton Avenue, Suite K, San Jose, California.
I have participated as an expert witness in licensing proceedings before i
I tne U.S. Nuclear Regulatory Commission (NRC) ; have served as a consultant to the NRC; have testified at the request of the Advisory Committee on Reactor Safeguards; have appeared before various committees of the U.S. Congress; and testified in various i
I received a Bachelor l
state licer. Jing and regulatory proceedings.
of Science in Mechanical Engineering from the South Dakota School i
I I
of Mines and Technnlogy in 1953.
From June, 1953, until February, I worked as an engineer and manager with the General Electric
- 1976, Company on a side variety of most of the aspects of power genera-tion equipmeat design, manufacture and operation.
During the I was in management positions in the last 10 of those 22 years, General Electric Nuclear Energy Division where i had the responsi-bility for managing the monitoring of operation of nuclear 1
l
power plants, for the implementation of solutions to nuclear plant operational problems, and for the development of a master performance improtement plan aimed at bringing about the long i
term improvement of power reactor perfornance.
2.
In my capacity as technical consultant'with MHB Technical Associates, I have provided technical advice to various gcVern-mental bodies and individual groups on subjects related to the design and operation of commercial nuclear power plants.
As examples of this work, in 1978 I served as a consultant to the United States Nuclear Regulatory Commission to review the NRC plan for research to improve the safety of light water nuclear power plants, and have served in var'ious* consulting capacities to the United States General Accounting Office, the states of California, Illinois, Massachusetts, New Jersey, Pennsylvania, to Suffolk County, New York, and to the governments of Sweden and Norway, all in the evaluation of nuclear plants or programs.
A statement of my qualifications and professional experience is appended to this testimony as Attachment A.
3.
My name is Gregory C. Minor.
A statement of my qualifications and experience has previously been provided to this Board as part of my testimony on Contention 1 and in Attachment B to that testimony.
l l
II.
STATEMENT OF CONTENTION 4.
The purpose of our testimony is to respond to Contention 10 as admitted by the Board as follows:-1/
The Staff recognizes that pressurizer' heaters and associated controls are necessary to maintain natural circulation at hot stand-by conditions.
Therefore, this equipment should be classified as ' components important to safety' and required to meet all applicable safety-grade design criteria, including but not limited to diversity (GDC 22),
seismic and environmental qualification (GDC 2 and 4), automatic initiation (GDC 20), separation and independence (GDC 3 and 22), quality assurance (GDC 1), adequate, reliable on-site power supplies (GDC 17) and the single failure criterion.
The Applicant's proposal to connect two out of four emergency power supplies does not provide an equivalent or acceptable levei of protection.
The results of our review of some of the important matters en-compassed by this Contention are summarized in the following paragraphs.
III.
DISCUSSION OF ISSUES III.A.:
Background and Summary of Position 5.
The essence of Contention 10 is that the pressurizer heaters, including the associated heater controls, should be 1/
ASLB Memorandum and Order, September 30, 1981.
On September 21, 1981, the Commission directed the Licensing Board to include in the full power proceeding Joint Intervencrs' low power Contention 10.
-4_
formally classified as " components important to safety" and, accordingly, be designed, manufactured, and constructed with all the care that should be afforded such components.
i 6.
The crigin of this Contention is the experience of the-Three Mile Island accident and the subsequent reviews performed to consider its significance.
This accident, along with its I
extended recovery period, demonstrated the need to reconsider the safety classifications and design practices for nuclear systems and components.
In particular, the inoperability of the reactor coolant pumps and the low pressure decay heat removal systems emphasized the importance of the ability to remove heat from the reactor via natural circulation and required assocIatsd'
~
systems.
Thus, the NRC Lessons Learned Task Force found that
" maintenance of natural circulation capability is important to safety."-2/ The pressurizer heater system is the normal and pre-ferred system for this capability.
In addition, the pressurizer heaters must also maintain physical integrity for the reactor coolant pressure boundary to be maintained.
While it may be possible to maintain natural circulation at hot standby condi-tions without use of the pressurizer heater and associated controls, such operation may be difficult to control and is contrary to the normal and emergency plant operating procedures.
In this regard, pG&E's response No. 45, dated october 26, 1981, 2/
NUREG-0578, p. A-2.
l
Second Set of Interrogatories provided to Joint Intervenors' ld the a list of emergency operating procedures that inc u e We have reviewed these procedures use of pressurizer heaters.
(to the use of the pressurizer heaters)
~
and find that " alternate" rs' pressure control metnods are not specified for the operato These procedures thus appear to place total reliance on use.
We therefore automatic or manual operation of the heaters.
- ified, that the heater system has been improperly class conclude:
in failing to or the procedurer, have been inadequately preparedbe at fault.
provide safety-related backup systems, or both may t the Further, plant safety may be affected by many things, no
+
least of which is the need to minimize the number of challenges h
rability to the total system integrity and to optimize t e ope i
or and controlability of the systems used in the mitigat on The logical response to the informa-control of abnormal events.
in our opinion, is to tion gained from the TMI-2 accident, t to safety clhssify the pressurizer heater system as importan so as to ensure its operability for response (safety-related) to accidents or transient conditions'.
dd It is important to place in proper context the inten e 7.
f Contention meaning of the phrase " components important to sa ety.
1981.
On 10 was formally accepted by the ASLB on September 30, 1981, Harold R. Denton, Director of the NRC's Office November 20, larify the of Nuclear Reactor Regulation,' iss'ued a Memorandum to c w
-~..s
3/
use of safety classification terms.
This Memorandum stated
~
that safety classification terms had not been consistently rpplied by the NRC S taf f, and that three terms, "important to safety,"
" safety-grade," and," safety-related," have been used inconsistently i
or interchangeably.
Mr. Denton's Memorandum goes on to identify 4
the recommended usage of these terms.
This should serve to make these terms more definitive when used in future licensing; however, our understanding of the usage intended in the Contention 10 a
language is that the pressurizer heaters and controls should be 3
classified as " safety-related" (as defined in the Denton Memorandum) and should,.therefore, be subject to the general requirements of the General Design Criteria (GCD) and 5 hat applicability of various GDC's should be judged by the guidance of 10 C.F.R.
4/
Part 100, Appendix A.
-3/
Memorandum from H. R. Denton to All NRR Personnel, November 20, 1981,
Subject:
" Standard Definitions for Commonly-Used Safety Classification Terms," Attachment B hereto.
4/
Our review of the Diablo Canyon pressurizer heater documenta-tion affirms our view that precise safety classification terminology is necessary and significant.
The NRC Staff believes the pressurizer heaters are considered " components important to safety" with respect to their pressure contCol function.
NRC Staff Motion for Summary Disposition of Contentions 10 and 12, p. 6.
The Applicant believes these components are not required to be classified as "important to safety."
Pacific Gas and Electric Company's Motion for Summary Dispositi-on, December 21, 1981, p. 4.
The Staff's position would lead to the belief that at least-some of the General Design Criteria have been applied, whereas the Applicant's position would indicate that none of the GDC apply (other than those that admittedly apply to the RCPB pressure retaining capability and to the breakers which can be used to connect the heaters to the onsite emergency power system).
Applicant's response to NUREG-0578 states that (Cont'd on next page)'
.r III.B.:
Importance of Pressurizer Heaters 8.
