ML20214R489
| ML20214R489 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/12/1984 |
| From: | Brady R Office of Nuclear Reactor Regulation |
| To: | Bosnak R, Knighton G, Lear G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20213D572 | List: |
| References | |
| FOIA-84-741, FOIA-84-742 NUDOCS 8706080167 | |
| Download: ML20214R489 (43) | |
Text
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION E
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- WASHlNGTON, D. C. 20555
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' \\ ',,,, 8 January 12, 1984 NOTE: This also went to RV.
G. Knighton, Chief, LB#3
< " 1 % v aas-R. Bosnak, Chief, MES F. Rosa, Chief, ICSB G. Lear, Chief, SGEB
- 0. Parr, Chief, ASB MEMORANDUM FOR:
V. Noonan, Chief EQB
,A. Thadani, Chief, RRAB FROM:
RichaidC.."Brady Management Analysis Branch Planning and Program Analysis Staff, NRR
SUBJECT:
REQUEST FROM BILLIE GARDE FOR DOCUMENTS
/
FOIA-84-21
(
REGARDING THE SAFETY EVALUATION REPORT RELATED TO THE OPERATION OF THE DIABLO CANYON NUCLEAR POWER PLANT y
The subject FOIA request is enclosed for your actionhN Please provide a copy of occuments you have that are subject to this request.
Please list the documents that'ycu are providing, along with a list of other dccuments which you know exist that are subject tc this request.
Incoming documents which are handled through the official distribution system are in the PDR and need not be provided.
However, please identify specific letters, reports, dates, applicable docket numbers, and accession numbers so the requester will be able to find the documents.
Internally generated documents do not routinely go in the PDR unless PDR is specified on the official file copy. Therefore, all relevant memos which have been generated without the FDR designation should be provided, even if it m' ans retrievino e
them from Central Files.
Do not indicate PDR on tne file copy of your response to me.
If you believe the search time will exceed two hours, please contact me immediately.
Also let me know if you think other NRR bran:hes might have documents which are subject to this request.
4 R n Management Analysis Branch Planning and Prografn Analysis Staff, NRR
Enclosure:
FOIA Requertl
.\\, - f 8706080167 070600
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TASK INTERFACE AGREEMENT DATE: NovEm;;
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'AC #: (t(7. gg O
-PROBLEM:
Diablo Canyon - RHR Allegations LEAD OFFICE: /[/I&E
/[/NRR
/_xf REGION
/ [/ JOINT
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NOTIFICATION:
REFERENCES:
Memo to DEisenhut from TBishop dated 10/26/83, subject: Request for NRR
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Assistance in Evaluating RHR System Allegations (including allegers letter of 10/20/83).
AC.* ION PLAN:
NRR:
Evaluate,silegers paragrcphs (c)(4), (e), (fl, (g), and (.1) as requested by Region V Tasks are specified on attached Evaluation Schedule. Consider a,1propriate paragraphs of Region V Inspection Report 50-275/82-42.
Region V: Evaluate remaining allegations.
Pro /ide NRC response to alleger.
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NRR:
Designate Lead Project Manager to assign TACS and coordinate correspondence, meetings, and reports (ORB #
/ LB# 3 Hans Schierling).
Prepare response to Region V for OL signature.
OFFICE COORDINATORS:
G. Holahan' (X27415)
R. Mattson d'
(X 27373)
([u (X 27425)
APPROVED:
T. Novak (X
)
T hsb(I&E),d(%
Mn (X
)
(Reg on J
F. Mira (X27492 )
c:
V. Stello, ROGR J. Heltemes, AE00 H. Thompson, NRR' R. Purple, NRR 8egional Adms.
-T. Speis. NRR R. Wessman, NRR -
+
J. Taylor, ISE D. Eisenhut, NRR G. Knighton, LB#3
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R. Baer, I&E G. Lainas, NRR F. Rosa E. Jordan, I&E R. Vollmer, NRR,.
B. Sheron.
G. Lanik, !&E H. Denton, NRR Ti Novak, NRR 3> \\.
E. Case, flRR F. Miraglia, NRR R. DeYoung, ISE R
G. Holahan Ma- %. Mattson, flRR 5 m u s, NRR ne a s a w w w
l Evaluation Schedule Diablo Canyon RHR Allegations
/
/
1.
Conduct pMiminary evaluation of RHR allegations to determine if they impact Comission decision on lifting Diablo Canyon license suspension.
Provide written evaluation results to PM. Evaluate items and review branches are as follows:
ICSB lead (a)
Evaluate RHR control circuit desi n aspects relating to RSB ' assist RHR Suction Valves (8701 and 8702 and consider effect of removin para C.4) g power from SSPS output relays(Alleger EF" W'.
(b) Evaluate FSAR statements regarding RHR System over-
]-
ICSB assist
- pressure /overtemperature protection (A11eger para., e) c ICSB., lead..
-(c)
. Evaluate potential for spurious RHR suction
~~RSB assist.,
... _._ ' closure, and effects thereof (Alleger para.. valve f) m-t l
ICSB lead (d). Evaluate acceptability of RHR pump trip annunciation..
~
~ and sugtion valve position indication (Alleger para. g)
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.(e) Evaluate configuration of RHR hot leg suction for compliance -
, ' s..with regulatory requirements. Evaluate-acceptability of
_not classifyin hot leg suction as "not safety-related".... :5."
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- 2. ~ If findings from preliminary " evaluation indicate more extensive evaluation is appropriate, conduct f'nal evaluation of allegations 1.a - 1.e,' above, and provide written results to PM. Review branches an as in item 1, above.
NOTES:
1.
Complete Preliminary evaluation by C.O.B. 10/24/83 to support Comission Meeting on 10/28/83 Complete}ina1'evaluationbh2/01/83) 2.
3.
Contacts win alleger or licensee must be coordinated with Region V.
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EXHIBIT 10b - NRC INS?ECTION REPCRT $0--275/S2-42.... 56-67
[N EXHIBIT 10e - NRC INSPECTION REPORTS 50-275/E2-25 1
j AND 50-323/32-13......................GE-7a
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EXHIBIT 11 LETTER John H.
Cooper to T.W.
Bisnop in rebuttal to NRC Incpection Reports 10 -2 0-63.......... 7 9-6 2 r.
EXHISIT. PORTICNS OF UNCFFICIAL " SHOP LCG".........
83 EXHILIT 13 - CVD INSTRUMENT SHOP PETITICN FOR MOPE INFORMATION..................
84-85 EXHILIT 14 MEMORANDUM John H.
Cooper to J. M.
Giscion describing procedure er'rors 4-3-79..................
86 EXHISIT 15 - REPLY to abd.ve memo, Mark Stephens to J.
Giscion 4-9-79...................a7-89
'5 - MEMORANDUM John H.
Cooper to EXHIS*T Juanito Diamonon noting ese' essivelly long times to correct errors in plant crawings 11-13-81...............
90 EX-IBIT 17a 50 E Y ANALYSIS REPORT, SUPOLEMENT 21
.f (appropriate portions)...............91-119 p.,
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EXHIBIT IS - TRANSCRIPT and comment s, 12-21-S3
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interview witn Jesse Crews and Al Johnson of NRC.................... 129-171 e
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AcFIDCVIT
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My name is John H.
Cooper. I am submitting this affidavit
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h freely and voluntarily, without any threats, inducements or coercion, to Mr. Thomas Devine, wno has identified himself to me
/
as the. legal director of the Government Accountability project i
(GAD) of the Institute for policy Studies.
Thi s st at ement
' evidences my concern that the Nuclear Regulatory Commission (NRC) i may permit operation of the Diablo Canyon nuclear power plant, despite$$ts vulnerability -to spurious closures of the Residual l
Heat. Removal (RHR) hot-leg suction valves (8701 and S703),
i causing loss of decay heat removal capability and possibly camaging the Residual Heat Removal Pumps, which are part of the Emergency Core Cooling System. I am also concerned about a management breakdown in the program to report and correct deficiencies both within the Pacific Gas and Electrie. (PG3E) organization and within the Nuclear Regulatory Commission.
I believe that the lack of response I have received from PG4E to my.
attempts to correct vari,ous problems in the Diablo Canyon RHR system is a symptom of a more widespread sickness"within the PG&E organization.
Finally, I am concerned about PG4E retaliation against.those who challenge these abuses.
~
~
From April 1931 to March 13S2, I woQked at Diablo Canyon.as a Ficid Engineer in the General Construction instrumentatzon deper ment.
crom Maren 1976 to November 1979, I workee at Diablo Canyon as,an Inspec or, an instrumentation Group Leacer, a
Control Technician, and a Temporary Instrument Maintenance
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Foreman. I have a Bachelor of Science degree in Electronic
/
Engineering form California State polytechnic University, and
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have also worked as a Startup Engineer at other nuclear and coal-j fired electric generating stations.
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D2:!CIENCIES IN THE DIABLO CANYON RESIDUAL HEAT REMOVAC $YSTEM A.
INTRODUCTION
~
I have been seeking resolution of the problems in the Diablo Canyon Residual Heat Removal system for nearly three years.
During this time, these problems were repeatedly brought to the attention of the pG3E management', who took no effective action to solve them.
I have also approached the NRC several times on'th'e same problems, again with no results. Although in conversations with the NRC, a n'd in the NRCi s own publications, the NRC staff 1
has conceded the cecuracy of my position, apparently nothing is going to be done to correct the deficiencies in the design of the RHR system at Diablo Canyon prior to operation of the plant.'
In May of 1961, I discovered a quirk in the control cir-cuitry for RHR pumps suction valves S701 and 8702 wheremy they l
would spuriously close wnen the power was turned off in a totally l unrelated portion of the power plant, the Solid-State protection
' 5cg b System (SSPS). At a simple analogy, consider the surprised person
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~g who purchased a new home.only to find that whenever th,e _1,ight s were turned off in the bathroom, the carbage disposal 'in the w
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e kitchen would come on!
The new owner would certainly not put up t
with this, but would simply call the Contractor who built the house and have the problem fixed. In the case of the Diablo
/
j Canyon RHR pumps suction valves, however, PG&E's response to this l
1
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proelerg was to " post a notice on the bathroom walla Gauli221
'f k!D21ME ibi 21"$A22 1E1122111 322f2CI AM""!D2 Sff 1h2 1211Cr222!
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.ligtig". pG4E's act ual words were: " Problem exists as stated.
3d.
Instead of re-wiring the circuit as suggested, the operating proceduje, step 18 is being revised to require the operator to open the. valve ' operator breaker when the valve has opened, thus preventing it from closing. " This is how the situation stands as i
of today.
I Of course, a nuclear plant is much more complicated than a r gg6 house, and the potential exists for danger to the public if its equipment coesn' t operate as designed. The two RHR pumps at
' ggs Diablo Canyon are each 400 Horsepower pumps, operating on 4160 VAC, and are part of the Emergency Core Cooling System. The RHR TCO',
system is a high-volume, low-pressure cooling water system, required in each nuclear plant by federal regulations. It is used
%"I for rem'oving " residual heat" from the reacter coolant system both" during a normal cooldown and during.the "long-term recirculation" pnase after an accident.
~ i, NOTE: Residual. Neat is the heat remaining in the nuclear fuel and reactor coolant system after the control rocs have been inserted into the reactor core and tne nuclear react ion is shut down. This heat comes from two s'ources. The primary b
source is the nuclear decay of atoms that have,been split in the nuclear chain reaction, as they.try to reacn a more g
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stable state. This heat amounts to a significant portion of tne full power of,a reactor,,and exponentially decays over a t
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persc4 of months. *he other source is the sensible:.b. eat g'c.
contained in the large amount of water and metal in'the reactor coolant. system. (As when ont removes a casserole from the oven - It still remains hot for quite some time) Resicual' heat must be continuously removed after shutdown, or the i
reactor core will begin to heat up again. At otner nuclear plants, this system may be known as the " Decay-Heat Removal system" or- "Shutcown Cooling System".
The problems with tne Diablo Canyon RHR system center en 5 e two gate valves in the 14". stainless steel line between the reactor coolant system hot leg #4 and the suction side of tne R.qR pumps.
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i RN NOTE: The RCS " hot legs" are the pipes carrying water heated
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i by the nuclear core twsy from the reactor, to the steam I
i generators, while the " cold legs" are the lines returning the water back to the reactor, after passing througn the steam generators and reactor coolant pumps.
i If either of these valves goes closed inadvertantly, all (r,.
decay heat removal througi t$is path is lost. Also( i f the-RHR W4
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pumps are operated for more tnan about 15 minutes with tne.
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,y suction valves closed, severe damage may occur to one or both pumps due to overheating and cavitation (which actually occurred
)
at Diablo Canyon in October of 19S3. In this instance, one RHR 1
pumo had t o be replaced, after it was run for one hour with a suction valve closed.)
B.
REGULATCRY REQUIREMENTS FOR RESIDUAL HEAT REMOVAL SYSTEMS The* minimum design requirements for a nuclear power plant built in -the United States are outlined by the NRC in Title 10 of the Code of Federal Regulations, Part 50, the General Design
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Criteria (GDC). According to these regulations, each reactor must
- f
have a system to remo<e residual heat. The regulation governing
,3 :n" -
the cesign of a Residual Heat Removal System, GDC #34, states:
A system to remove residual heat shall be provided. The system Eff21% fSDSil2D shall be to transfer fission product decay heat and other, residual heat from the reactor core at a rate such that specified acceptable fuel desig*n limits and the design conditions of the reactor coolant pressure boundary are not exceeded.
Suita 1e"reduncancy in component s a'nd features, and suitable interconnections, leak detection,,and isolation capac211 ties i
shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available)
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and for offsite electric power system operation (assuming y'pg onsite power is not availacle) 103 3v3133 33f31y furSAA2D E*D
<<d i f-h~h ki tE22221A5bt2r ti22WiDg i siDgig !sklyDg. (emphasis added) s I
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10 C== 50.46 " Acceptance criteria for emergency core cooling systems for light water nuclear power reactors" also states in paragraph (b) (5) :
pp.c '
(5) L2DE:12CT E2211Dg. After any ea
- eulated successful.
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initial operation of the ECCS, the calculated core temperature sna11 be maintained at an acceptable low 'value and cecay heat shall be removed for the extendec. period of time required by the long-lived racicactivity remaining in the core.
Also, in ReguIatory Guide 1.133 " Guidance for Residual Heat Removal", the NRC states",
... experience shows that there have been events t5at recuired eventual cooldown and long-term cooling until the RCS was cold enough-to perform inspection and repairs.. It is therefore obvious that the ability to transfer heat from the
'9 reactor to the environment af ter a snutcown 15 AD iT22DIADI
' P" #"I.
2Affi% f2DSEA2p for both pWRs'and BWRs. Consequently, it is essential tnat a powe-plant have the capability to go from i
hot-standby to cold-shutdown conditions (when this is
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determinec to be the. safest course of action) undeh~a'np j
accident concitions. (empnasis adced) 1 4
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19 can bo seen from tho abovo Yoforences that the',NRC feels
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that the RHR system in a nuclear plant performs an import ant hg;,'
g safety funct ion and must-still function after the failure of any V ~^
!g g g 3 3 single piece of equipment, and uncer any accicent conditions. It
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',js is important to note this NRC definition of the gaftly function
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J of an RNR system, since it differs so greatly from the AG4E
'p,y cefinition. PG&E dces not even consider a major portion of the Diablo Canyon RHR system as important to safety!
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C.
DIgCUSSION OF DIASLC CANYON RHR SYSTEM
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The RHR system at Diablo Canyon, although one system, can be made to perform several functions by opening and closing various valves.
This system may be " lined up" to take water from one of three different sources, cool the water, and then pump it back into the reactor (or other places) via a number of different paths. The suction of the RHR pumps may be connected to 1)the j
Reactor Coolant System (RCS) " hot leg" #4, 2) the containment recir,- 4 culation sumps (which remove water spilled on the floor of the i
containment after an accident), and 3) the Refueling Water j
Storage Tanx. The PG4E cescription of this system is contained in the Final Safety Analysis Report (FSAR) for the plant, section
]
- 5. 5. 7, woien states:
"The Residual Heat Removal System transfers neat frem the I
Reactor Coolant. System to the Component Cooling Water System i
to recuee the temperature of the reactor coolant to tne cole I
snutcown temperature at a controlled rate during the latter
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part of normal plant coolcown and maintains this temperature until the plant is started up again. CPG &E consicers tnas a f
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DSO 1AISly function]
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As a seconcary fu6ction, the Residual Heat Removal 3,gs, tem also serves as part of the Emergency Core Cooling System curing the injection and recirculation phases of a. loss of coolant accident." EpG4E consioers tnis tne safety function of the system]
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The Residual Heat Removal System also is used to transfer refueling water between the refueling water storage tank anc
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t,e refueling cavity before anc after the refueling operations.
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NOTE: information in bracket s added by affiant 4
AG&E defines the 2*f tf% fung 11hn of tne Diablo Canyon R. ;
y system as only the recirculation and cooling of water spillec on
- ne containment floor after a large-crean less-of-coolant aces-
'"y, cent (LOCA). By'tnis definition, the remaincer-of the Resacual
' eat Removal System (tne portion normally usec to remove restdual neat from the core) performs no safety function and hence is not b
repuired to tolerate a single failure.
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,p During normal heatup and cooldown (removal of cecay anc g ga
-ecteuel heat), the RHR pudps~take suction from RCS*not Ie;"4 gt
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t through a ningle 14" pipa contcining two valves in Eeries, valves 6701 and 8703.
The RHR pumps then pump this water through two
{~'N heat exchangers to cool it down, and then beek to the RCS cold legs. Two valves are proviced in series in the suction line from i
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hot leg'4 to insure that when the RHR system is not being used, q,oc <
the high pressure in the RC3 (3235 psig during power operation)
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can' t possibly get through to the RHR system, even'if one valve
, leaks through Unfortunately, if githgr of these valves fails to g ;c
.Lw open when neeced for cooldown, then the RHR system can' t be used
'*i h A; to remojb residual heat from the reactor.
La sch The Diablo Canyon RHR system, then, is redundant 20h for a 1arge-creak loss of coolant accident, and not,for normal residual heat removal or any other kind of accident, such as a small-break LOCA, where large amounts of water are D21 spilled onto the L(J ~
containment floor. If an accident were to occur at Diablo Canyon such that the reactor core were damaged but only small amounts of
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water were released into the containment, then the reactor core would have to be cooled via the " normal" suction line, whien does 3 gg not meet the single fai1*ure criterion. Of course, before Three SME' Mile Island, this type of accident was not considered " credible" by the nuclear industry-or the NRC itself.
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- a. -[9 pG&E, probably made aware of the ce'ficiencies in the design gmc,
of tnis system by the new famous accinent at.Three Mile Islanc,
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nas dene some fancy-footworn in orcer to Justify tne lack of 7,7),
recuncancy in the RH; system. Tnis nas =ent erec mostly on the argumer.: :nat altnougn tne RWR hot leg suction line coes not meet
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f' 'g 55E
=~.e sir.gle failure criterion requirec by 10 C.n 50 GCC#34, the
=rebar:1:ty of a single fa21ure in valve S701 or 5702 (or tne 1"**,
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pip;ng it sel f) is so low as to be "increcible". AGAE's arguments N
are as.follows:
- 1. "A failure, suen as the spurious closure,of a d5foF operatee valve tnat as normally open anc remains open upon
,e i
loss of electrical power or the spurious opening of a motor
! a,,p.:
operated valve that is normally closed anc remains closed 7d. e upon loss of electrical power, cannot be cefinec as an active 4..
or passive failure in the no-mal sense of tne cefinitions'
- r. ch I
use: above.
Because of cesign features and criteria applie-to prevent tnis type of failure, its occurrence curing tne i
period when these valves are required to be in a specifiee-.*
pos: tion in orcer to perform a safety function has not meen cons: cered ereciale." (F3AR section 3-1, p3)
- 3. "The random shorting of the manual closing switch contacts or a hot short in the cable run to the closing coil have been identified as control system failure moces whien coule lead to spurious movemen,t of a passive motor operatec valve.
The estimated prosabilaties of tnese of tnese faults, cased on MIL-HDBM-217A with credit for periodic tests are one in 40 i
milli.on per valve-hour anc one in 2.5,billson per valve-nour respect ive
- y.-
The p-esasility of any spurious valve closure
/
is therefore one in 3h 37_0,073 per valve-hour.
In tne case
?y-of 3mergency Core Coe'.:ng System valves, the maximum m'ission I'
i i
- E* ~
I i
J' \\
i
--~.,n-
~~. ~ '.
" - '. ' ' ~. ~. ' ' '. --
. ~.
-n---~~
- ~ ~ ~ -
-~-
~.
C fa n
tidh occurs for thoco.volves wh'ich cro required for the switchover from cold leg recirculation to hot leg recirculation; i. e., approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For this i
maximum mission time, the probability of a spurious motor-operated valve movement is computed as one in 1,633,344 per
\\ -
val.ve-mission at any given time.
7he the procamility of the
~
failure of any one of these valves coincicent.with an event requiring the actuation of the ECCS is considerec sufficiently low so as to be considered incredible." (FSAR seeffon G.3, p34a)
- 3. " Westinghouse does not consider spurious operation of i
electrically centro 11ed valves as a credible single active f.ailure." (FSAR section.15.4, pS)
\\
Unfortunately, the operating experience of otner nuclear p
plant s-in the U. S.
shows that the spurious closure of an Rhn g
suction valve is not only " credible", but happens on a regular basis'. It is difficult to believe that neither PG4E nor Westing-house (the Nuclear Steam
- Supply System vendor) is aware of this fact. On a cursory search of Lice.nsee Event Reports, I turned up sixteen reported cases of spurious RuR suction valve closures between I??3. and 1551, or over 5 per year. Th i s wor'a s out to a pro abil,ity of spurious RHR suction valve closure of more like 1 in 100 per valve-mission, vastly different than the S in ten m:111on enance calculated by PG3E. Even my
- 1. in 100 figure coesn't take into account the many incidents which went jg, unreported to the NRC (such as the ones at Trojan on 7-21-75, i"[y !
(
Diablo Canyon on 3-28-81 and 10-27-43, and at Palo Verde on 5 -
S 3. None of these incidents were caused by " random shorting of j
('~
sw' itch contacts" or other near-impossibilities that AG4E takes great delight in analyzing, but by three simple. factors:
.s
- 1. Door design of the control circuitry intended td'~ prevent l
the opening of these valves at the wrong times, especially in the Westinghouse reactors, and i
Human error caused by inadequate drawings of the shstem
~
~
2.
3.
Lack of control room annunciation to alert the operators of loss of RHR flow l
Westinghouse has, apparently realized the deficiency in this design, since their newer plants (designed after 1965) have 1*2 redundant RHR hot leg suction lines, so that a failure either one of them won' t cause a loss of decay heat removal capability. One must assume that Westinghouse would not spenc the collars td add
.l's another 14" stainless steel suction line to their RPR system d.5),
cesign without good reason.
Other nuclear plants, such as the Dalo Verde Nuclear Generating Station (NSSS by Comoustion g3.
Engineering) go to even greater lengths to increase tne
. g. "
reliability of this system, with two redundant suction lines and
,0*
t ree isolation valves in each line. Of course, in both the new g,p (*
(
Westinghouse and Cembustien E,ngineering designs, the RHR hot leg j
j suet zen is consiceree saf ty relatee.
aja
- ~ < - - -
t
.a
- gy fvt
,s s [,
s
, of -
t K.n
~
3
. - - ~. -.
M
-,y-
.-..7.._
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~ 3 -~~3 3 3 ~;;3 3 *3 ~, ~ } ~ ' _ **[__,
-~ _ _ _ _
3 (p
l L
Since FG15's argureents for the reliamility of this system do r
ne really nold up uncer analysis, as a last resort PG4E also
~5 g argues snat the RHR system is not really needed anyway, since gg f.
- )
- they can use the Auxiliary Feecwater System for long term gg cooling. In supplement 7 to the Safety Evaluation Repert, PGSE y,,2 claims that "...long term heat removal can be accomplished incefinitely with the steam generators and the auxiliary feed-d' [
=ater system rather that tne residual heat removal system." (SER
["
supplempht 7, p3-3)
Whetner this meets General Design Criterion i
34 is questionab*e, since residual heat must be reliably removed ig,5~
j if there is no offsite power available. With no offsite power,
- haJ, i
- ne Aeactor Coolant Dumps mignt be unavailable a'nd " natural circulation" of water into the reactor core would be the only.
. j,.5 y means of heat remova*.
'I NCTI: cGSI Startup Test #44.1, 44.' 2, and 44. 3 "Special Low i
towe-Test Program - Natura,1 Circulation" will be performed q' g y,'
by D332 during tneir low power testing program to cemenstrate ene acequacy of " natural circulation". Unfortbnately, these
! M $#.
tests demonstrate only.that the plant can be kept in the " he t - f "' "'[,
c:ancby" concition, not brought,to the " cold-shutcown" conca-L('
t:en as -recomt.1encet by the NRC in Regulatory Guice 1.*33.
, ((t, ;
.Jak
=
i n_ C sa..-,. m ef... C..,.
,N. nasa.....wn..w--.
---.m s.
s mm,.,m w eYS.
.3, c..
s
-a...
. w
.m
.~
~
f Cne woul: tnink tna: the NTC neule te interes ec in 23ese i
protieres in the Diat*o Canyon ib.3 system *for many reasons:
j f
1.
The N30 censicers the entire question of resicual heat i
\\
retcova*.
o ce a " Category A Unresolvec Safety Issue".
I e
j NO I4 Ca egery A is cefinec as " hese generie tec.y;ca*
j cet:vi-iss Jue;e-1y the staf f to warrant, priority
+
at t er.t i e n :r. terms of rean;ower anc/or funds to attain ear *y 1
i re sc *. ut ion.
hese matters incluce incse, the resolution cf w-icn cou*.c (*)
provice a significant :ncrease in assurance c f tne nea* tn and safe y c.? the puslic, or (2) have a s:gn,:fican- - im;ac: upon ene reae ce licensing process."
e NAC wac c-carec ey tMe Congress, througn section 2:0 ef t e Energy 3eorgar.izat:en Act of 1974, to identify anc Th cascri:e issues relating to the safety of nuclear power plaets w31c3 were unresolved at the time of enactmer..