The pressurizer heater system used at the Diablo Canyon plant provides an important function, namely, the ability to control primary coolant pressure under various conditions.
Not only is the system used during normal power operation, but is especially needed for control of pressure and of natural circula-tion capability in the hot standby mode.
The NRC Staff's recom-mendations emanating from the TMI reviews recognize that maintenance of safe plant conditions depends on maintenance of pressure control in the primary system for the associated main-tenance of natural circulation capability.
The Staf f, therefore, recommended upgrading the pressurizer heaters and associated 4/
(Cont ' d) equipment identified as non-safety-grade will not be quali-fied for the Hosgri event, implying that the heaters, therefore, are not seismically qualified.
Pacific Gas and Electric Company Response to NUREG-0578, April 21, 1980,
- p. III-B-5.
The Westinghouse specification under which the pressurizer heaters were procured seems to confirm that only the coolant boundary GDC's were applied.
Furnished with Applicant Pacific Gas and Electric Company's Supple-mental Response to Joint Intervenors' Second Set o f Inter-rogatories, December 23, 1981, Immersion Heater Spec. 3E3A701.
The specification provides no design requirement on the radiation exposure the unit must withstand (the specific concern is the insulating boot at.the electrical connection),
nor does it address seismic loadings.
No information is given on heater sheath supports along the length of the heater (the heater rods are approximately eight feet long and are 7/8" in diameter).
These omissions provide little assurance that these important aspects have been adequately considered so as to produce a reliable source of pressure control.
(
5/
The NRC Staff's controls to achieve greater reliability.
Motion for Summary Disposition states that pressurizer heaters the hot standby are recuired to maintain system pressure at PG&E claims that heaters are not required for 6/
7/
condition.
We hot standby pressure control and natural circulation.
for this agree with the Staff that the heaters should be used The basis of this position is that this is the normal function.
and that that the procedures specify this mode, control mode, it is difficult for the operators to follow a different and infrequently used procedure under stressful conditions.
PG&E's intended reliance on the pressurizer heaters i
9.
is indicated by frequent mention of them in the Diablo Canyon No less than nine such procedures Emergency Operating Prc edures.
8/.
PG&E claims
~
call for the use of the pressurizer heater system.
NUREG-0578, NRR Lessons Learned Task Force Short-Term 5/
Recommendations, page A-2.
NRC Staff Motion for Summary Disposition of Contentions' 10 6/
and 12, page 5.
Affidavit of John B. Hoch, page 1, a part of Pacific Gas and Electric Company's Motion for Summary Disposition, I
7/
December 21, 1981.
Applicant Pacific Gas and Electric Company's Answer
[
8/
Governor Edmund G.
page 47.
l that alternate means (to the heater system) for pressure control are available; however, none of the cited emergency operating procedures specifically direct the operator how to proceed with alternatives if the heater system becomes unavailable.
(See Paragraph 11 for further discussion of procedural inadequacies.)
10.
The NRC Staff states that primary system pressure control is not a prerequisite for natural circulation as the Westinghouse design will provide natural circulation as long as adequate water is provided to the secondary side of the steam generators, even if the primary coolant pressure decays to bring the system to a saturated condition.
Applicant and NRC Staff also cite test data obtained at the Sequoyah Nuclbar Piant that supports the claim that the Diablo Canyon primary system pressure will decay at about 100 psig per hour if the pressurizer heaters are lost.
It has not yet been demonstrated, however, + hat these character-istics are true at Diablo Canyon.
Further, the Applicant has provided no directions in the Emergency Operating Procedures as to how the characteristics would be u.tilized to assure proper operation.
If it is the Applicant's intent to rely upon these claimed reactor characteristics, they should be demonstrated and necessary operator action (s) should be fully described in the
~
procedures.
III.C.:
Deficiencies of Present Pressurizer Heater System 11.
The purpose of the pressurizer heater system upgrading
? 1 1
l l
required by NUREG-0578 (and 0737) is to assure that primary coolant pressure control will be available when needed.
The time when this need is the greatest is during or following transient and/or accident conditions.
Emergency Operating Procedure OP-13, Malfunction of Reactor Pressure Control System, i
is intended to provide guidance on now to maintain primary pressure control when the pressure control devices malfunction.
This procedure only assumes control channel failure or failure to deenergize and therefore provides corrective action by placing the system in manual control.
No guidance is given as to how to proceed to " feed and bleed" or the other " alternate control methods" claimed by the Applicant.
Sin 41arly, EP OP-23, Natural Circulation of Reactor Coolant, has as' a basic assumption that offsite power and the heaters are available, making it incomplete for certain accident sequences.
12.
PG&E appears to be in a paradoxical situation.
On the one hand, PG&E has argued that the pressurizer heaters are not required for natural circulation; rather, other methods are available to ensure that this important cooling mechanism occurs.
However, in the Diablo Canyon Emergency Procedures (OP-13 and 23),
no other methods are provided for the operators' use.
- Thus, in our opinion, at a minimum, either the heaters should be up-t graded to safety grade or the other methods which presumably rely on safety grade systems should be specified.
Since the
O f -
other methods are not specified in the procedures at this time, there can be no assurance that Diablo Canyon opegators would, in fact, utilize such other systems if the non-safety-grade heaters l
were unavailable.
Thus, the procedures are inadequate or the 9/
~
heaters ', classification is inadequate, or both.
l 13.
Another deficiency affecting pressurizer heater re-liability during emergency conditions is the method required to transfer some of the heaters to the onsite emergency power system.
The NRC Staff claims that the dispatching of an operator to a remote (the 100 foot level of the Auxiliary Building) location to perform electrical breaker manipulations is an " acceptable alternative" to actually meeting clarifIca' tion item 4 of TMI requirement II.E.3-1, which specifies that transfer is to be 10/
[
accomplished in the control room.
The Staff does not state how I
this conclusion was reached, whether or not area radiation monitors will be available to assure immediate access to these areas j
l under accident conditions, or whether they have independently l
verified the operator radiation exposure of 10 mrem claimed by 9/
Use of the pressurizer heaters is clearly the preferable method of maintaining natural' circulation.
Thus, even assuming other methods may exist and assuming they may sub-sequently be identified in -the procedures, we believe the
~
heaters should be upgraded to safety grade to ensure to the extent feasible that this most useful equipment is available.
l 10/
NUREG-0675, Supplement 14, Safety E/aluation Report,
- p. 2-21.
s.
It also is not clear that the Staff has adequately Applicant.
/
assessed the potential delays and disruption to area accessibility inherent in a confusing post-accident situation.
Impact of Upgrading the Safety Clari-III.D.:
fication The possibility of upgrading all of the pressurizer 14.
heater system components to a " safety-related" classification recommended has been considered in the past and was, in fact, by one of the major NRC groups assembled to review the TMI Ihe recommendations presented included:
accident.
The pressurizer heater systgp should be classified as safety grade which wou12-assure emergency power availability and protection from failures due to environ,-
mental conditions. 11/
if followed, would have required full ad-This recommendation, herence to all applicable safety requirements and qualification of the components to appropriate saismic and environmental con-There are no reasons to believe that such upgrading ditions.
could not be done (from a " state-of-th<s-art" standpoint).
If safety classification vpgrading were to be required, 15.