- see NUR23-04;O, N'*SE3-0660, vol 1,
and NUMIG-C&c5, :ne "p;.a Book") In tne " Aqua Book", Task A-45, "S.uscown De:ay.ga:
Removal Mequirements", the NMC states that "The overa**.
purpose of Tasa A-45 is to evalua e 13e acequacy of current licens:ng cesign requirements, in crcer to er.sure t9at i
nuclear power plants ce net pose an unacceptacle ris. cue to failure te remove s atcown cecay hee:. " One ef t.e A-45 Suo-J Tasks is tne " Assessment of existir.g, plants to identify those in whien T.2." tecay -eat lez vall systems requ:re improvemer. ".
ne results of nis stucy wi'1 no
~be available i
,m untti late ;355.
~ '
i.
1
.....--_.,-__m..,
-....,m...
-.. _ _. ~ -. -. _ ~ -. -.,.... - -. _ _... _.
1
'C__
- e D
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L D
- 2. Since the accident at Three Mile Island, Unit 2, the focus of the NRC's nuclear safety program has s
enanged. Instead of analy=ing only large, catastrophie (but very unlikely) acci-l s
cents, tne NRC is now taking a realistic look
~,
at smaller, likely to occur, mishaps. The Electric Gower Research more Institute (EPRI) put this very well by stating. "One very important lesson of TMI-2 is that small errors or,malfune-tions in a nuclear power plant can compound into an accident whose consequences i
are potentially as serious as those caused by'a' single, large less-of-coolant accident, which in the past has been the main focus of licensing requirement s. "
j
- 3. NRO Regulatory Guice 1.139, " Guidance for Residual Heat 1.
Remova.*" stresses the importance ~to safety of the RHR system The importance of reliable systems that remove decay heat from the reactor coolant system (RC3) while the latter is at or near normal operating temperatures is indicated by
. tne result s of WA5H-1400, '" Reactor Sa fety Stucy" (RSS). The capasility of a typical PWR plant and a typical PWR p*. ant e remove decay heat following a plant trip was evaluated in the RS3 on a pro abilistic basis... However, in the ever.t of a plant trip even with a sucpessful operation o' tne R;5 1rea:,cr protection system 1,
- y;2225 20 2 21 22 1 f3;1u-es tna 1ee to ene in;;111;y 1; r;;;g; g;;3y ;g;.;
C2521222 AD t bi222C 2 22&211122 21 & 25C2 2212 2222 1221 2:2512122 ft: b 12:22 %999 22" botn ***s 225 EsB2
(
(emanasts ac:ec)
\\
s
- a. Regulatory Gu ce 1.135 also ' outlines several cesign "eatures wnsen snoulc de incorpo6ated into this system, none of w-ich are present at Diacio Canyon. The two meste. glaring cefie:enezes are:
. The Diablo Canyon hot leg suction line will not perform
' fe e its function assuming a single failure.
~
- 2. Tne Dia "o Canyon RFA system coes not meet the criterion
- }.'d t..at
"=ailure of a power sup *y should
{.4 not eaase any valve a ' q],
to change position."
5.
ne spurious closure of an RHR suction valve is a (rI-v "recure:rg commen cause" fault, wheresy a single "ailure in a
" ' ~
non-safety-related (according to PG4E) system can ca.ase tetn i
sa'ety-related RHR pumps to fail.
7 g'
- 5. This syst,em poses a tnreet to tne tne en-ire So*ic State Drotection System, wnereby a s i ng.' e e.* ect r i ca *. '"a u * : in a
" g' non-safety-related (according to PG3I) system woulc cisable an entire train of tne reactor protection system.
Un fort u na t'e l y, neitner the NRC nor PG32 have snown muen ar.terest in correcting t.ece serious ceficiew.e:es in Dianic-Cany on' s CHR system.
(
,s s
e
.w.
m.m.e-es,==-*'
"~
\\ L) c
'o E.
CHRONC'CGY OF MY ATTEMPTS'TO CCERECT THESE DE.ICIENC:25 1
The following is a.cnronology and a brief explanation of my I
attempt's to correct this problem over the last 2 1/2 years. As I
{
continued my researen, I discovered that many other nuclear-plants in the US hac experieneec reliability proslerss with this system.,Several RHR pumps had been camagec, and in,some cases, reactor ecoling was lost and the core began to heat up.
I also ciscovehkd that most other plants have redundant RHR suction jyQi d'
E j.
lines to. prevent suen eccurrances, but that for some reason uf4" 2
Diaolo Canyon coes not. :f for any reason, either valve 6701 oc g)ic.A 5703 coulc not be opened after a TMI type acci, dent, the RHR system ceule not be used for long term cooling, and the plant l
woule have to rely on " natural circulation" for ecoling the core, ',
a poor alternative, out the only one available. In Regulatory j
Guice 1.133, the NRC staff states that
" Natural circulation may lead to slow and unequal cooling.
Slow coeling will re'sult in longer cooldown t iines, which, in
\\
turn, recuire a larger clean feecnater inventory.
Unequal lW A eco*ing may lead to hot spots and high vessel stresses.
It ifA i
- s essenttal that acecuate cooling,be provided uncer these i d<l circumst'anees to kee; the intege:ty o.f tne reactor coolant pressure bouncary and maintain the,, reactor core in a coolable form."
l' f
Cn 5 4.y 19, 1951 wrote a memorandum to J.M. Giscion (Dept.
1
(
Heee. Ceast Valleys Div:stor CCVD, now known as Nuc* ear power f
0:eest ens: 'ee7nical and Engineering Departments) explain:r.g
(,
snat valves 5701 anc 8702 woule fail closed wnen SSDS out put f'cp
- user wet *e removee and that emergency procecure CD-8 " Control tror e ae:es s i':1.11ty','. was in error. I also sent copies of this meet t e N*.
Dat t erson (Dept. Heae, ~CVD Cperat ions Destat,'-Pa'ul G:lereats (General Construction CGC: Instrutnentation Supervisor),
are JT W1' son (GC Instrumentation Foremen). (The asove memorancum is enclosed as Exnimit 1)
~
On May 25, 1541 I was called to to my superviser's o'ffice e
anc te'.
never again to sene a memo cirectly to Mr.Giscion, but to use tne " chain of commene" in the future. He told me snat I l
was over-reactzng to a " minor procecure problem". I explained 4 ",'
c that :t was my responsim'ility both professionally and under NRC regulations to make tne responsible PG4E management aware of any nuclear safety promlems whien I discovered. I continued that if PG8E was not interested in correcting this preslem, it was aise my responsibility to contact the NRC.
.*y supervisor state: 15 at i'
- f I contacted anyone outside tne company : "mi g P. t lose my ;om".
t t
On July 30, 1981, since I had received no response from Mr.
Giscion, Mr. Patterson, or Mr. Gilbreath to my memo of 5-19-31, I ca*1ed the NRC Site Representative, Mr. M,arvin Mencorca anc explaine'd the proslem to him. Mr. Menco::a apparently ciscussed (f.
t.e situation witn.9r. Gisple.n or Mr. Ken Doss (CVD :nstrument Sueervisor) anc Nuclear p; ant proslem Report (Np3R)* w' 'DOD-51-T:-
(
i e
.e-i
^
...+.....
_- - _. _ - _. =.. -. - - -.. -,
T,_, ~ ~ -' " ~- ~ ~. _ ~ ~ * ' * * ' ~ ~ " * ~ ~., _ -
_. _. ~.... _
fV
?
n---
~
p0237 was initiatec to revise procedure Op-8.
Nothing was cone
{
to correct the wiring proclem.(The NG34 is enclosec as Exhicit 2)
On August 3, 1981, I was approached by Mr.
R. D.
Et:1er (GC i
- Site Superintencent) concerning my having contactec the NRC. he st at ed that although it was my right to contact the NRC,.he would appreciate it if in the future I would go through him first. I do not know how he was made aware that I had cont act ed the NRC - I 4
certainly didn' t tell him.
On August 6, 1931, M r. Thornberry (CVD plant Superintendent) j put a
" hold" on my security clearance, which m,eant I Mac to be j,
i escorted when inside Unit 1.
I callec Mr. Mencorea amout this, 4
and explained that I felt I was being punished for talking to him i
about the RHR system.
On August 7, 1981 my security clearance was reinstatec.
On September 22, 1981, Diablo Canyon received Facility Operating License eDpR-76 On or about Septemmer 23, 1931.I sent a copy of my hanc-crawn circuit diagrams for valves 8701 and S702 to Joe Rappa (CVD Instrument Foreman) with a note explaining tnat removing tne ssp 5 output fuses would-cause these valves 'to fail elesee.
l Cn September 23, 1S31, while Lnst I was ope-ating on tne RHR closec wnen the SEDS output fuses were removec from S535 Train 3.
system in the hot-leg rectreu*.at ion me<e, valve S70 spuriously j
(The fuses were pulled cy Tony Lees CCVD Control Technician; as.
part of a test of the SEDS. Tony Lees' foreman was Joe Rappa.) As d:-i l
a consetuence, RHR pump
-1 ran for approximately 5 minutes with
~1 i
its suctten isolated. (I ce not kWow whether the mini-flow valve, whtes is supposed to provice a minimum flow of 500 Gp* gnrough
]
tne 9W3 pump performee it s function or not. )There was no control i
roem annunciatton of the problem, but the control room was
~
notifiec wnen a consciencious painter working at the 64' *1evel in t
j the auxiliary building became concerned with the loud ban'ing g
noises the pump was mawing anc callec t,e control room. NpDR
=CC1-S;-CD-31057 (enclosee as Exhisit 3) was inatiated by the Sen.or Control Room Operator to document tne spurious closure o*
l valve 3721. An "On-tne-Spot Drocecure Cnange" was initiatec by i
Fr. Steve = ridley (CVD Operations) to 02erating procecure 3-12::
"Resicual Feat Removal System-Normal Operation". This enan e l
(enclosee as Exhibit 4) instructed the emerator to remove tne pcwer from valve S701 and S702 motor operators after the valves i
i were open.
1 i
f
't NOTE: It is the NRC staff's posttion tna all operator actions necessary to take the plant from normal operation to cold shutdown shou!c be performed from the control room. In
.N Diablo Canyon.3afety Evaluatien Report (SEM) Supplement No.
7, May 197S, page 3-3, p3&E assurec tne N9C t,at "Al'. of the i<
pg operator actions neeced to perform plant coolcown-(except for j
(
periodic surveillance c f One boron concen rat ion) can be i
i 1@
i M
k i
i, J
m
,.-we.es===apinesi.-*w**D*-M-
~
~
l 1
cecomplished from the control room assuming no ingle failure." The above change in-procedure B-2:II requires an
. [ lh h l
c operator to leave the control room to energ t =e and ce-
+
energi=e tne breakers for valves 4701 and 4702 motor
' lud *..
operrators anc thereby violates the above commitment.
On or about Sept ember 30, 1981, I telephoned Mr. Mendorca 1
and told him that the potential. problem with the RHR system whicM I had previously described to him had now become an actual problemV t
On September 30, 1981, I initiated an M-88 form " Material /
(
Engineering problem Report" on the RHR system,. and two G-44 forms
" Recommended Changes to FSAR" (Final Safety Analysis Report) g j
wnien documented errors in FSAR sections 5.7 and 6. 2.
(The t wo G-
)
44 forms are enclosed as Exhibit 5)
On Cetober 10, 1931, I was told by my supervisor to void the.
two " Recommended Changes to FSAR" forms which I had written, and to resummit tne "Materid1/ Engineering problem Rep 6et" as a
" Design Change Request", whien I did. The Design Change Request was sumsequently assigned the number DCO-GE-2513 (Enclosed as Exhibit 6).
l 1,
On Ceteser 29, 1981, the On-Site-Review-Group (CSRG) met to eiseuss my Design Change Request. The* proposed resolution (the
- sj asanconment of tne pressure / temperature interlock system) was f
f' turned down on the casts that this would, increase the probamility 7
)
of everpressurization of the RHR system. * (notes this informa-
.i t:en supplied my the NRC)
,'NC '
]
On Novemmer 11, 1981, I discovered that NDpR #DC1-St-CP-
}(['
u-j Pie 57 nad been signed.off as completed witneut any plant 1
management review,.as requiree by Administrative procedure C-12
)
" Resolution and Reporting of plant problems", and wtthout calling L
for any follow-up action to be taken. One would expect at least-
{3 l
an incuiry into the cause of the incicent and an inspection of.
~
1 RHR pump 1-1 to determine if-it had been damaged. I called Mr.
v.
Steve Foat in the CVD Guality Control Department and asked him to looH into the satuation.
-7' 4
1 On Novemmer 28, 1961, the 05MG met to discuss loss of MMR-*.
suction on 3-23-31. Apparently a NPAR was initiated at this time to perform a pump test on RNR pump 1-1.
(notes this information
{
supp12ed by the NRC) 1 On January 5, 1952, I contacted Steve Foat again concerning i
h:s investigation into the violation of Administrative Procedure i
C-12 during the processing of NPDR # DC1-81-CD-D1057. He said 1
that the problem had been corrected. The NDDR had simply been pulled from tne closed file, the " Initial plant Management j
Review" Llock signed off (by Jim Sexton,,CVD Operations), and the N39 was returned to the closed file - end of proclem. I then 1
(-
contacted Mr. Mike Domre=enshi of the Quality Assurance (CA)
I Oe,a-tment, anc sescri=ed this violation of Admints'tritiVe' '
l
[
w sit
\\
1
,M
..h
. _ _ _.. _. _. 1. _ _
..-.../ --, - -,
-- -- ~ ~. ~. -.'
j _ _.
9
.A Procecure C-12 to him. I don' t recall ever receiving any response.
t
\\l On, January 7, 19S2, I called the Instit ute for Nuclear Dowee
. Operations (INEO) and req uest ed a Licensee Event Report (LER)
{
search %ns all events at Pressuri ed Water Reactors ( PWR' s) i involving the RHR system since 1378.
On January 14, 1932 I received the results of the LER search from INPQ and discovered that there had been numerous similar
~
events at'other nuclear plants.
1 On January 16, 1962, I wrote DCR #DCO-GE-2518-Rev.1 to which I attached copies of LERs describing 16 inadvertant RHR suction i
valve closures at various US nuclear plants during the previous j
three years (Enclosed as Exhibit 7).
On February 26, 1382, I received a " Notice of Special Assignment" and was sent.to the Helms Project, a remote, snowbound site in the Sierra Nevada mountains. I was only given 3 days i
notice of the change, and it was explained to me that I was the only one available who could do the Job at Helms - to compile various lists of vendor prints. I quickly completed the. task assigned'to me, and notified the management both at Helms anc Diasic Canyon that,I was cond. They told me to stay, for unscecific cuties. I recogni:ed at that time that I was being punisned for speaking out on safety concerns at Diablo Canyon, and accordingly, on March 3, 1952, I returned home as an act of
(.-
resignation. It was approximately,one ween before tne PGSE
(
management realized that I wasn' t in the Sierra Nevadas any more, and called me to inquire what had happened.
4 6
In July of 13S2,'while working as a Startup Engineer at the 4
Palo Verce Nuclear' Plant near Phoenix, Arizona, I was e'oSt' ct ed
~
a by an NRC Representative from Region V. He asked to talk to m,e concerning the Diablo Canyon RHR system, anc I agreed to meet him i
tne next time he was in Phoenix.
On August 2'0, 1982, I met with Mr. Eugene Powers cf the NRC Of fice of Inspection and Enforcement at the Camelback, Inn in Scottsdale, Arizona. During tnis interview, which was uncer oat.)
and tape recorded by Mr. Powers, I discussed the following:
- 1. The problem with the control circuits for the RNR suction
)
valves 6701 and 8702, and how these valves were prone to i
spurious closure.
I gave Mr. Powers copies of the circuit diagrams for this system, and the LERs descrioing similar
)
problems at other power plants. Fy major point was that j
routing the control circuits for valves 8703 and 8702 *through the Solid State Prot ect ion Syst em performed no function what-
!;d.
)
soever, reduced the reliability of the RHR system, and caused a potential for camage to the RHR pumps. I stated tnat 'the pressure / temperature inter *.ock system should be wired dir-(
ectly from the Control and Protection Racks to the.; Auxi-liary i
e, Safeguares Racks wit 9eut going tnrough the 3335, or tnis i
I s
$} Q l
3
. ~.
?
-l N $ $ ?,, -
m.,_.,,
~
,a
cyatcm chould be ebendoned altogether, since it wasn' t being e.
y' used anyway.
(~
f
- 2. The fact that there was no electrical schematic or logie
[
diagram which incicated that removing the power from the SSPS l
s ou,tput relays woule cause valves 8701 and 3702 to fail closed.
I
- 3. That there was a problem with dissemination of f
information at piablo Canyon. Speci fical ly, operating experiences at other plants were ret mace available to the plant Maintenance -
Personnel, and in the case of the RHR system, even the plant Management was apparent ly unaware of problems at other plants. Also, althougn I mace the PG4E management aware that N
removing the SS;S output fuses would cause valves 4701 and 5702 to fail closee, they apparently never passed on this l
information to those under them.
t
- 4. That the loss of Decay Heat Removal capability on 9-26-61 lt8i was a "significant event" as defined in le CFR 50.72, anc that by law, the NRC should have been notified, but wann't.
(,
5.
7 hat the manner in wnien NPDR #DC1-81-CP-A1057 was handled in violation of Administrative Procecure C-12 "Reselbtion was anc Reporting of Plant Problems",,sihee the N3DR was signec off as complete without any initial plant management review, i
investigation into the rect cause of the proslem, or follow-I up action suen as an RHR pump test.
G.
(
That there is no control room annunciation of low 6f-RHR flow j
or pressure, auc hence the RHR pumps may be damaged before qd tne operator is aware of any prenlem.
I recommenced tnat annunciator alarm be insta11ec. (Unknown to me, in Safety ly,11 an l
Evaluation Re; ort, Supplement e 13, the NRC staff NId* required
.I' PG33 to install suen an alarm during the first refueling
,cg d" outage.)
tF 7.
That the F3AR cescription of tnis system was incorreet,
~
anc DGar refusec te enange it.
d,, ',
S. That I had repeatecly notified :G4E management (inclucing a memo to Mr. Thorn =erry, the plant manager, which is enclosed as Exhibit 6) of an error in Plant Manual, Vol is, but tnat the error had not been corrected eight months after the original notification.
9.
That NPDR's DCO-73-TI-P0006 and DCO-73 I-20117 wnsen*:
had written to document erewing errors tnree years previous *y were still unresolvec, and apparently unreviewee.
Lf6 i
- 10. I askec the NRC why the RHR suction from the Reactor Coolant System (RCS) het legs was not, considered Safety-Related (and hence mace recundant) at Diablo Canyon, but wg3
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consicerec Safety-Related.and gas reduncant at
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plants in the US.
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9 During the interview, other items, not related to the RHR system were discussed, and I gave the NRC drawings and written f'
information to support my position with the understanding that i
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,this material would be copied by the NRC and returned to me.
l l
4
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j y
On August 20, 1983 after not receiving any communication,
- j from the.NRC for.one year, I wrote Mr.. Eugene Powers and
.t requested the return of the material which I had given the NRC, '
j, *
}
and an explanation of what action they had taken, if any.
l (My letfer is enclosed as Exhibit 9) i 1
In September, 1943, my materials were returned by Mr.
Powers j
j witn an apology. He went on to say that my conc' erns had been
).
j addressed and that I woul'd be contacted.
j i
On October 17, 1983, I received a letter from the NRC signed a
by Tolbert Young for Mr. T. W.
Bishop, the Resident, Reactor j
projects and Engineering programs, Region V.
Mr. Bishop also apologi=ed for the delay in contacting me and said that an investigation had been conducted.
He included copies of Inspect' ion Reports, Nos. 50-275/S2-25, 50-323/50-13, and 50-j 275/82-42, which had supposedly addressed my concerns. These inspections Eere conducted on Septemmer 17 and December 2 of 1982.
1 (tne letter and i
inspection reports are,, enclosed as Exhibits 10a-c) s-j On Octoeer 20, 1983, after reading the amove-referenced inspection reports, I felt tNat tne NRC had only gone through the f
motions of an investigation into these problems, but tnat the l
i s,
questions which were asked of PG8E were trivial or incorrect
]
(
representations of those concerns which I conveyed to Mr.'
Powers.
! 3:
In acettion, it appeared tnat AG4E's answers to even those ques-i t
tions were acceptee witneut questier, or follow-up. In some cases, j. *.
tne answers PGA3 gave to the NRC were false. For example,iduring j
the inspection, pG4Z management claimee ignorance of any errors i
l in Giant Manual, Volume 16, even thougn I had written memorandums j
en the subject to botn Jonn Giscion anc Mr. Thornberry, the plant i
manager, and given copies of these memos to tne NRC the pEevious j
August.
1
-7 The NRC inspector performing the audit appeared to be i
i unaware of the basic NRC requirements for an RHR system. For l
example, in his inspection report he " concluded that there i
appeared t o be no requi rement or necessity" to have a control room annunciation of low RHR flow, even tneugh in April, 1941, the NRC staff, required PG4E to install suen ris, al arm.
In other instances, the NRC confirmed the facts of regula-j tory violations, but then ignored them. This was the case with tne requirements in 10 CFR 50.71 to maintain an accurate, current
{
FSAR:
le CFR 50.71 Maintenance of records, making of reports 4
i f
(e) each person licensed to operate a nuclear power reactor
\\
pursuant to tne provislons of paragraon 50.21 or 50.-22 of
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thic part guill M2(313 ggriggiggily, as providME in v*
3 paragrapns. (e) (3) and (4) i f this section, ing fingl 33f313
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i A011%211 C222ri 1E3531 originally summitted as part of the
.N apelacation for the operating license, to assure that the
/
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information included in the FSAR contains tne latest material
/
developed.
i This submittal shall contain all the changes
- y rrecessary to reflect information and analyses submitted to l
the Commission by the licensee or prepared by the licensee i
pur'suant to Commission requirement since the submission of l
}
theteriginal F5AR ori as appropriate, the last updated FSAR.
j The' updated FSAR sna11 be revised to include the effects of t all changes mace in the facility or procedures as described l
}
in tne FSAR; all safety evaluations performed by the licensea l
either in support of conclusions that changes did not involve an unreviewec CNCTE: now called unresolved 3 safety question; anc all analyses of new safety issues performed by or on j
benalf of the licensee at Commission request. The updat ed
! LIy, information shall be appropriately located within the FEAR.
(empnasis added)
)
l (3) (1) A revision of the original FSAR containing those l
j original pages thas are still applicable plus new replacement i
pages snail be filed within 24 month,s of either July 22,1540, i
i or the case of issuance of the operating license, =nienever j
is later, and small bring the FEAA.up to case as of a maximum j
of 6 montns prior to the cate of filing tne revision.
.{
1 eecause c? these, anc ether inadecuacies in the NRC investigation f
i
{
(
ir.to my concerns, I wrote a letter to Mr. Bisnop of the NRC remutting their report. (This letter is enclosed as Exhibit *11) i I
(
i On Cetoder 27, 1383, a spurious closure of one of tne RHR I
leg suction valve's caused camage to an' RNR sump, not 8
after was The RHR puEp*'su'it j
run for one nour w'ith its suction isciated.
s-tained cartage to it s seals, and a bent shaft, and Mad,to be replacec. + (note: this information pro.videe by the NRC)
's '
t Over the ' course of November and early Decemoer, 1943, Mr.
i Jesse Crews of NRC Region V telephonec me several times to ciscuss my allegations. Mr. Crews offerec encouraging responses s
]
tnat essentially concedec that I had a valid critteism anc sq '
i t.t p l i e c t h a t the problems might be fixed.
He did not make any-.*
l commitments to tnat effect, however.,
b On Decemmer 21, 1943, I met with Mr. Crews and another NRC Q4 Region V representative, Al Johnson, for approximately 2 to,3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
I was accompanied by Mr. Devine, wne servec as my lega*
representative at the meeting.
I have prepared a rougn transcription anc commentary of our meeting, whien,was tame recorded. (The transcript is enclosed as Exhibit 18) A review of
}
t9e transcript is revealing both of PGAE$ s breakdown in organiza-tional accountaniitty, and the NRC's lack, of response.
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F.
SUMMARY
OF NRC INTERVIEW On page 4, Mr. Crews complitnent ed. my diligence and stated F
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.that I deserved "a great coal of credit" for making the RHR issue a thing of praiority within the NRC's Office of Nuclear Reactora Regulation (NRR). That brings up a major point of this wnole exercise Why is pG4E unable to handle simple engineering '
i prodlems like this within their organization. Why is it necessary for an espployee of the licensee to have to cause.a " focusing of 1
the Comewission resources" and thereby place his Job in Jeopardy in order to force pG4E to correct a minor problem whien they should be eager and willing to fix?
I, i
I On pages 4 and 5, Mr. Crews mentioned the existence of
)
acministrative centrols on tne valves. (Adrainistrative controls are procedures which direct the operators and maintenance person-J nel to turn of f the circuit breakers for these valve motor -
l operators after moving the valves to the desired position. ) This i
l raises the question of unauthorized changes to the plant g,5 l as
' Ig )
1 described in the Final Safety Analysis Report (FSAR). Removing 4
4 ene power from these valves modifies the operation of tne RHR
_ system and cisables tne pressure / temperature interlock system as Ti coscribed in the F5AR. Since the RHR system descriptions found in l
the c5AR were used as the basis for botch 'pG4Es' anc the Staffs'
,;.g i
analyses of postulated accicents involving tnis system, taese
,,g analyses are no longer valic witn tne power removed frorn the i
va*ves. 10.CFR 50.53 specifies that the holder of an operating i
j
(
license may not make changes in the facility as described in the ij FSA1 if tne enange involves an unresolved safety questien. The I-
)
{
RMp. system g2gg involve an unresolved safety question (Ta s e< A-1 I
j 45). It seems as thougn PG4E has violated 50.E3 by tur-ning off g
j the pc er to the suction valve operators witneut prior comentssion g
l approval.
." m
.e L
On pages 5 and 4, Mr. Crews revealec that in October, 1943, I
h j
en RHR purnp at Dianto eenyon was severely damagec by yet spurious closure of or'.e of the RHM suction valves during,another=
a surveillance test. According to the NRC, "inacequate,peccedures" caused the accident. The " incredible" had occurred for the second i
tirne in three years.
t l
On page E, P'r., Crew's revealed that pG4E apparently is reaI:h i
anc able to fix the cesign flaws..That is not surprising. The most troublesor.e design flaws could be corrected merely my rerouting fouva pairs of wires and adding an ann'unciator alarm
!'f input from two existing low flow switches. 1 estimate that irme s-l total cost for these changes would be less that sie, eco,
' i including paperwork and drawing revisions. Certainly these I'
changes would cost less'tnan a replacement RMA pump.