Plant the pressurizer heater system should become more reliable.
safety would be improved by the minimization of the number of from R. D. Martin,.NRC, i
Menorandum for J. M. Allan, NRC, 1979, p.
11/
" Operations Team Recommendations," October 10,
~~
23 (emphasis added).
l l
challenges to the system and by the optimization of the operabil-ity 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 hhe preferred components for this capability.
It is our opinion that such upgrading would impose more of the safety design criteria on this system and its operability.
GDC 20 requires, for example, that the protection system shall be designed "to initiate the opera-tion of systems important to safety."
Standard Review Plan Table 7-1 extends the applicability of GDC 20 to all instrumenta-12/
tion and control functions important to~ safety.
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 i.
the Auxiliary Building and verbal confirmation that such action has been taken.--13/This procedure does not meet the automatic initiation requirements of GDC 20.
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 12/
NUREG 75/087, Section 7, Table 7-1.
--13/
See Philip A. Crane to Frank J. Miraglia, January 26, 1981,
- p. II.3-14.
accordance with GDC 2 (seismic and environmental qualification),
GDC 22 (protection system independence, " separation"), or GDC 3 (51:e protection).
Since these components have not been classi-fied as important to, safety, the requirements of GDC 1 (Quality standards and records) does not appear to have b'een applied.
IV.
CONCLUSION i
16.
The discussion in Part III above indicates a number of reasons why the pressurizer heater system components should be classified as safety-related components.
It also indicates some 'o'f ETbeMfits to be obtained by such classification.
e We therefore conclude that this action should be taken at the Diablo Canycn plant.
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ATTACHMENT A EXPERIENCEAllDhUALIFICATIONSOFDALEG. BRIDENBAUGH Dale G. Bridenbaugh is a~ Professional Nuclear Engineer licensed by the State of California (license NU 973), president of MHB Tech-nical Associates, and a member of the American Nuclear Society.
Bridenbaugh received a B.S. in Mechanical Engine-ring from the South Dakota School of Mines and Technology in 1953.
He has been intimate-ly involved with.the commercial nuclear power program since 1958, when he was first assigned in a supervisory capacity for the General Electric Company in the construction of the Dresden Nuclear Power Station near Chicago, Illinois.
Subsequent to that assignment, he has accumulated over,20 years of nuclear experience, including re-sponsible management of_ positions in construction, startup, op.eration, maintenance, and prod'ct improvement planning with General Electric's u
nuclear program.
Included in his background experience is the man-agement of the design, constraction, and checkout of the fi.rst me-bile test facility assembled by the General Electric Company for the on-site testing, under simulated environmental operating conditions, of various nuclear system safety and re? ief valves.
s'Nss S
Bridenbaugh has been involved with Pacific Gas and Electric's (PG6E) nuclear plant programs since 1966, when he was assigned re-sponsibility for liason with utilities on all operating nuclear plant matters.
This included the ongoing engineering effort by General Electric in support of PG6E's Humboldt Bay No. 3 nuclear unit.
After the formation of MHB Technical Associates in 1976, he has participated in the review and licensing process of the Dia-blo Canyon plant, presenting testimony before the ASLB in 1976 on expected plant c'apacity factors.
Bridenbaugh has analyzed the operations of numerous nuclear plaats in his previous and present positions.
He evaluated the re-sponse of the Sacramente Municipal Utility District Rancho Seco Plant to equipment and operating procedure recommendations made as a result of the TMI accident.
Results of this evaluation were pre-sented in direct testimony on behalf of the California Energy Com-4 mission in a hearing before the ASLB on Rancho Seco in 1980.
He has testified on similar matters before the ASLB on the Black Fox (Oklahoma) case.
He has also served as a consultant to the NRC on the review of the NRC safety improvement program and on the safety goals assessment program.
Bridenbaugh has testified on nuclear safety, reliability, and economic matters before the NRC (Commission and ASLB), before the Committee on Atomic Energy of the United States Congress, and Joint befo,re the energy and utility commissions of Ohio, New York, New Jer-sey, Massachusetts, and Californit.
He has also served as consultant to private and governmental bodies in Pennsylvania, Massachussetts, New York, Illinois, Texas, Oklahoma, and Oregon, as well as in Sweden, !
and Australia.
Additional information on the professional l
Italy,ication of Dale G. Bridenbaugh is set forth in the following:
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4 P RO FE S SION AL O UALI FI C ATI ON S OF DALE G.
E RIDEF? AUGH DALE G. BRIDENSAUGH 1723 Hamilton Avenue Suite K-Sen Jose, CA 95125 (408) 266-2716 E XPE RIEN CE :
19 7 6 - P RISENT President - MHB Technical Associates, San Jose. California.
Co-founder and parcner of technical consulting firm.
Specialists in energy consulting to governmental and other groups interested in evaluation of nuclear plant safety and licensing.
Consultant in this capacity to state agencies in California, New York, Illi-no is, N ew Jers ey,,
P enns ylvania, Oklahoma and Minnesota and to the Norwegian. Nuclear Power Committee, Swedish Nuclear Inspectorate, and various other organizations and environmental groups.
Per-f ormed extensive safety analysis for Swedish Energy Commission and contributed to the Union of Con'c e rn ej S cien tis t's Review of W AS E-14 00.
Consultant to the U.S. NRC - LWE S af ety Improvement Program, performed Cost Analysis of Spent Fuel Df.sposal,for the Natural Resources Defense Council, and contributed to the Depart-ment of Energy LWR Safety Improvement Pro gram f or S anc'.ia Labora-tories.
Served as expert witness in NRC and state utility commis sion hearings.
i 1976 - ( FEB RUARY - AUGUS T)
Consultant, Project Survival, Palo Alto, California.
l Volun teer work on Nuclear S af eguards Initiative campaigns An California, 0,regon, W a s h in g t on, Arizona, and Colorado.
Numerous presentations on nuclear power and alternative energy options to civic, government, and college groups.
Also resource person for public service presentations on radio and television.
1973 - 1976 Manager, P erf orman ce Evaluation and Improvement, General Electric Company - Nuclear Energy Division, San Jose, California.
Managed seventeen technical and s even clerical personnel with responsibility for establishment and management of systems to monitor and measure B oiling Water Reactor equipment and system operational performance.
Integrated General Electric resources in customer plant modifications, c o o rd in a ted correction of causes of forced outages and of efforts to improve reliability and par-f ormance of BWR sys tems.
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1373 - 1976 (Coned) of Division Mas ter P erf ormance Responsible f or development Plan as well as'for numerous Staff special assign-Improvement ments on long-range studies.
Was on special assignment for the management of two different ad hoc proj ects formed to resolve unique technical problems.
1972 - 1973 Manager, Product Service, General Electric Company - Nuclear Energy Division, S an J o s e, California.
of twenty-ona technical and four clerical personnel.
Managed group interface and liaison personnel Prime responsibility was to direct involved in corrective actions required under contract warranties.
Also in charge of refueling and service p lannin g, performance analysis, and service communication functions supporting all com-supplied by General pleted commercial nuclear power reactors Electric, both domestic and overseas (S p ain, Germany, Italy, Japan, India, and Switzerland).
1968 - 1972 l
Manager, Product Service, General Electric Company - Nuclear Energy i
Division, S an_ J ose, Calif ornia. -
Managed sixteen technical and six clerical personnel with the responsibility f or all customer contact, planning and execution of work required after the customer acceptance of department-supplied plants and/or equipment.
This included quotation, sale and delivery of spare and renewal parts.
S ales volume of parts in cre as ed from $1,000,000 1.n 1968 to over $3,000,000 in 1972.
1966 - 1968 Man a g er, Complaint and Harranty S ervice, General Electric Company Nuclear Energy Division, S an Jose, California.
[
six persons with the responsibility for customer Managed group of planning and execution of work required after customer of department-supplied plants and/or equipment--both
- contacts, acceptance domestic and overseas.
1963 - 1966 Field En gineerin e Supervisor,. General Electric Company, Installatiot i
and Service Engineering Department, Los Angeles, California.
Supervised approximately eight field representatives with responsi-I bility for General Electric steam and gas turbine installation and l
maintenance work in S outhern < Calif ornia, Arizona, and Southern Nevada.
During this period was re s pons ible for tbs ins'tallation of central s tation s team turbine gene rator units, plus l
eight different much maintenance activity.
Work included cus tomer contact, prepa-I ration of quotations, and contract negotiations.
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1956 - 1963 Field En g ine er, General Electric Company, Ins tallation and Service Engineering Dep a r t= en t, Chicago. Illinois.
Supervised installation and maintenance of steam turbines of all s
than'o'ne hundred men, s ite s.
Supervised crews of from ten to more depending on :he. job.
k'orked prinarily with large utilities but had significant work with steel, petroleum and other process industries.
Had four. years of experience at construction, startup, trouble-shooting and refueling of the first large-scale commercial nuclear power unit.
1955 - 1956 Engineerina Trainine Procram, General Electric Company, E rie,_
P enn s y lv an i a, and S chenectady, New York.
Training assignments in plant facilities design and in steam two General Electric Factory locations.
turbine testing at 1953 - 1955 United S tates Army - Ordnance S chool, Aberdeen, Maryland.
Instructor - Heavy Artillery Repair.
Trught classroom and shop l
disassembly of artillery pieces.
1953 En g in e e rin g Training-Program, General Electric Company, Evandale,_
1 Ohio.
j f
Training assignment with Aircraft Gas Turbine Department.
1 i
EDUCATION & AFFILIATIONS:
1953, South Dakota S chool of Mines and Technology, BSHI j
Rapid City, South Dakota, Upper k of class.
j Certificate No. 0973.
P rof es sional Nuclear Engineer - Calif ornia.
j Member - American Nuclear Society.
career including Pro'f es-Various Company Training Courses during sional Busines s Management, Kepner Tregoe Decision Making, Effective Presentation, and numerous technical seminars.
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HONORS & AWARDS:
Sigma Tau - Honorary Engineering Fraternity.
General Managers Award, General Electric Company.
?
l PERSONAL DATA:
B orn November 20, 1931. Miller, South Dakota.
Married, three children 6'2",
190 lbs., health - excellent H on o r able discharge from United States Army Hobbies:
S kiiin g, h ikin g, work with Cub and Boy Scout Groups.
P UB LIC ATION S & TES TIMONY :
Operatine and Maintenance Experience, presented at Twelfth 1.
' Annual Seminar f or Electric Utility Executives, Pebble Beach, California, October 1972, published in General Electric NEDC-10697, December 1972.
and In-Service Inspe tion, presested at IAEA Maintenance 2.
Symposium on Experience From Operating and Fueling of Nuclear P ower Plants, B ridenbau.gh, Lloyd & Turner, Vienna, Austria, October, 1973.
Op era tin g and Maintenance Experience, presented at Thirteenth 3.
P ebble Beach, Annual Seminar f or Electric Utility Executives,
Calif ornir, November, 1973, published in General Electric NEDO-20222, January. 1974.
4.
Imoroving P lan t Availability, presented at Thirteenth Annual P e'ob le B e a ch, Cali-S eminar f or Electric Utility Executives,
November 197 3, published in General Electric NEDO- : fornia, 20222, January, 1974 5.
Auplication of Plant Outage Experience to Improve Plant Per-Bridenbaugh and Burdsall, American Power Conference,
- formance, Chicago, Illinois, April 14, 1974.
6.
Nuclear Valve Testine Cuts Cost, Time, Electrical World, October, 15, 1974 7.
The Risks of Nuclear P ower Reactors :
A Review of the NRC Reactor S af ety S tudy W ASH-14 00, Kendall, Hubbard, Minor &
3ridenbaugh, et al, f or the Union of Concerned S cientists,
August, 1977.
A-5 L
8.
Swedish Ranctor Safety Study:
B ars ebh*ck Risk As sas smmu t,
l HEB Technical Associates, January, 1978.
(Published by the Swedish Department of Industry as Document Ds1 1978:1) 9.
Testimony of D.G. Bridenbaugh, R.B.
- Hubbard, G.C. Minor to I
the Calif ornia S tate As sembly Committee on Resources, Land Use, and Energy, March 8, 1976.
10.
Testimony of D.G. Bridenbaugh, R.B. Hubbard, and G.C.
Minor before the United S tates Congress, Joint Committee on Atomic Energy, February 18, 19 7 6, Washing ton, DC '(P ublished by the Union of Concerned S cientis ts, Cambridge, Massachusetts.)
11.
Tas timony by D.G. Bridenbaugh bef ore the California Energy Commission, entitled, Initiation of Catastrophic Accidents at Diable Canyon, Etarings on Imergency Planning, Avila beach, California, November 4, 1976.
12.
Testimony by D.G. Bridenbaugh before the U.S. Nuclear Regula-tory Commission, subject: Diablo Canyon Nuclear Plant Perfor-j mance, Atomic S af ety and Licensing B oard Hearings, December, 1976.
13.
Testimony by D.G. Bridenbaugh before the California Energy Commission, subject: Interim Spent Fuel S torag e Considerations, March 10, 1977.
14.
Testimony by D.G. Bridenbaugh bef ore the New York State Public S ervice Commissio. Siting B oard Hearing s concerning the James-port Nuclear Power S tation, subj ec t : Effect of Technical and Safety Deffeiencies on Nucler.r Plant Cost and Reliability, April, 1977.
15.
.Tes timony by D. G. B ridenbaugh bef ore the Calif ornia S tate Energy Cor71ssion, subject:
Decommissioninc of Pres surized Water Reactors, S unde s ert Nu cle.ar Plan t Hearings, June 9, 1977.
16.
Tes tinony by D.G. B ridenbaugh bef ore ~ the Calif ornia S tate Energy Commission, subject: E conomic Relationships of Decommissioning, Sundesert Nuclear Plant, for the Natural
- Resources Def ense Council, July 15,.1977.
17.
Testimony by D.G. Bridenbaugh bef ore the Vermont State Board 1
of Health, subj ect : Ooeration of Vermonr Tankee Nuclear Plant and Its Impact on Public Health and Safety, October 6, 1977.
18.
Tes timony by D.G. B ridenbaugh bef ore the U.S.
Nuclear Regula-tory Commission, Atomic Safety and Li cens ing B oard, subj ect :
Deficiencies in S a f e ty Evaluation of Non-Seismic Is sues, Lack of a Definitive Finding of Safety, Diablo Canyon Nuclear Units October 18, 197 7, Avila Beach, California.
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Commission Bridenbough before ths NorvGgidn 19.
Testimony by D.G.