.s.
s i On pages & througn 1, we discussed the respense within pG4E I
i wnen I first raised tnis issue; To my knowledge, the aesign j
(.
cnange request wnien 1 initiated has been sitting in some l
3 engineer's in-casket s i nce-1941, unreviewes and une,eso.1v.ed,. unt il j f I
,15 1
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h my continuse proosuro en the NRC has only recently horced a ph
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serious review of my allegations. The CVD On-Itte-Review-Group i
{"N gave snis Design Change Request a hasty anc flawee review in Cet o =er, 13ai, but ceciced
.[--)
to take no action on the proelems in tne RHR system. The NRC considers this an " adequate response",
j
/
but.I disagree.
The 03RG, ef fectively vetoec the Design Change Requesta grous of managers anc administrators, in a single,
' factually erroneous sentence, after a " review" of unspecified
)
metnodology. The results of the review were not documentee on the l
}
DCM fory'a~d returned to me. The CERG did not, at any time, n
i engage in any communications with me.
The NRC appears satisfied that the issue,is finally being i
seriously reviewec, two and a half years, two spurious valve closures, anc one Junked RHR pump after it was first raises. Mr.
i Crews excusec the long celay my stating that PG4E had other
{
priorities, such as the ongoing eesign verification program. He also speculated that the pG4E was afraid to make a change that woule create an unresolved safety question. (I have seen no 1
1 j
cocumentation to support' either of these positiend, so it may be assumed that they are conjecture, on the part of Mr. Cruse)
{
l 1
If G42,was too busy to consieer significant questions amout
{
t.*e reliatility of it s Emergency Core Cooling System, then what pt; c;9er serious questions concerning the. design or construct ion of
,,5 1!
j t'i $ '
Dia-;:
Canyen may nave gone unansweree? The position tnas G4Z co.ic taee no action in this case because tne solution to tne g
i pr: = l ere j
(
" signs involve an unresolved safety ques' ion t
a Just coes j
no: maxe sense. Woule p332 rea* 1y have us believe that it gives j
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more att er.tien to routir.e Design Chanae Requests than those more l
serious OCis anien might involve a safety question?
s 1
Cr. pages 9 anc ic, I askee what facet of the N9C's program
{
prever.ts A3&I frem Just not acting on problems that haie'seen
}
inentifiec, sut for whatever reasor., the utility enoeses not to j
act upon.
F.P. Crews responded witn a reference to the NRC 2
i
" reporting requirements snat
- f tnere's any discovery of.a defect f
in a system or cesign...which coule involve substantial safety j
consicerations, a
somewhere in our nocy of reporting requirements tney' re going to have to report it." He was probably referring to I
10 CTR*50.55, which statess i
10 C.TR 50.'55 Concitions of construction permits (e) (1) If the permit is for c'enetruction of a nuclear power plant, the holeer e.f the permit small notify tne Commtssten of eacn deficioney found in design and eenstruction, ub1's3s es:s 1112 bara essaires sessersstass ss211 haxe affssies asynessix its safsix sf estralises sf aus casimac asmsc i
alsca si acx 11ms sursusceut its smassing 11fs11st sf the 214216 geg wnten represents:
i
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(ii) A significant cef tetency in final design as approvied
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and releasee for eenstruc31oe suss 12s1 1hs CasAsn.512A csi
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22cf2:2 is abs sclascia 20s Dassa slaisa in are aAfsix ALA1x:
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51S C222ri_cr construction permit. (emphcoin cdded)'
N E.
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In this case, of course, AG4E didn' t make any kind of a
\\s report to the NRC. The NRC excused PG4E s failure to report under
[
10 CFR,50.55e by saying that:
- 1) It doesn' t consider this a 112Difi22Di deficiency.
- tf The' NRC excused PG4 E' s f a i l ure t o report u nder 10 CFR 59. 72 4
l by sayigg that
,t e
}
1)*Although these regulations might seem to require reporting the of loss of decay heat removal capability at any time, that was not their real " intent".
l n
- 2) There was no fuel in the reactor, therefore no danger'to i
i the health and safety of the public.
{
i-The NRC excused pG4E's ' failure to report under Section & of the
{
Technical Specifications.by saying thata
- 1) The plant was in no " mode"-(i.e. Refueling, Cold Shutdown, Hot Standby, etc.) as cefined by the technical specifica-
- tions, therefore no violation of the.. Technical Specifications was'possible.
4, J
Cn pages 21 ant 22, Mr. Cruse die concede, however, tnat the less of decay heat removal casability would have been reportable
(
under 10 CFR 50.72 as a "significant event" if tne plant has been l
i operating."
~ '
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~
This concession by,Pr. Crews adds weight'to the argurev.
that OG32 violated the reporting requirement of 19 CFM 54. 5L(e),
which requires cisclo' ure to the NRC of the discovery of.signif t-s I,
cant design deficiencies that could adversely affect the plant's safe operation at any point in it s li fetime. Surely discovery of vulnerability to a significant event vould qualify. _ 5ection 50.!5te) applies to construction permit holders, whien was pGSE's status in May, 1381, when* I first repcetoe the problem to my supervisor anc to Mr. Giselen. Ce*mplience with 10 CFR,50.55(e) is a cor.dition of the construction permit.
PG12 did not responc to my original May it, 1941 letter texcept to threaten the less of my Job if the information got i
out side the company).
In fact, Mr. Giselon did not even document l
the problem on a NEAR until after I called the NRC and.Mr.
yendo=ca asked him macut it. (A person wne becc*es aware of an I
actual or potential problem in the plant or in a procedure is i
i required by Administrative Droceeure C-12 to document tne problem i
on a NppR.)
In the subsequent NRC aucit, a proslem report *was I
wr:tten to document that removing the output fuses from the ssp 3 would cause the RHM suction valves to close, and to revise Emee-d l
gency Operating procecure 0P-3. At this time no analysis of the i
f l
?
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basic cause of the pros *em was performec, and no constseration of 1
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- w.. at other peccedures mignt be affected was documented.: Neithee l
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j did tho NRC toks Gny ceticn ogainst PG4E for violotTon of
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4 Acministrative Erocedure C-12.
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On pages 12 and l'3, Mr. Crews conceded that the NRC has no i
[
]
. fimec criteria to determine whether a utility has displayed a l
breakdown in organizational accountability requiring enforcement action.'
Further, Mr. Crews described his approach to evaluating j~
'the organization as either " sort of Setting a feel for the
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personalities involved" or looking at the "better document ed" l
cases. The approach of looking only at the well documented, cases t
)
iherentTy misses the cases where the utility has not responded at I
all, thepehy generatin'g no documentation. I believe that this i
weakness compromises Mr. Crews' review of the PG4E organization.
i i
I Cn page 13, Mr. Crews referred to a " communications" problem I
i
- hat prevented the design flaw from being resolvec. That is an -
j l
l uncerstatement. General Construction personnel could not. initiate plant protlem reperts, which left them only the option of writing memos, as I did, in order to notify the operations civision of a l
probi er.1 in the plant, or with their precedures. I. was r1tprimancee fer wr: ting such a memo directly to the responsible maneger, s;
wit 5eut going through the proper chain of command. The, proper enain of coramanc whien was outlined to me was as follower 1 - My
- ,i su;ervisor, 2 - tne Electrical Resident Engineer, 3 - tne Constructien Superintencent, 4 - the Vice presicent e* General C c.stru= tion, 5 - the Vice 3 resident of Coast Valleys Division j
.N.c w see: ear Power C=erations), & - The Dia=1o Canyon #1 ant i N q
Ou;st intsacent, 7 - The Department beam of tne affectec j
(*
i r.c. vi e w al, & - The incivieual to wnom the original umeno was 4<
3 accreasec.
t 1
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4:
1 Cr. ; ages 15 ar.c 15, Mr. Johnson clearly revealed tne limita-tions o' 'G: regu*atien.
I have always felt that Just as it is impessi:' e t e legis* ate morality, it is also impossib1'e'1e legis-
! ate (2ality. True quality requires the personal commitment o*
al; parties 16volves, as well as legislative guidelines. Mr.
Jennson ex;;ained that it is'not a regulatory violation to camage j
vel pumps. :nereculeus, : inquires, "You can stanc tnere anc say t
i it's net a violat ten... it doesn' t nother you to say that, when pum;s are teing wreckee all over the Country?" To wnie5 mr.
J:.r.sc *. resp:r. cec, "chnolutely not..."
he went on t.o say that
"*r::.* a sa'ety stand;eint... when tnis thing happens you' re not,.in a ;esition where you need these pumps.
If you have something go wrong, you re;1 ace the macnine." (Whien is exactly wnat tne A342 ratepayers di: last Getecer)
It seems obvious to me, but apper-ently net to the NRC, that the proper time to correct a promlem
}
is ag*2C3 tne equipment is needed to mitigate sne consequentes of an accident and the prehlem becomes a canger to tne healt3 anc sadety of the public. M r.
Jonnsen empnasized a cistinctien tnat must always he consideree: "You can take a safety injection pump, er any pump at 491s plant that's not needed for snat paase of l
operation - say it's cefueled - therei s no fuel in it - if sney I
want to go and tear up their pumps, taat's not a safety proslem, j
g sut it's dumb and it's an economic proslem, mut i t ' s not a safety j
33 prim;er - it's not a violatten. So you' ve got to a; ways _ separate l
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P those t,wo." This attitude displayed by the NRC brings in to acute i
focus tnat the public's concept of the function of the NRC is i
g, h vestly dif#erent from the NNC's concept of its own function. It f
is discouraging to learn Jutt how limited the authority of the j
f NRC actGally is in enforcin; quality, which.is the founeation on i
which nuclear safety is built.
.i:
On.page 17, Mr. Crews mace a factual error which exposes a serious weakness in the PG42' structure.
He i ncorrect ly, assert ed, "Those gUelear plant problem reports can be initiated by f
a n y o n e.*.. "
In fact, problem reports may gely be initiated by NPQ personnel. Construction and contractor personnel are not allowed to initiate NEERs. That it why 1 was forced to write memos to j
make NGO aware of the promlems with the RHR system.
Similarly, on page 13, we discussed the fact that there.is no provision of the Design Change Mequest procedure requiring a management response to the originator.
This may. explain how the defects !
icentidied in the Diablo Canyon RHR system could languish foe
{
years witnout being seri,cusly adcressed.
- l
}
Cn,; age 15 we discussed my allegations from the auditors perspective. In orcer to learn whetner an organi=ation is capable o' success *ully resolving problems, an auditor looks to see tnat wnen a pro *.em is identi*ied to the organ,',i:ation, steps are taken 1
0:
11 'cetermine the root cause of the protlem, 2) cetermine the
)
f.!!-exten:
whet 9er tnere nave teen any similar types of I
pr:: leys, '
take corrective action to insure that tne pron-and 3) i late won' t reoccur in t 7e *ut ure. ?G32 nas.failee tne test on all f
tMree o' tnese criter:4. In fac;, SG42 re*uses to acmit tnat it 0
4
}
(
sig e pro:lem in $51s system, *et alone taken tne first step Nd I
toas.rt selving it. As previously mer,tioned, tne root cause of tne j
spuricus Idi sue: Lon valve closures is poor cesign o* the inter-
't lock cir=uitry controlling these valves, compouncee by inacequate crawirgs of sne system, and the lace, of control.reem ashunciation
?
},
l o a* ert the operators of the prealem. FG S E, however, continues acamar.t'.y to cling to tne bankrupt philosophy, twice p'roven
$ 't f
j incerrect, that tne root cause of tnis pro =lem is "inacequate
}
precedurse" er."*ack cd planning". This problem will never be
{
st
- vec until 9342 admit s that t9e pres; ems witn this,, system are j
in t e eeuipment, not the prececures.
t ve. Crews fermally washec his hanes of tne secone anc ta:Fi j
au=iting criteria: "I' m rea* ' y not prepared to offer you any j
steps $9at we' re going to take formally, bases on wnat we' ve i
seen."
In suppert, he asserted tnat EG4E's correct ive " steps i
j we-e reasonably e*#ective... in dealing with the sumstance of my concern." How can the NRC say that the pG42 solut*en was t
reasonably e**ective when the problem cont te.ues to reoccur? What active steps did PG4E take in this case?
I' m net aware of gny action taken by DG4E o' its own volisier.. unless you allow tnat i
tne decision by the CSRG to tav.e no action, is in itself an r
I action.
a
)
j
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From pages 04 to li, we ciscussed anoiner prosjemJ 4
.nacequase appecvec cesign drawings tna: " ail to snow :nat the s
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t'Q, pewer coureo for eartcin control relaya in the RHR Eystem comes V-I.
from a totally unrelated syst em, the 55DS..There is effectively d'
i
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no way for a competent. tecnnician or engineer to ascertain tnat L'
removing the power in the S535 will cause the RHR valves to
/
I
,close.. Mr. Crews coneeced the inacequacy of the existing
_s crawings, but explained that a new, accurate composite crewing had.
seen preparec.
Unfortunately, he cic not feel it necessary that
,'i 231s composite drawing become a part of the plant's approved ces.ign document s.
Rather, the accurate version is informal, and was given to tne technicians as a tool "to sort of help everyone I
better enderstand it..."
This attitude is not reassuring.