Re actor S a f e rv /Ris k. _ October 26, on Nuclear P ower, -subj ect:
1977.
20.
Testimony by D.G. Bridenbaugh bef ore the Louisiana S tata on Natural Resources, subject: Nuclear Legislature Committee P ower Plant Deficiencies Imp a c tin e on Safety & Reliability, Baton Rouge, Louisiana, February 13, 1978.
f prepared by D.G. Bridenbaugh report 21.
Soent _F.uel Disonsal Costs, Def ense Council (NRDC), August 31, for the Natural Resources 1978.
and R.B. Hubbard Tes timony by D. G. B ridenbaugh, G.C.
Minor,
in the matter 22.
the,Ato=ic Saf ety and Licensing B oard, of the Black Fox Nuclear P ower S tation Construction Permit bef ore Hearings, September 25, 1978, Tulsa, Oklahoma.
N of D.G. B ridenbaugh and R.B. Hubbard bef ore the'N Nuclear Plant and P ower
's, 23.
Testimony Louisiana Pul.lic S ervice Commis sion,
'N L ouis ian a.
.s s
Generation Cos ts, N ovember 19, 197 8, B aton Rouge, the City Council and Testimony by D.C. Bridenbaugh bef ore Texas, Design, Con-24 Electric Utility Commission of Austin, struction, and Operating Exoerience of Nuclear Generating Facilities, December 5, 1978, Austin, Texas.
Testimony by D.G. Bridenbaugh f or the Commonwealth of 25.
Massachusetts, Department of Public Utilities, Imeact o f_
Unresolved Safety Issuec, Generic Deficiencies, and Three on P ower Generation Cost, Mile I sland-Initiated Modification s June 8,
1979.
the P ropos ed Pilgrim-2 Nuclear Plant,
at 26.
Inurovine the Safety of LWR P ower Plants, KHB Technical Associates, prepared for U.S. Dept. of Energy, S andia Laboratories, September 28, 1979.
for 27.
BVR Pine and N o zzle Cracks _, MEB Technical Associates, (SKI), October, 1979.
the Swedish Nuclear Power Inspectorate 28.
Testimony of D.G. B ridenbaugh an d G. C. Minor before the in the matter of Atomic S af e ty and Lice'n s in g B oard,
Rancho Seco Nuclear S acramento Municipal Utility District, Generating S tation f ollowin g THI-2 accident, subject:
Ooera tor Training _ and Human Factors En gfineering, for the Calif ornia Energy Commis sion, February 11, 1980.
29.
Italian Reactor Safety S tudy:
Caorso Risk As s es smen t, MHB for Friends of the Earth, Italy,
' Technical Ascociates,
March, 1980.
30.
De c on t amin a tion of Krypton-85 from Three Mile Island Nuclear P l an t,,
H. Kendall, R.
Pollard, & D.G. B r id enb augh, et al, the Governor of Concerned S cientis ts, delivered to
~he Union of Penn sylvania, May 15, 1980.
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Testimony by D.C.
th e N ew Jorscy Board of Public Utilities',' B ridenb augh b ef oreJersey ?ublic Advocate's 31.
on behalf of New Office, Civisien of Rate Counsel,
'An a ly s is of 1979 Salem-1 Refueling Outare, August, 1980.
32.
Position S tatement, Proposed Rulemaking on th3 S torage and Disposal c f Nuclear Was ta, Joint Cros s-S tatement of Position
.of the New En gland Coalition on Nuclear Polluti'on and the Natural Res ources Defense Council, September, 1980.
33.
Tes timony by D.G. Bridenbaugh and Cregory C. Minor, before the New York S tat'a Public Service Commission, In the Matter of Long Island Lighting Company Temporary Rate Case, prepared for the Shoreham opponents Coalition, September 22,.1980, Shoreham Nuclear Plant ' Cons truction S chedule.
34.
Supplemental Testimony by D.C. Bridenbaugh before the New Jersey Boad of Public Utilities, on behalf of New Jersey Public Advocate's Office, Division of Rate Counsel, An aly s is of 1979 Salem-1 Refueling Outage, December, 1980.
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. 3 ca g 'e, UNITED STATES
., f/
- n NUCLEAR REGULATORY COMMISSION
,(.
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,I WASMNGTON. D. C. 20555
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NOV 2 01931 i;?:
MEMORANDUM FOR:
All NRR Personnel l
FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation-l
SUBJECT:
STANDARD DEFINITIONS FOR COMMONLY-USED SAFETY CLASSIFICATION TERMS' Litigation of one of the principal issues in the TMI-1 Restart Hearing brought to light the fact that 'there is not complete consistency among all elements of the NRR staff jn the application of safety classification terms used frequently in the conduct'.c,f NRR's safety review and licensing activities.
More specifi-cally, it appears that terms "important to safety," " safety grade," and " safety-related" have been used at times interchangeably, or in w~ays not completely consistent with the definitions and usage of such terms in the regulations, and which d'o not fully reflect the intent of the regulations or current licensing practice.
Efforts have been underway for some months now to develop guidance for the consistent usage of these terms.
These efforts ha.we included:
(a) review of a large number of Reg Guides and SRP's, in conjunction with parts of the regula-tions upon which they are based, for consistency in the application of safety classification terminology, (2) extensive discussions cmong cognizant NRR, RES (Stds. Devel.) and ELD representatives regarding proper interpretation and application of such terms, including consideration of alternative " standard" definitions and (3) consultation with the cognizant ACRS Subcommittee regarding these matters, and consideration by the fall ACRS as well.
As a. result of these efforts, I am endorsing and prescribing for use by all NRR personnel the standard definitions set forth in the enclosure to this letter.
It should be noted that in connection with long-tenn efforts to develop means for ranking reactor p' ant systems with respect to degree' of importance to safety, and in connection with related efforts to develop a graded Q.A. approach in reactor licensing, the general question of safety classifications and safety classification terminologies will be reexamined; and this could result in changes to the defini-tions set forth in the enclosure or perhaps in development of a completely new scheme in this regard.
For the time being, however, the definitions in the en-closure should be considered " standard" and should be applied consistently by all NTP. personnel in all aspects of our safety review and licensing activities and should be appropriately reflected in our regulatory guidance documents.
B-1 D U')
D#iT8 C-
All "NRR' Personnel.
It is expected that minor editorial revisions will have to be made to some existing Reg Guides and SRP's in order to make their worcing consistent with these definitions',f You should review the regulatory guidance documents within your purview in this regard and recommend the necessary changes; it is not expected that this will involve extensive revision efforts. 'I'want to make clear that my interest here is only in establishing consistency in the language -
used by all cognizant groups within NRR in expressing our technical requirements.
It is not my intention by this action to dictate new technical requirements, to modify existing technical requirements, or to broaden the existing scope of NRR licensing review.
Harold R. Denton, Director Office of Nuclear Reactor Regulation Enclosure :
Definition of Terms O
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DEFINITION OF TEPJ45'
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Imoortant to Sa'f6 y Definition - From 10 CFR 50, Appendix A (General Design O'riteria) - see first a
paragraph of " Introduction."
"Those structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public*."
- =T Encompasses the bfcad class of-plant features, covered (not necessarily e
exolicitly) in the General Design Criteria, that contribute in important way
- t. safe opera tion and protection of the ptblic in all phases and. aspects of facility operation-(i.e., normal operation and transient control as well as accident mitigation).
Includes Safety-Grade (or Safety-Related) as a subset.
a Safety-ela ted Definition - From 10 CFR 100, Appendix A - see sections III.(c), VI.a'(1), and e
VI.b.(3).
Those structure, systems, or components' designed to remain functional for the SSE (also termed ' safety features') necessary to assure recuired safety functions, i.e.:
(1) the integrity of the reactor coolant pressure boundary; (2)' the capability to. shut down the. reactor and maintain it in a safe
. shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures compar.able to the guideline exposures of this part.
Subset of "Important to Safety" e
Regulatory Guide 1.29 orovidesan LWR-oeneric, function-oriented listing of e
" safety-related" structures, systems, and components neecea to provide.;r perform' required safety functions.
Additional information (e.g., NSSS type, BOP design A-E, etc.) is needed to generate the complete listing of safety-related SSC's for any soetific f:cility.
~
l Note:
The term " safety-related" also appears in 10 CFR 50, Appendix B (Q.A. Program Requirements); nowever,. in that context it is framed in somewhat different language than its definition in 10 CFR 100, Arpendix A.
That. difference in language between the two appendices has contributed to confusion and misunderstanding regarding the exact meaning of " safety-related"land its relationship to "important to safety" and " safety-grade."
A revision to the language of Appendix B has been proposed'to clarify this situation and remove any ambiquity in the meanina of these terms.
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Sa fegy-Gra de e Term not used expliciIly in regulations but widely used/ applied by staff and industry in safety review process.
s Equivalent 9b "Sa'fety-P. elated," i.e., both terms apply to the same subset of the broad class "Important to Safety."
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yEuTED CCCW"?T'M EXHIBIT 4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFOPI THE ATOMIC SAFETY AND LICENSING BOARD
.. )
In the Matter of
)
)
PACIFIC GAS AND ELECTRIC COMPANY
)
Docket Nos. 50-275 0.L.
)
50-323 0.L.
(Diablo Canyon Nuclear Power
)
Plant, Units 1 and 2)
)
)
PREPARED DIRECT TESTIMONY OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR ON BEHALF OF GOVERNOR EDMUND G. BROWN JR.
REGARDING N
s>
CONTENTION 12 lyj
-D le JAtJ 151982 > Z 4
If' tu::e e e., ser.3,y D::h::q Qg f (b,
ECt' n J
January 11, 1982 Ng jpta3\\s Y
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PREPARED DIRECT TESTIMONY OF DALE G. B_RIDENBAUGH AND GREGORY C. MINOR REGARDING CONTENTION 12 I.
INTRODUCTION l.
1.
My name ib Dale G. Bridenbaugh.
A stateme'nt of my 1
qualifications and experience has previously been provided to r
i this Board as part of my testimony on Contention 10 and in Attachment A to that testimony.
2.
My name is Gregory C. tiinor.
A statement of my qualifications and experience has previously been provided to this Board as part of my testimony on Contention 1 and in Attachment B to that testimony.
II.
STATEMENT OF CONTENTION i
3.
The purpose of our testimony is to respond to 1/
~
Contention 12 as admitted by the Board as follows:
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 addition, their failure can cause or aggravate a LOCA.
Therefore, 1/
ASLB Memorandum and Order, September 30, 1981.
On September 21, the Commission directed the Licensing Board to includ.; in the full power proceeding Joint Intervenors' low power Contention 12.
l
these valves must be classified as com-ponents important to safety and required to meet all safety-grade design
~,
criteria.
Further, the Appeal Board's order of December 11, 1981, expands Contention 12 to include "the testing and vsrification of these same components" since " testing and verification of these com-2/
ponents is an integral part of the qualification process."
including the Thus, the adequacy of the qualification process, is included in the expanded adequacy of the EPRI testing program, scope of Contention 12.
The results of our review of some of the important matters encompassed by this Contention are summarized in the following paragraphs.
III.
DISCUSSION OF ISSUES The NRC's Criteria for Equipment Classification III.A.:
are Confused There is confusion as to the meaning of terms used to 4.
describe the safety significance of structures, systems, and 3/
The NRC issued a memorandum components in nuclear power plants.
2/
ASIAB Order, December 11, 1981, p. 3.
3/
Memorandum from H. R. Denton t.o All NRC Personnel, Novem-i ber 20, 1981,
Subject:
" Standard Definitions for Commonly-used Safety Classification Terms."
which provided definitions'of the most often used safety classi-fication terms as follows (see Attachment A for the full text) :
Important to Safety:_
Those structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
Safety-Related:
systems, or components designed Those structures, to remain functional for the SSE (also termed
' safety features') necessary to assure required safety functions, i.e.,:
(1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the guideline exposures or tnis part.
Safety-Grade:
Term not used explicity in regulations but widely used/ applied by staff and industry in safety review process.
Equivalent to " Safety-Related,"
i.e.,
both terms apply to the same subset of the broad class "Important to Safety."
The writing of Contention 12 preceded the issuance of the clarification document.
If Contention 12 had been written using the definitions of the Denton memo, the term
" safety-related" would have been used instead of "important to it is From our review of the Applicant's submissions, safety."
4.-
unclear to us how the Applicant is using the safety classification terms and how it defines important to safety. "..
The Safety Significance of the PORVs' III.B.:
and Block Valves' Functions Justifies Safety-Related Classification The <'esign of Diablo Canyon includes 3 PORV's and 3 5.
's Two of the relief valves are described associated block valves.
having tant to safety" and, the third is not, by PG&E as "im been added to provide capability for 100% lesad rejection without 4/ The three block valves are also described as reactor ~ trip.
"important to safety."
The PORV's and/or Block Valves perform several functions 6.
which have safety significance along the lines of one or more of These functions are:
the definitions,in paragrapg 4.
Maintain integrity of the prinary pressure a.
boundary.
Provide pressure relief for Low Temperature b.
Overpressurization conditions.
Reduce the number of challenges to the c.
safety vilves.
Reduce the number of challenges to the BCCS.
d.
l Provide a bleed capability during the feed-and-(
bleed mode of operation to remove decay heat e.
from the core.
5/
(PG&E's response PG&E response to Interrogatory No. 46.
includes the term " local rejection" which is interpreted 4/
as a typographical error for " load rejection".)
As used in the TMI-2 accident and as referred to in NUR 5/
0578, Sec. 2.2.1 and page A-1.
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Each of these functions is consistent with the definitions of "important-to-safety," and the first two functions are also consistent with the definitions of " safety-related."
The FSAR is vague as to the safety classification of 7.
the POHV's, Block valves, and their circuits and controls.
The Applicant has stated that the qualification level of the three PORV's and their circuits are not all identical.
- However, documents which the operator relies on for guidance in operating the plant during emergency conditions (Emergency Operating Pro-cedures) and deciding on an acceptable plant configuration (Diablo Canyon Technical Specifications) provide no evidence of differentiation between the greater =or less " qualified valves or associated equipment."
The Block valves and/or PORV's are called upon to be 8.
operated or checked for misoperation in several of the Emergency Operating Procedures.
For example, EOP-20 calls for checking the PORV's as a possible source of excessive. leakage from the coolant system (i.e., a small LOCA).