I
~
5ac prepared a similar crewing some two years earlier, anc_it was j
ciscarded or " lost". The technicians certainly aren' t the only people with a need to know how this portion of' the plant operates. Operators, Engineers, procecure writers, or the NRC might have a neec for this information at a later date. The i
assurance of the drawing control syst em is that accurate, and up-to-cate informat ion is readily available to any person who mignt need it.
,t The formal drawings of this system remain cefective anc
- naesquate. Mr. Crews tried to reassure me that this "new" composite drawing woule.always be availamle in the future because "y.c. ave to maintain as-built drawings anc schematics of the j
- 1 ant, anc these will always exist..." He cic not succeec in
]
cor,vincing me, =ecause the "as-tuilt",crawing of this system w : c.. I placed in t_he Instrument snop files in August of Isa:,
j mu-was apparently "lest". What assurance can he give me that i
t..:s re crawing will always he available.to those wne need it?
I
(
- y original competite crawing may have been lost in tne i
7 -s g-en-f:le purge of I?!2. karly in 1942, that year, tne CVD l
(
1rstrumentForemen{hent t h'rou gh the loop test files in the
- r.s rument snop anc removed anc costroyee all.hanc-drawn, "as-
.=si*t" erawings tall crawsngs and information weten were, ret part. '
of t e crawing control system). Apparently, the Justiff'eation for s
.,inis was tr.at tne NMC was very concerned that " unapproved" l
~~
crawir.gz were being usec in the plant.
In this manner, tne
[
techn: cal literature and i
notes wnich had seen collectec *er years 7
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dy t,e nstrument anc control technicians simply disappearec.7 Wnat :s to prevent tnis latest = composite' drawing of the RHR '
systen from being " lost" in a similar manner unless i"t is ineo--
- oratec into the official plant drawing control system?
\\
j Dn page 29, tnere is arether illustration of tne NRC going out of its way to excuse PGSE misconduct. As I previously mentionec, I hac complained to the NRC that an error wnsen 1 I
discoveree in the plant Manual, Vol is " Annunciator 8eseense" was
- still uncorrected 4 months after I had first reportec it.
i NOTE: Thas problem was ultimately not corrected until January of 1983, over 20 months after I icentifiec the proslem. Here j
again, AGAI toov. no action until the NRC aucit, wnen a i
problem report (NppR NO. DC1-83-TN-90001) was initiates. The l
NRC, true to form, made no note of tnis violation of acminis-trative proce:ure C-12 eitner.
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T83 " Annunciatcr Rasponso" boo.-k io o recdy rafareneo manual 1,
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s for the control room operators which provides them with a list of l
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all inputs which could, bring in each annunciator alarm, along l
^
with possible causes of the alarm, The' error wnich I found was in
,tne Reactor Coolant Pump section. This section incorrectly listed i
j the Reactor Coolant pump oil level alarm as a low alarm only, when in reality it is a high-low alarm. I first pointed.out this
' error in. a mek to Mr. Giscion, written on 4-24-81. I received a reply on 6-9-41 from Mr. Rob Fisher who stated "I am revising the 4
i manual and have made a change to reflect actual conditions. "
The f
y following December, the change has still not been incorporated, i
j so I wrot'e another memorandum to Mr. Thornberry asking what had' j
happened. Two weeks later I received a reply,,again from Mr.
j Fisher, stating that the change would be incorporated soon.
I described my concern - that the control room operatoes at
'l j
Diablo Canyon were operating with incorrect information due to i
the excessive amount of time it was taking to update the manuals
{
- to Mr. powers on August 20, 1982, as previously described._
In i
the ensuing audit of pG4E on this " allegation", pG4E representa-l tives claimed that pG4E 'was "was not previously aware of this
]
pro bl em,". Although the NRC was in possession of my memos proving 4
otherwise, this false statement was allowed to pass unchallenged.
In the inspection report the NRC merely. printed pG4Es statement as if it were accurate.
pecording to this same inspection j
report, both Mr. Giscion and Mr. Thorn, berry were present at tne d
exit meeting, and presumably had copies of tne NRC's finding, but enose to remain silent.
l 1
.97 Crews at t empt ed to explain away the false statement with j
f the following hypothecatec excuser "Mayce the NRC did not bring
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tne topic up et the meeting".
The point is that it came up at
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least once - the NRC report quoted the cisclaimer - and the response was false.
The Commission should learn why, and the
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l entent of the problem.
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On pages 31 thougn 33, Mr. Crews revealed the misplaced allegiance of the NRC. He said he felt that the staff would be willing to permit operation of Diablo Canyon without resolving l
tne RHR* issue, because~other plants are operating in a similarly i
vu*reeable state.
As a mat %er of " fairness", Diablo Canyon l
s3ould not.se held up.
That raises a beste question, fairness to I
l whcm? To the utility that owns the nuclear plant, or to the I
pubise w.3o must endure tne consequences of unsafe operation?
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Mr. Cruse's final, analysis was that he feels comfortable with the plant emerating in its current eendition.' His explanation l
again fail.
to provi'de the intended reassurance: "you may not be l
able to bring it Edown3 on RHR, but you could being it on the
)
steam generators - you can do that - we' ve done it... and you' re i
right, they couldn' t get on RHR, but they could very likely use gus,
i the steam generators. ' You keep a steam generator
- intact, and j
natural circulation will bring 'er down.
1 1
The overall impression lef t me by the December meeting l
was
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DG8E's organi avional miseenduct and leen of effective.aceton that the NRC was still more concerned witm finding excuses for
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than with forcing PG4E to correct these problems.
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Apparently, at t he t ime of t h i s Decem ber 21 st meeting, the NRC staff hac already determined that no action would be taken on f
any of'my ellegations. Many of the responses I heard from Mr.
Crews during that meeting re fl ect ed precisely the finoings
,j
,containee in NUREG 0675, Supplement 21 - The Safety Evaluation Report (SER) on Diablo Canyon.
On January 7, 1984, I received e 4
eppy of,this repert, whien basically formali=ed the philosophy i
presentpd more tentatively by Mr. Cruse and Mr. ' Johnson during the meeting.
Overall, with respect to the issues I raised, the, l
SER represent s a determined ef fort on the part of the NRC to j
excuse violations of'the Atomic Energy Act and. Ener2y Reorgani=a-tion Act. (My detailed critique of the SER response to each alle-gation is enclosed as E On balance, the NRC's handling of my RHR system allegations l
can be characteri=ed as first, an attempt to ignore the proslems,,
t then an attempt to whit ewash them, followed by a still more sophisticated attempt td wnitewash them. The " engineering review" performee by NRR reveals a basic lack of understanding of how the l
RWR suction valve interlock circuitry operates, contains numerous centradictions with other statements ma,d,e by the staff, ane proviced no evicence to support its conclusions, as if the mere statement that tne plant is safe is sufficient to make it so.
?.t e, t.is discencerting document brings in to question tne astlity o.f One NRR staf f to cenduct en unbiased review of tnese allegations.
~
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At this point, after a thorough reading of the work of the C
NPR staff, I recommenc that.the wnole question of the adequacy of the Dianlo Canyon RHR system be t urned over to an indepeneent a "second opinion".
I group of qualified en*gineers for I!. GI.NER*C PREAMDOWN CR THE PG3E ORGANI2ATION ~IN THE, EFFECTIVE RESCLL'TICN OF PROBLEMS The lack of response I have received both from PG3E and from tne NRC te my RHR, system allegations appear to indicate a larger problem at Diablo Canyon, and w % teansd u,cze u pgs From my f
esperiences, : een only conclude tnat at as the unwritten policy ef tne PGAE management to attempt to cover up any plant ceficiencies wnien. if admitted, mi ht dela-
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During my first two years at Diablo Canyon, as an Inspector
- or pG8E General Construction instrument ation department, I was positively impressed with the quality of the work bet ng cone in our group. At no time, during tnis period, would my superv1 sors allow any shc4cy or incomplete work to be performec, or any design question to go unanswered. At that time, the General Construction Instrumentation Department was run in an effietent,
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p-ofessional, and quality,_ conscious manner.
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r, B$ ginning in January, 1978, I took a job with pG4E Coast U
Valleys Division,' the division responsible for the actual
{h operation anc maintenance of Diablo Canyon.
This job was a j
f permanent position as a Control Technician; I made the change
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from GC because this new job would allow me continuing employment in the. area. It became immediately apparent that the management
~
I of Coast Valleys Division had an entirely different attitude toward n,uelear safety than did that of General Construe' tion. In the CVD instrument department, problems were routinely ignored by.j j
the manjheme,nt; reports documenting significant plant deficien-eies were either " pigeon-holed" in the foreman's or supervisoe's 4
in basket - sometimes for years, or just thrown in the trash without notifying the originator. As an example of the former case,'in August of 1978 I discovered an error in the wiring of the RHA suction valve This problem was documented, interlock comparators and relay modules.
along with the proposed solution, plant Design Comment #403, dated 8-15-78. As of March the fol-on lowing year, although a simple wiring, change would have correctedabsolutely no action it. Finally, 1
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when instructed to calibrate the equipment 'in question, I corree-ted the. problem myself, and was reprimanded by my supervisor for 4
doing so. Hence the entry in the unofficial " shop log" stating:
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"You BD2w where management's priorities lie when your super-
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visor instructs you that a defect in a safety related system is not consicered a nonconformance, b u ~. correcting the defect d
without written approval of the plant superintendent 13 a I
nonconformance. Sometimes it's hard to remember that our job i
is only to iggetify problems." (a partial transcript of the t
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" shop log" is enclosed as Exhibit 12) i
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Examples abound where pGSE has been slow to respond to problems.
NppR DCO-79-TI-p0117, documenting a simple drawing change, was still unresolved f i
2MC 222C1 after I initiat'ed it.
NppR QCO-73-TI-90006, documenting another erawing erece,
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took over three years to close out. (My memo to the Quality Control 1
Supervisor on this subject i s enclosed as Exhibit 16.) These are
)
not isolated cases, but typical examples of what I consider j
excessive time celays in correcting simple problems..(The NRC's i
definitive statement on this was that "no particular safety or regulatory significance could be attached to this situation.)
As an example of the management simply destroying problem',
reports, tne " shop log" also contains a list of 8 problem report i
numbers that were " voided" There were many other problem reports l
that simply disappeared, as a review of the plant NppR log for 1378 and 1973 will show.
4 problem reports documenting errors in
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safety-related procedures were routinely " voided" by the management.
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j Not only did the management destroy problem reports which had been written by the technicians, but the technicians were discouraged from writing any at all. On July 13, 1978, I was told
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by my foreman not to write any problem reports without his approval in advance.
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Ratcliation won takan against tcchnieicnn who "found too
,s, many proclems", by assigning them to the more distasteful jobs at
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the plant. As an example, tchen I wrote problem reports documen-f ting errors in several safety-related procedures, I was assignec
. to rewr'ite these procedures, although this was not part of my job cescription. I was given a table in the corner of the "cof 'ee i
room" in the temporary administration building, anc isolated from all my co workers for 4 months.
Anfther form of retaliation taken against whistleblowers was to give them po'or performance reviews, which went in their persorr-nel file and might prevent them moving up in the ecmpany. As an example, my 1973 performance review while a CV.D control technician contained the following statements:
" John is a skilled craftsman in troubleshooting and cali-l beating.
However he is intent on redesigning most equipment he is assigned to test or repair or rewriting most procedures,
he is assigned to perform. To complicate the process of evaluating his critici=ms (sic) of equipment design and pro-eedures John usually becomes indignant and recalcitrant if responsible supervision does not agree with him or respond as qu2ckly as he would like."
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" John meet s the overall job c'equ'irements.
However he tends to deviate from assigned tasks into areas, such as design of the plant, for which he is not responsible.
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i would be more valuable if he would concentrate his efforts on
(
assigned tasks."
i The CVD technicians during this period (here routinely denied j
access to necessary information, such as newly issued or revised I
plant crawings and NRC ISE Bulletins describing proelems at other similar plants, necessary to perform their JobsyfOn Apr'il'3,1979, 12 technicians from the CVD instrument shop signed a letter i
formally requesting the necessary information. (A copy of this l
1etter, with signatures omitted, is enclosed Giscion, the Technical Department Head, accressedthetechn{zeians as Exhibit 13) Me, the following day, saying "I can' t believe you people knew what
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S you were s2gning", accusing us of " rabble rousing", and basically i
telling us to keep quiet and follow orcers.9 2n the question anc I
answer period following this speech," the poor condition of our-*
safety-related procedures was brought up.
I pointed out that the'se procedures were not kept current or accurate.
Mr. Giscion sazd that he couldn' t believe this and asked for evicence. Imme-j diately following the meeting, I wrote him a memorancum documen-i ting several current deficiencies in our procedures (enclosee as Exhibit 14).
The problem of inadequate reference materials was not cor-rect ed. Although new drawings were routed to the instrument shop
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for a few months after the request was made by the technicians, i
tnis practice was discontinuec.
At no time were NRC bulletins or
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any of the other requested information routed to the instrument shop. As a result, the tedhnicians were forced to gather wnat i
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.information they could on their own. Informal drawings, vendor
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instruction manuals, catalogs, and notes from other plants were
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collected in informal files.l,Early in.1962,. however, the manage-
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ment destroyed these. files, apparently fearful that the NRC would
. cite th'em for the use of " uncontrolled document s".