EOP-38 (ATWT) ' describes the need for automatic opening of the PORV's and checking later to see that they are not stuck open in the event of a pressure decay and EOP-2 describes the actions to prevent challenges coolant loss.
to the pressurizer safety valves in the case of loss of secondary i
It too mentions that the transient may cause the PORV's coolant.
to open and' requires that their resetting be checked, thus insuring against a small LOCA in the primary. coolant.
1 P
y The emergency operating procedure for Emergenc 9
describes conditions where the use of a Shutdown (OP-22) rize the primamf PORV/BV combination may be needed to depressu d for boration.
loop so the safety injection pumps may be use sed for throttling The PORV would be opened and the block valve u to any The procedure does not restrict the operator f ty-grade alternative the flow.
particular PORV nor does it identify a sa e Thus any of the PORV/BV component to accomplish the task.
i h this safety-related combinations should be able to accompl s tor-task.
Ercrgency Procedure OP-13 on Malfunction of Reac ORV/BV combination 10..
Pressure Control System calls for use of a P The same procedure as a back-up pressure contro1 technique.
ORV's which identifies techniques for finding stuck open P l pressure control may be leaking coolant and exacerbating norma 6/
l Specifica methods.
Section 3.4.9.3 of the Diablo Canyon Technica 11.
be operable during Hot Shutdown tions requires that two PORV's There is no conditions for overpressure protection.
the more qualified (Mode 4) guidance, however, to the operator to choose l
i ted as an It also notes that a stuck open PORV is des gna f
ffsite
" UNUSUAL EVENT" and requires notification of o(EOP G l
6/
personnel per the emergency. proceduresdix Event of
~
Emergency).
an
0
- l PORV's.'
Thus, all PORV's should be qualified to the same level the operators ' EOP 's should restrict which-two valves are to
,cn:
be used.
12.
During operating modes 1, 2 and 3 (Power Operation, Startup, and Hot Standby), Tech Spec Section 3.4.4 re,quireb that each PORV must be operable or isolated by an operable block valve which is then deenergized.
For these cases either the PORV's or their associated block valves are relied upon to protect the integrity of the primary pressure boundary.
However, accord-ing to the Technical Specification, it is possible to block the two higher-qualified valves and rely only on the lesser-qualified valve and its associated controls.
ITe feei there should be
~ ~
instructions to the operator to prevent this situati' n if the o
difference in valve classification continues to exist.
III.C.:
The Fact that Diablo Canyon Has More PORV's than Other Plants is Not Necessarily an Advantace 13.
Since it is permissible to operate with one or more of the PORV's isolated by their block valves, and there are no restrictions on which valves are isolated, it is possible that the PORV with the lesser classified components would be the only valve operable at a time when PORV operation was called upon by a transient or accident.
The fact that the Diablo Canyon design has more valves 14.
than most plants is commendable but it is not always a virtue.
The addition of the third valve may help the reactor ride through thus preventing a challenge to the a load rejection transient, protection system, but it alsc creates additiona'l failure points additional common mode failure which conid result in a small LOCA, mechcnisms, and the possibi2 2 ty of systems interaction which 4
could impact other safety-related functions.
Oualification of PORV's and BV's i
III.D.:
is Incomplete Proper safety classification-of the PORV/BV and their 15.
controls and instruments should insure proper design and quali-fication for worst case conditions and plant-specific evalua-7/
~
tions.
the qualification of the Diablo Canyon 16.
- However, The Bv's have not been PORV's and Block Vibes is incomplete.
and there apparently are no plans for further fully. tested, 8/ The full range of conditions, including ATWS, has not testing.
PGEE Memorandum relating to EPRI safety 7/
See June 16, 1981raising a potential issue regarding Diablo valve testing,
~
Such testing designed to ensure Canyon safety valves.
qualification of valves will increase reliability of and confidence in Diablo Canyon systems.
Second Set of See NRC Staff Response to Joint Intervenors' 8/
Interrogatories, No. 61(e).
~
e
~9-been tested and the plant-spec!fic analysis has not been prepared to cover Diablo Canyon's design of PORV/BV's and their components, systems, and structures.
Thus there can be no assurance that the configuration meets.GDC 2 and 14.
Also, tha scheduled completion of the valve tests and the plant-specific analyses have been 9/
delayed until July 1, 1982.
This may not be soon enough to satisfy the terms of the Low Power Testing License, Section 1,
- p. 6.
A s.
IV.
CONCLUSIONS 17.
Based on the functions and required operations of the PORV's and Block valves, as described above, and according to the NRC definitions of safety terms, the PORV's/BV's and their instruments, Oi m zuls and structurcs, should all be classed as 10/
~~
" safety-related."
18.
There are insufficient test data and plant-specific analyses to show that the Diablo Canyon PdRV/BV's and associated equipment and ctructures have been properly qualified.
i 9/
Ibid, No. 61 (d).
10/
" Safety-grade" is also appropriate since it is defined as equivalent to " safety-related" by the NRC.
. j,s H c o
UNITED STATES
- '[
,,' h NUCLEAR REGULATORY COMMISSION VvASMlNGToN D. C. 20$55
,.g p
%, A 4
/
NOV 2 01931 J:
HEMORANDUM FOR:
All NRR Personnel FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation-
SUBJECT:
STANDARD DEFINITIONS FOR COMMONLY-USED SAFETY CLASSIFICATION TERMS' Litigation of one of the principal issues in the TMI-l' Restart Hearing brought to light the fact that 'there is not complete consistency among all elements of the NRR staff in the application of safety classification terms used frequently in the conduct of NRR's safety review and licensing activities. More specifi-cally, it appears that terms "important to safety," " safety grade," and " safety-related" have been used at times interchangeably, or in ways not completely consi:: tent with the definitions and usage of such terms in the regulations, and which d'o not fully reflect the intent of the regulations or current licensing practice.
Efforts have been underway for some months now to develop guidance for the consistent usage of these terms.
These efforts have included:
(a) review of a large number of Reg Guides and SRP's, in conjunction with parts of the regula-tions upon which they are based, for consistency in the application of safety classification terminology, (2) extensive discussions among cognizant NRR, RES (Stds. Devel.) and ELD representatives regarding proper interpretation and application of such terms, including consideration of alternative " standard" definitions and (3) consultation with the cognizant ACRS Subcomittee regarding these matters, and consideration by the full ACRS as well.
As a. result of these efforts, I am endorsing and prescribing for use by all' NRR personnel the standard definitions set forth in the enclosure to this letter.
It should be noted that in connection with long-term efforts to develop means for ranking reactor plant systems with respect to degree of importance to safety, and in connection with related efforts to develop a graded Q.A. approach in reactor licensing, the general question of safety classifications and safety classification terminologies will be reexamined; and this could result in changes to the defini-l tions set forth in the enclosure or perhaps in development of a completely new l
scheme in this regard.
For the time being, however, the definitions in the en-closure should be considered " standard" and should be applied consistently by all NRR personnel in all aspects of our safety review and licensing activities and should be appropriately reflected in our regulatory guidance documents.
i 1
A-1 1
9e
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- =====w-
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wa===**
( Ml' NRR Personnel
-2
~
(
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~
It is expected that minor editorial revisions will have to be made to some existing Reg Guides and SRP's in order to make their wording consistent with these definitions',T You should review the regulatory guidance documents within your purview in this regard and recommend the ne:essary changes; it is not expected that this will involve extensive revisicn efforts.
I want to make clear tha.t my interest here is only in establishing consistency in the language used by all cognizant groups within NRR in expressing our technical requirements.
It is not my intention by this action to dictate new technical requirements, to modify existing technical requirements, or to broaden the existing scope of NRR licensing review.
42e h Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
Definition of Terms 9
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DEFINITION OF TEpf45' Imoortant to Sa'fe# y Definition - From 10 CFR 50, Appendix A (General Design Criteria) - see first e
paragraph of " Introduction."
"Those structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the, health and safety of the public"."
Encompasses the broad class of plant features, covered (not necessarily exolicitly) in the General Design Criteria, that contribute in important way e
te safe operation and protection of the pLblic in all phases and aspects of facility operation-(i.e., normal operation and transient control as well as accident mitigation).
Includes Safety-Grade (or Safety-Related) as a subset.
e Safe y-Rela ted Definition - From 10 CFR 100, Appendix A - see sections III.(c), VI.a.(1), and P
e
~
VI.b.(3).
Those structure, systems, or components desigFed to remain functional for i
d ft the SSE (also termed ' safety features') necessary to assure recu re sa e y functions, i.e. :
the integrity of the reactor coolant pressure boundary; (1) the capability to shut down the reactor and maintain it in a safe (2) shutdown condition; or the capability to prevent or mitigate the consequences of accidents which could result in potential off-site exposures comparable to the (3) guideline exposures of this part.
Subset of "Importent to Safety" e
Regulatory Guide 1.29 providesan LWR-oeneric, function-oriented listing o
" safety-related" structurcs, systems, ano conponents neeceo to provid e
perform ^ required safety functions.
B0p design A-E, etc.) is needed to generate the complete listing of safety-related SSC's for any soecific facility.
The term " safety-related" also appears in 10 CFR 50, Appendix B Noh.e:
(Q.A. Program Requirements); however, in that context it is framed in somewhat different language than its definition in 10 CFR 100, That difference in language between the two appendices has contributed to confusion and misunderstanding regarding the exact Appendix A.
meaning of " safety-related" and its relationship to "important toA re safety" and " safety-grade."
B has been proposed to clarify this situation and remove any ambiguity in the meaning of these terms.
g--
- A T_-
o Sa fety-Gra de
[
e Term not used explicitly in regulations but widely used/ applied by staff
~,
and industry in safety review process.
o Equivalent *.b " Safety-Related," i.e., both terms apply to the same subset of the broad class "Important to Safety."
e 3
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EXHIBIT 5 l
1 E
g.nr> C9r.nESFOSUDC 41 UNITED STATES OF AMERICA
> *~ l NUCLEAR REGULATORY COMMISSION f.
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 1
)
In the Matter of
)
)
Docket Nos. 50-275 0.L.
PACIFIC GAS AND ELECTRIC COMPANY
)
50-323 0.L.
)
(Diablo Canyon Nuclear Power
)
(Low Power Test Proceeding:
Plant, Unit Nos. 1 and 2)
)
)
II OPPOSITION OF GOVERNOR EDMUND G. BROWN JR.
TO THE NRC STAFF AND
),
PACIFIC GAS AND ELECTRIC COMPANY MOTIONS F.O_R RECONSIDERATION AND
SUMMARY
DISPOSITION m
~
N
\\
Byron S. Georgiou I
t' Legal Affairs Secretary
'l J
l Il Governor's Office
(
2
-4 State Capitol JAN 151gS2 >
Sacramento, California 95814
{ l.
0l!$!3.l th S::re,,7y u sua 4
n L "h Herbert H. Brown 2
6 Lawrence Coe'Langher 4
HILL, CHRISTOPHER AND PHILLIPS, P.C.
c 1900 M Street, N.W.
L Washington, D.C.
20036 F
Attorneys for Governor Edmund G.
d f
April 24, 1981 Brown, Jr. of the State of Californ:
L J
- 1 I
u
i IV.
The NRC Staff and PGLE Moticns for Summarv Discosition Must Be Denied.
The Staf f and PG&E motions request summary disposition of 1
j emergency planning,1/ water all of Governor Brown's Subjects:
level indicators,8/ and relief and block valves.1/
M The Governor
+
demonstrates below that the Staff and PG&E motions must be denied.
W ell-s ettled law and NBC administrative practice require the PG&E motion be denied summarily.
Section 2.749(a) of that G
annex to its sum-the NPC's regulations requires that the movant short and concise statement mary disposition motion "a separate, o f the material facts as to which the moving party contends that there is no genuine issue to be heard."
This is a mandatorv pro-vision of law.
See Houston Lichtine and Power Co. (Allens Creek C
Nuclear Generating Station), ALAB-629, CCH Nuc. Reg. Rpt.
1 30,562 (1981).
PG&E completelv icnores this mandatorv recuire-ment.
PG&E cannot ccmplain abcut"this Board's summary denial of PG&E's motion.
In the Stanislaus case, where PG&E similarly was the applicant and the movant for summary disposition, PG&E was summary disposi-chided for f ailing to comply with the very same tAon requirement.
In Stanislaus, the Board stated:
1/
Governor Brcwn Subject 3 and Joint Intervenors' Contentions 4 and 5.
8/
Governor Brcwn Subject 13 and Joint Intervenors' Contention i.;
13.
-9/
Governor Brcwn Subject 14 and Joint Intervenors' Contention
- 24..
l
[
Subsection (a) [o f Saction (.,N9] clearly requires I
that "There' shall be annexed to the motion a sepa-rate short and concise statement of the material facts as to which the moving party contends that there is no genuine issue to be heard."
PG&E has f
failed to file this recuired statement of material facts.
Such a recui~ement is not merelv a croce-
+
dural technicalltv, but it is of substantive significance.
This statement is necessarv in order to m. pose upon other carties a duty to file I
a statement of material facts as to which it is contended there exists a genuine issue to be heard under penalty of having uncontroverted material facts deemed to be admitted.
It is necersarv for the Board to have this information in a readily a vailable form in order to evaluate the merits of a motion for summary disposition.
PG&E's lengthy
]
(77 pages olus numerous exhibits) and arcumenta-tive motion for summarv disocsition wholly falls to ccmolv with the recuirement of a concise state-ment of material facts as to which there is no In re Pacific Gas and Electric Co.
cenuine issue.
~
CCH Nuc. Reg. Rp t.
(Stanislaus Nuclear Project),
't
? 30,211 (LSP 1977) (emphasis supplied).
4 PG&E has again The same situation exists in this case.
lengthy and argumentative motion with numerous exhibits filed a that, as the Board stated in Stanislaus, "sholly fails to comply with the requirement of a concise statement of material facts as to which there is no genuine issue."
PG&E, like other partici-pants, must adhere to the NRC's regulations.
For this blatant g),*?
Violation of reculatory reouirements, the Governor submits that f
PG&E's motion should be summarily denied.
M[
The Staff has also failed to follow strictly the NRC's sum-mary disposition regulations, although the Staff's violation is sweeping as that of PG&E.
Section 2.749(b) requires that not so
"[a]ffidavits shall set forth such facts as would be admissible in evidence...."
The NRC Staff has violated this requirement.
For example, paragraphs 8-10, 12-13, 16-17, and 23 of the St =.f f's as to which the Staff alleges there is emergency planning f acts,