(This was I
hardly likely considering the NRC's reluctance to cite pG&E on clear-cut violations of federal regulations.)
_~
Mr. Giscion's response to the discovery that the safety-related$* procedures for which he was responsible were defective,
was not to re-evaluate his department's policy of discouraging the writing of problem reports on procedure errors. Instead, it was to chew out my foreman, and make him work overtice to write a response to the procedure problems which I had the audacity to point out (My foreman's response is enclosed as Exhibit 15 ).- My foreman, in turn, chewed me out for bringing up the problems in the first place, and accused me of.being an " agitator". Following this, I was " exiled" to the temporary administration builcing to write procedures, as I have previously described. At the time, I was not unhappy with this decision, since I considered it an opportunity to correct some of the procedure errors acout wnich I was eeneerned. It became apparent, however, that I would not be allowed to a,etually correct anything.
As I mentioned, my "of fice" was a corner of the cof fee-breaR room, separated from 3
all the information which'I needed "to. update the procecures, and without a phone.
When I did produce revisions to several procecures, they were not even read by my supervisor for over two months. At this time it became obvious that I was merely being (s
isolated from the rest of the control technicians, and I transferred out of Coast Valleys Division and back to General Construction.
f Unfortunately, when I returned to General Construction as a
' ~-
Field Engineer, supervisory personnel had changed, and"there was no longer that commitment to quality and professionalism whien I had so much admired during my previous employment there. It was during this period which I discovered the major problems with the RHR system. To my dismay, I found that my new supervisor made repeated attempts to downplay my concerns and discourage me from documenting these problems, as I have.cescribed.in th'e chronology of events. I was reprimanded for noti fying Coast. Valleys Divisi,on of a problem with the RHR system and one of their Emergency Op' era-ting *Drocedures, and threatened with " losing my Job" if I talked 1
to the NRC about it. The forms I filed documenting errors in the FSAR were voided with no reason given, and the Design Change Requests I filed were simply ignored. Eventually I was sent on "Special Assignment" to the Sierra Nevada mountains, anc quit PG&E socn thereafter.
In summary, my experience with the PGAE management at Diablo Canyon after 1977 was consistently one of intimidation, retalie-
, t ion, records destruction, and inadequate corrective action to deliberately circumvent problems instead of directly addressing i
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a I have reason to believe that the attitude toward solving problems at Diablo Canyon has not changed for the better, or 4
pernaps deteriorated. Recent_ statements mace by pG4E during my f
cont'inued attempts to force resolution of these problems -
t indicat~e a sloppy, "do as little as possible" attitude towaM quality. Recent discussions with PG&E employees still at Diablo Canyon have convinced me that retaliation is still taken against 9
those who " find too many problems" by assigning the " trouble-makers",*to distasteful or boring Jobs, or placing unfavorable performjnce reviews in their personnel files.
As a result, problem reports on design or procedure problems are rarely initiated.
Robert A. Heinlein, in his recent book " Friday" made an observation that pG4E would do well to ponder. He said " Brainpower is the searcest cornreodity and the only one of real value.
Any human organi=ation can be rendered useless, irnpot ent, a danger to itself, by selectively removing its best minds while carefully leaving the st upid ones,in place". I feel that pG&E's policies are destined to remove many of the best minds from within-its organi=ation - those with the imagination and foresight to. dis-cern potential problems where their co-workers did not, and with the gumption to speak out.
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In my opinion, this situation can c'nly get worse in the
~
future, given the historical attituce of the pG4E rnanagement towarc dissent, and the perpetuation of these attituces by promo-ting to positions of responsibility only,those who "go along" and
" don' t rock the boat". With the new weapon pG4E now possesses in f
the " Behavioral Reliacility program" more direct methods will
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surely be employed to weed out whistleblowers.-It is no longer merely irritating to the enanagernent - for an employee to show a
" critical attitude toward the company", or to "not follow the appropriate chain of corninand". It is now considered a,stgn of
" emotional instability", punishable by involuntary psychiatrie examination and loss of-job.
I do not believe that the management of Nuclear powed' Opera-tions at Diablo Canyon possess the necessary character and compe-t tence to operate a nuclear power plant.
- 7 I have read the above 27 pages of affidavit and it i's truE,'
accurate and complete to the best of my knowledge and belief.
JOHN H.
COOPER Subscsi. bed and Swan t.o befosa me t.ALs t_ ant.y t. Aid cyanuasy,198ti
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EXHIBIT ~17b y
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SER' R EB U TT A L. ' 1 - 1 GD -84 On January 7, 1984, I received a copy of NUREC- 0675, Supplement No. 21 - the Safety Evaluation Report relatec to the operation of Diablo Canyon Nuclear Power Plant,-Units 1 and 2.
This document contains the official NRC responses to the
" allegations" (NRC's terminology) which I made to Mr. Eugene Powers of the NRC Of fice of Inspection and Enforcement on August 20, 1981, almost two and one-half years ago.
One would think,'given the amount of' time available to the NRC to evaluate myceoncerns, that their engineers would have done a thorough job of evaluating them. From reading the appropriate.
portions of this supplement (Allegations #37-45), however, it-becomes apparent that/this document-has been hastily prepared during the last few ' weeks in an attempt to mollify the growing public concern over the large number of outstanding problems at Diablo Canyon. The nummer of typographical errors contained in 1
this document gives some indication of the amount of time spent in it's preparation. Eut more serious, to me.
is the lac < of res=onsiveness to my original questions, anc the lack of Justifi-cations supplied by the NRC for for what I consicer to se inace-quate answers. NRC personnel, in their " responses" to my concerns, continue to ignore the basic facts of my case, the operating experiences at Diablo Canyon and at other.-nuclear power i
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plants, and they continue to insist that, Code of Federal
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Regulations notwithstanding, repeated malfunctions in a safety-related system are not considered to be a "significant safety
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concern" unless the system malfunctions when called upon to actually perform it's safety function during an-accident. I am convinced - and I think most thinking individuals would agree -
that the time to correct problems with a safety-related system is kgf2CE it is actually needed to prevent an accicent or safely _
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+
shut down the plant.
NRC representatives have told me that if tne malfunctions i
about which I am concerned had happened when there was fuel in the reactor, or when the system was called upon to function, tnen they would be considered a significant safety. concern. Since, however, the system failed (twice) - before f uel was - loaded, there was no threat to the health and safety to the public, therefore no safety problem. This philosophy is legally, morally, anc logically bankrupt, and is akin to saying that a high-speed automobile bearing down on a pedestrian is no threat to his health and safety until it actually hits him.
The following paragraphs contain my analyses-of, rebuttals to, and comments on the NRC " responses" (or'actually, lack of
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responses) to my " allegations".
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ALLEGATION No. 37 "The solid state protection system (3505)
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relays that initiate closure of'RHR letdown isolation valves s'
.8701 and 8702 perform no safety function, reduce'tne relia-bility of the RHR system, and cause a potential for RHR. pump 4
damage. Therefore, these relays should be removed."
The NRC response to this concern is very disturbing,'since it displays a total lack of understanding of how this system.
i functions. To me, this is especially disturbing since this response is from the NRC engineering staff in Washington, DC -
j the supposed " experts" on nuclear power. plants.
In the first place, the SSPS gsg3 est initiate the closure of valves 8701 and 8702. The automatic closure of these valves is initiated by the changing of state of a "comparator" module l'
locat ed in the protection and Control Racks in the cable l
spreading room (directly below the control room). From there, the
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eircuitry passes througn four input. relays in the SSPS (located l
in a separate room, adjacent to the control room) anc then bacu 3
down to the cable spreading room to the Auxiliary Safeguares j
Racks, and then to the motor control centers. for the valves. The SSPS AD D2 "f% ampli fies or changes the signals, or performs any
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logic function with them - the signals merely pass through tne 4
relays in the SSPS. Why PG4E-continues to insist enat "The solic i
state protection system-completes the logic function anc
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generates a larger output signal (amps.) which in turn actuates i
relays in the auxiliary logic cabinet" and the NRC cont inue to insist that "This automatic isolation function Eis: performec by
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the Westinghouse designed SSPS" is truly amazing.-Certainly both
(
j of these organizations have access to the circuit diagrams for this system and the expertise to understane them. One magnt i
almost conclude that this is a deliberate attempt to make tnese circuits appear to be a part of the engineered safety features of the SSPS when they really ' aren' t.
4 Secondly, both PG4E and the NRC continue to insist that the r
l RHR system has a temperature / pressure interlock system to automatically close valves 8701 and 8702 if the temperkt ure or pressure in the reactor coolant system exceed precetermined values, thus preventing an " inter-system LOCA". Both cG3E anc the l
NRC are aware that the Diabic Canyon Technical Specifications require that the power be removed from the actuators for these valves during the period when the automatic closing action woulo be desired. It is a mystery to me how both organizations continue to make this clatw when they know that the power'is removed from the valve actuators, preventing them from automatically closing; unless, again, there is a deliberate attempt to mislead the public.
4 The NRC asserts that " diverse indications and alarms are i
pecvided in the control room (including a RHR system low flow
]
alarm to be installed during the first refueling outage) to allow l
the operator (s) to assess RHR system status and to alert Ahem to 1
potentiel system degradation." George Orwell.would be proud of' the author of this sentence; it is such a fine example of "new-l speak". In the same sentesce'" the presence of an almem eis- "
claimed, and the schedule for installing it is given, perhaps the y
NRC would like to explain how it was possible, with all these i
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" diverse indications an'd alarms" for an RHR pump to be run
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h without suction, unnoticed by the operators, for 933 hgyr, until'
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it was damaged, as happened at Diablo Canyon last October..
(The pump had to be replaced. ) -
Lastly, the NRC states that "this allegation does not involve considerations that question plant readiness for power ascension testing or full power operation", and yet goes on to.
}-
say, several pages later, that "during the first cycle of opera-j tion, plants operate more frequently on the RHR system, testing and training requirements for a new plant. Thus, the period of vulnerability to a spurious RHR suction MOV closure may be greater than in subsequent cycles." Apparently, the NRC sees no i-contradiction in these two statements.
l ALLEGATION No. 38 pG8E is ignoring evidence that the spurious closure of a motor operated valve is not " impossible".
1 i
The NRC's position, here, was very well-timed. They state "The staff has examined in depth the licensee's.act ions in response to an event involving the spurious initiation of RHa motor operated valve closure as well as the concerns expressed by the alleger regarding the potential for such event sic 3, and concluded that timely evaluation and corrective measurec were taken to preclude repetition of such conditions." In the few weeks since this statement was mace, *D21U2r "s;urious initiatier of RHR suction valve closure" at Diablo Canyon caused camage to I
an RHR pump. This is twige now that the NRC has proclaimed that
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I the problems with this system have been " resolved" after making i
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only paper changes. How many more times will the Diablo Canyon j
RHR pumps have to suffer damage before the people in char e out there realize that the problem is inherent in the circuitry, not l
the procedures? As of today, pGAE and the NRC co'ntinue to espouse
}
the same philosophy which has in the past led to pump damage and loss of decay heat removal capability at Diablo Canyon anc at i
many other plants around the country.
f Over three years ago, I gave pG3E copies of.16 Licensee Event Reports documenting cases of " spurious initiation of RHP suction valve closure" in various plants around the country.
I' m i
sure that many more cases must have occurred since that time besides the latest one at Diablo. But pGSE still continues to
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contend that "A failure, such as the spurious closure of a motor operated valve...has not been considered credible." (FSAR, page 3.1-3), and that " Westinghouse does not consider spurious operation of electrically controlled valves as a credible single active failure" (FSAR, page 15.4-8), and that "The probacility of i
any spurious valve closure is therefore 2.54 x 10 to the minus 1
8th power per valve-hour. " (FSAR, page 6.3-34a). In the face of i
the overwhelming evidence that spurious ' valve closures happen quite regularly, I can only interpret the failure of pG82 and the j
NRC to recognize this evidence (and act upon it) as yet another attempt to mislead the public as to the safety of the RHR ' system.
The NRC states that "It does appear'that the l i,censee.i s,
(" s giving proper attention to the spurious closure of the valves in t
i question". I would like to point out that mere sligD112p 'is only 1,
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.he first step toward correcting the deficiencies in the RMR
-(1 system. AEt12D is necessary.to.actually correct the problem anc J'
prevent it from happening again. How many more times will an RHR pump have.to be damaged at Diablo Canyon before tne necessary action - wiring the interlock circuitry directly from the "hagan racks" to the " Auxiliary Safeguard Racks", without it. passing through the SSpS,-and adding the low RHR flow alarm - will be taken?
1 ALLEGATION No. 33 "There is no control room annunciation provicec to alert the operator (s) when the RHR letdown line has been isolated during Modes 4, 5,
and 6 (hot shutdown, cole shut-down, and refueling respectively).
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PGAE was instructed, on April 2, 1981, to install a low RHR flow alarm in the control room, but was allowed to wait until after the first refueling outage to do so. Apparently GG&E intends to wait until the last permissable minute to install this alarm, even though its presence would have prevented the damage to the RHR pump wnich occurred two months ago. This attituce coes not boost my confidence in PG3E's commitment to the safe, i
reliable operation of Diablo Canyon.
~
The NRC intends to achere - to-their original' schedule for the
.1 installation of this alarm, even in the face of this second ancident, stating: "The staff has concluced that the existing control room indications and procedures are sufficient to assure
- adequate decay heat removal in the interim."
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ALLEGATION No. 40 "The question raised was with regard to whether or not the single RHR pump suction line from.the RCS hot leg meets safety related standarcs. The newer AWRs are cesigned with redundant RHR pump suction lines from the RCS hot legs."
Ag a a n, the NRC missed the boat on this one. My contention j '
was that this system should be remnalyzed in the light of the TMI accident. PGAE claims that the single RHR suction line is not i
i safety related and is only used during the normal coolcown of the plant. I disagree. At TM1, their RHR system was used to mitigate the consequences of the now famous accident there. I propose that this portion of the Diablo Canyon RHR system is inadequate, since a single failure in this line would prevent decay heat removal 2
i via this system. I offer these facts as evidences
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' 'i.~ hhe proven unreliability of the suction valves in this e'
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j line, both at Diablo Canyon and at other plants.
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That 199s IffflY Cfifigg suction lines from the RCS to the
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RHR are provided in CE, B&W, and the newer Westinghouse j
designs. Why would these extra' lines be provided i f t hey 2
weren' t needed?
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In Regulatory Guide 1.133, the NRC staff states that the
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Residual Heat Removal System 13 required to be redundant, and t
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' withstand gay _ type of an accident, not just a large break t
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Apparently, the NRC is aware of the deficiencies in this and i
perhaps other areas in the design of Diablo Canyon, since they t
consider it a " Class a plant". They go on to state that "A single RHR suction line from the RCS hot leg is considered acceptable for a class a plant as long as a single failure could be 1
corrected by manual actions inside or outside of containment, or the plant could be returned to hot standby until manual actions (or repairs).are accomplished." I would like to pose a question at this _ point. iT4cw, under Cadi_ological conoitions which wou l d --* ;
i prevent human entry into th'e containment - (such were found inl the containment at TMI), could valves 8701 or 8701 be opened j
manually or repaired?
- Both of these valves are located inside the containment structures at Dtablo Canyon.
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ALLEGATION No. 41 The power source of certain relays is not snown 4
on certain drawings and this caused an operational problem, the failure (closure of RHR isolation valves)"
Although the' contacts of the SSpS input relays are shown on the electrical schematic of this system, the power source for tne j
relay coils is not shown 20 20% REEC2Vid 21301 GCAW1Dg. The NRC continues to state that the Septemmer 1981 incicent was due to a
" lack of pre-plarning" implying that if the technician i
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responsible for the incident (and his foreman) had just done
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It seems to me, that i f the informat ion is not on any drawings, then no amount of pre planning will help, and errors are bound to j
be made. The NRC takes great pride that a technician-drawn
- I
" composite drawing" of this system has been put together. I put together a similar drawing back in May of -1981, and gave'it to j
the instrument foreman, but that obviously didn' t prevent the 4
incident 3 months later. Hand drawn, or " bootleg" drawings can j
get lost, or not be distributed to the persons who need the information. That is the whole purpose of the Drawling Control j
System - to make sure that Aggyr31gs yg 12 delf 1Df2rTall2D is l
readily available and accessable to those who need it. Not only i
is this idea Just good common sense, but it is part of the Coce of Federal Regulations 1
i 10CFR50 Appendix 9,' III Design Control 1
Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in I
paragraph 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings,'
procedures, and i nst ruct ic os.."
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pp g '5 ALLEGATION No. 42 Licensee management was unresponstve to recom-(
mendations to prevent spurious closure of the isolation valves on the residual heat removal (RHR) system.
Closure of the valves disables operation of the RHR system for decay heat removal.
The chronology of events in this case speaks for itself. At no time did pG8E take any effective action to resolve any of my concerns without the intervention of the NRC. Most of the the original problems which I brought first to PG&E's, then the NRC's attention remain uncorrected even today, after three years of unsuccessful attempts on my part.
In the NRC assessment of the safety significance of this problem, an attempt is being made to give the impress 1on that PG&E, Westinghouse, and the NRC have all been actively working for the past three years solve these problems. I find this very hard to believe, since no evidence of any kind has been produced to support this position.
As far as I' m concerned, the last consideration that this problem received was in November of 1991, wnen the Diablo Canyon Onsite Review Group cecided to take no action to correct the problem.
ALLEGATION No. 43 The loss of the residual heat removal (RHR) system on 9/39/81 due to unplanned closure of the RHR iso-lation valves was an event whien should have been reportec to the NRC in accordance with 10CFR50.72. The l icensee' s failure
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to make such a report was in violation of NRC regulations.
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10CFR50.72 Notification of significant events (a) Each licensee of a nuclear power reactor licensed uncer paragraph 50.21 or 50.22 of this part shall notify the NRC Operations Center as soon as possible and in all cases within one hour by telephone of the occurrence of any of the following significant events and shall identify that event as being reported pursuant to this section:
"(6) personnel error or procedural inadequacy which, during normal operations, anticipated operational occurrences, or accident conditions, prevents or ggitig gegver1, by itself, the fulfillment of the safety function of those structures, systems, and components important to safety that are needed t o... ( i) remove residual heat following reactor shutcown..."
(my empnasis)
The NRC claims that "The loss of residual heat removal capacity during a time when significant fission product decay heat is present in the core would have safety significance. In tmis particular instance, fuel had not been loaded into the Dieb1'o Canyon Unit
- 1. Therefore, no fisuson procuct cecay heat was present and loss of RHR capab.ility had no actual safety (I$
significance." p ain, the NRC doesn' t perceive any safety'p'roblem g
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until the speeding automobile actually hits the pedestrian.
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kk~ j ALLEGATION No. 44 The licensee failed to properly process a j
Nuclear plant Problem Report.
1 In my original complaint I stated that the Nuclear plant problem Report documenting the first loss of RHR suction at l
Diablo Canyon was " signed off as complete without any plant i
management review... classified as ' non-report ab l e' and without j
any follow-up action such as an RHR pump inspection or investigation into the cause of the event." The NRC admits that the above concerns are true (although two months after the event, another problem report was initiated to perform a pump test), they just believe that this is an acceptable way to run a power plant. In my discussions with Mr. Jess Cruse of the NRC, 2
who interviewed the " principles" in the handling of this problem report, he stated that "no-one denies that it could have happened 3
Just the way you said it Cdid3, and I sort of concluded most l
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likely it did happen Cthat way3". What action did Mr. Cruse take?
i None. Mr. Cruse also stated that this was not ' reportable' i
because there was no fuel in the reactor at the time. Again, it 1
would seem to me that the NRC would wish to know of problems i
ggfgr3 there was fuel in the core, but this is apparently not so.
I As for the analysis of the problem to prevent reoccurrance, this has not been done by AGAE even to this day, as evicenced by the recent (Nov. 83) replay of the September 1991 incident. PGS.E and j
the NRC both claim that " strict procedural-controls" are adequate to prevent reoccurrance, although this method has been proved inggggggig twice before.
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ALLEGATION No. 45 Section 5.5 of the Diablo Canyon FSA3 describes j
gg the autoclosure interlock for the RHR suction line isolation i
valves (8701 and 8702). Sect ion 3. 4. 3. 3. a o f the Diablo
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Canyon Technical Specifications requires power to be removec from these isolation valve operation during modes 4 (Hot N
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shutdown when RCS cold leg temperature is less that 323 i
i degrees F), 5 (cold shutdown), and 6 (- fueling). This re-(
quirement defeats the function of autoc.5 'are interlock for the valves.
In their lengthy analysis of this simple allegation, the NRC l
admits that removing the power from these valve operators defeats i
the autoclosure interlock to the RHR suction valves as described
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in the FSAR. I contend that either the FSAR should be corrected so it accurately describes the RHR system at Diablo Canyon, or the RHR system should be operated in conformance with the FSAR.
4 i
The Code of Federal, Regulations is clear in both cases: The NRC j
must be notified if the plant design does not conform to the criteria and bases in the FSAR, and the FSAR must be kept up to 1
date.
1 10CFR50. 55 Conditions of construct ion permit s l
fe~ N (e)(1) I f the permit 15 for construction of a nuc1'e'aF'pswer i
plant, the holder of the permit shall notify the Comm'ission i
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of each deficiency,found in design and construction, whien, A
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I l-adversely the safety of operations of the nuclear power-
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plant at Any time throughout the expected lifetime of the plant, and which represents:
f (ii) A significant deficiency in final design as approved and released for construction sygd ibAl ibt gg313D g221 D2%
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10CFR50.71 Maintenance of records, making of reports J
j (e) each person licensed to operate a nuclear power reactor l
l pursuant to the provisions of paragrapn 50.21 or 50.23 of this part 3bg11 ggggig ggriggiggily, as proviced in paragraphs (e)(3) and (4) if this section, kbg final ggfsky
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A0f1X512 C222C1 JEgA31 originally submitted as part of the i
l application for the operating license, to assure that the information included in the FSAR contains the latest material developed. This submittal shall contain all the changes e
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necessary to reflect information and analyses submitted to the Commission by the licensee or prepared-by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last updated FSAR.
j The updated FSAR shall be revised to include the effects of J
all changes made in the facility or procedures as described I'
in the FSARg all safety evaluations performed by the licensee i
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either in support of conclusions that changes did not involve en unreviewed safety question; and all analyses of new safety j
issues performed by or on behalf of the licensee at j
j Commission request. The updated information shall be t
f appropriately located within the FSAR.-
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(3) (1) A revision of the original FSAR containing those original pages that are still applicable plus anew i
replacement pages shall be filed within 24 months of either July 22, 1980, or the date of issuance of the operating license, whichever is later, and shall bring the FSAR up to cate as of a maximum of 6 months prior to the date of filing i
the revision.
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The NRC, in it's analysis, curiously avoids any mentien of the above two regulations, but goes on to say that operating the j
plant with the power removed from these actuators is a violation
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of their Branch Technical position RSS 5-1, position 9.1.C.
They l
also state that "There have been many occasions of spurious RHR i
suction valve closures on Csic3 operating plants. This has i
resulted in not only a loss of decay heat removal, but also an i
overpressure event due to the loss of the letdown flowpath. " They l
l continue that "During the first eyele of operation, plants j
operate more frequently on the RHR system as a result of l
- fb maintenance, testing and training requirements for a new-plant.
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Thus, the period of vulnerability to a spurious RHR suct i'on MOV j
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closure may be greater than in subsequent cycles. They basically {
i agree with everything I've been saying for all these years - and I
their " response" is to do D21h1Dg to cor rect the situation until the first refueling outage. I never cease to be amazed at the workings of the bureaucratic mind.
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GOVEF.NMEN'T ACCOJNTADIUTY PROJECT
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.T vo3-g - D C 2JOOQ (2C2, 2 R G '22 January 9,1954 HAND-DELIVERED 9 January 1994 FREEDOM OF INFORMATION ACT REQUEST
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Director
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Office of Administration
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U.S. Nuclear Regulatory Commission Washington, D.C.
20555 To Whom It May Concern:
Pursuant to the Freedom of Information Act (FOI A), 5 U.S.C. 5552, the Government Accountability Project (GAP) of the Institute for Policy Studies, requests copies of any and all agency records and information, including but not limited to notes, letters, memoranda, craf ts, minutes, diaries, logs, calendars, tapes, transcripts,
summaries, interview re;;rts, procedures, instructions, engineering analyses, draw-ings, files, graphs, : harts, maps,. photographs, agreements, handwritten notes, studies, data sheets, n:tet oks, books, telephone messages, comoutations, voice recordin;s, and any other data compilations, inter,in and/or f.inal r m ets, status reports, and any and all other records relevant to and/or generated in connection to_the_cFBf'iFLof the,0faiilo Canyon
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with the SafAty Evalua. tion Report rel3 tad o
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Nuclear Power Plfnt,' Units 1 and 2, hUREG-0575 Supplement No. 2.)J(December,1933)
@k"$ This includes, but is not limited 6." anfanfal'1 NRR regu1 tions, 3
"-mgalanil ins, and judgments used to evaluate any of the allegations referred to in SSER 21. We request that each responsive document be identified by the allega-
"titrrrMi5e7('s) to which it may relate.
If any of the material covered by this request has been_ destroyed.and/or removed, please provide all surnunding documentation, including but not limited to a des-cription of the action (s) taken,7eETint~date(s), and justification (s) for the action (s).
GAP requests that fees be waived, because " finding information can be considered as primarily benefitting the general public," 5 U.S.C. 5552(a)(4)( A).
The Government Accountability Project is a non-profit, non-partisan public interest organization concerned with honest and open government.
Through legal representation, advice, national conferences, films, publications and public outreach, the Project promotes whistleblowers as agents of government accountabili.ty.
Through its Citizens Clinic, GAP offers assistance to local public interest and citizens groups who seek to ensure the health and safety of their communities.
The Citizens Clinic is currently assisting citizens groups in the California area concerning the Diablo Canyon Nuclear Power Plant.
For any documents or portions that you deny due to a specific FO!A exemption, please provide an index itemizing _and, describing the documents or portions of docu ents withheld.
The index should provide a detailed justificatien of your
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grounds for claimin; each exemption, explaining why each exemption is relevant to the d0c; tent er portion of the document withheld.
This inde> is required unde"
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.Vaughn v. Rosen (I), 434 F.2d E20 (D.C.Cir.1973'), cert. denied, 415 U.S. 977 (1974).
We look forward to your response to this request within ten days.
Very truly yours.
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,i r-Billie Pirner Garde Citizens Clinic Director BPG:me 1
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