ML19350E307
| ML19350E307 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 06/09/1981 |
| From: | Latham S SHOREHAM OPPONENTS COALITION, TWOMEY, LATHAM & SHEA |
| To: | LONG ISLAND LIGHTING CO. |
| Shared Package | |
| ML19350E302 | List: |
| References | |
| NUDOCS 8106170286 | |
| Download: ML19350E307 (86) | |
Text
i .
f UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Mater of )
)
LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322
)
(Shoreham Nuclear Power Station,)
Unit 1) )
ANSWERS OF SHOREHAM OPPONENTS COALITION (SOC) TO APPLICANT'S INTERROGATORIES DATED MAY 21, 1981 Beginning on page A-1 attached, SOC herewith submits Answers to the Interrogatories submitted by Applicant, dated May 21, 1981. The responses have been captioned in a fashion identical to that utilized by the Applicant (by letter 'nd SOC Contention number).
. The responses do not contain a letter "E" (SOC Contention 7a(ii)) since no interrogatories on that Contention were submitted by Applicant. For the reasons stated on page
-- "M
, .1.",- SOC objects to the- interrogatories submitted on Contention 19.
The Answers to the Interrogatories filed on each SOC Contention begin with a preamble or narrative introduction, followed by supplemental responses to the specific interrogatories. Where appropriate, the preambles are cross-referenced in the supplemental 810 6170 2J($
3
.I Interrogatories A.11 and A.12). SOC reserves i ts right i
to' object to providing further responses to these or i
other interrogatoties beyond the answers presently submitted on thu. grounds of privilege, undue burden or other grounds should Applicant seek to colipel additional answers in accordance with the Cornission's Rules of Practice.
i s~
- e
.w s
g m e ee e 6*
O 8
s y e
--en ,y g-g,-w.-new, ,,ew q . - - - - . - , , l, 7 , 7,w wa,-,,,,,,--5,n-,---w,p--,-ew,wrew- --e,my--ym-rw-. -, e,w,--- ,w-e,r-e- ~ew--.--gn
i .
A. S_0C CONTENTION 1 The Shoreham-specific emetgency planning criteria have not yet been developed by LILCO, by local authorities including Suffolk County, or by state authorities. Furthermore , neither the NRC or FEMA review of emergency preparedness has been con-ducted at this time. Thus, the Shoreham-specific distance cri-teria which will be utilized as the basis of the co=bined appli-cant, state, and local plans has not yet been documented and, as a result, SOC answers to this interrogatory are premature.
However, SOC will provide responses in so far as practical on the basis that the applicant, state, and local authorities will develop the emergency planning zones of about ten and fifty miles for plume and ingestion as provided as guidance in NUREG-0654.*
First, SOC contends char. neither the Staff or LILCo have conducted a Shoreham site-specific accident consequence assess-ment utilizing the NRC's CRAC Code (from WASH-1400) or equiva-lent. Such an analysis based on site-specific core inventory, release fractions, release time estimates, shielding factors, health treatment facilities, demography, topography, meteorology, land characteristics, access routes, and local jurisdictional boundaries is necessary to develop the appropriate plume ex-posure and ingestion Emergency Planning Zones (EPZ). As dis-cussed by the Staff in the Shoreham SER** LILCO has not yet sub-mitted emergency plans required by the new regulations which were published in the Federal Register on August 19, 1980, and became effective on November 3, 1980. The regulations contain a revised Appendix E to 10 CFR Part 50, " Emergency Planning and Preparedness for Production and Utilizat, ion Facilities," which establishes minimum requirements for an acceptable state of on-site e=ergency preparedness, and a new Section 50.47, " Emergency Plans," which specifies standards which must be met for both on-site and'offsite emergency response. This latter section incor-porates the joint NRC/ FEMA standards for use in evaluating state and local radiological emergency plans and preparedness. The Staff also acknowledges NRC and FEMA have agreed that FEMA will make a finding and determination as to the adequacy of state and local government emergency response plans. NRC will determine
- " NUREG-0654 (FEMA-REP-1), Rev. 1, Criteria for Preparatice and Evaluation of Radiological Emergency Response Plans and Preparecness in Support of Nuclear Power Plants, page 17, for example.
- NUREG-0420, Shoreham SER, page 13-1.
A-1 e
, - . --w- - - - - n- - > - - - - - ~ , - . - , - , - - , , , - - , - - , , - - , - - , - , , - - - , ,-e-. ,
L f emergency response the adequacy of state and local governmentto the standards listed in Section 50.4 plans with respect 50, the requirements of Appandix E to"10 CFR of 10 CFR Part Criteria Part 50, and the guidance contained in NUK'IG-0654, for Preparation and Evaluation of Radiolog. of calNuclear Emergency Power Re'-
sponse Plans and Preparedness in SupportAfter the above determinations Plants," dated November, 1980.the staff will make a finding in the licensing by NRC and FEMA, process as to the overall and integrated state of prepa Such a review will be published, 'at some future time by the Staff in a Supplement to the Shoreham SER.
The President's Commission on the Accident at Three in Mile Island (Kemeny Commission)* noted that a central concept the then current siting policy of the NRC is that (LPZ), an area reactors should be located in a " low population zone"in which How-around the plant appropriate pro be taken for the residents in the event of an accident.
ever, Kemeny concluded that the concept is implemented in a To determine the strange, unnatural, and round-about manner.the utility calculates the amount o size of the LPZ, in a very serious hypothetical ac-released into the containment Using geo-cident but assumes no failure of the containment.the utility then calculates graphical and meteorological data, rems or more to the whole body, during the entire co The 25,000-millirem standard accident. This area is the LPZ.
is an extremely large dose, many times more serious than that re-ceived by any individual during the entire TMI accident.
The Kemeny Commission believed that the LPZ approach has serious shortcomings. First, by which its size is determined, the LPZs for many nuclear power plants are relatively small areas, Second, two if an accident miles inasthe serious case asof TMI the as well as Shoreham. it is evi-one dent used to calculate the LPZ were actually Third, the toTMI occur,that acci- man er, but still massive doses of radiation.the LPZ has little distances relevance to the dent shows that the public -- the NRC itself was considering evacuation as far as 20 miles, even though the accident Kemenywas therefore far less serious concluded
__,than,those postulated during siting.
al, Report of the President's Commission on KeLeny, John, etat Three Mile Island, pages 16-17.
the Accident A-2
,.--...----<.,-e,n ~ - - , , , - - - - , . .
3-- - - - - - - - - -
I that the entire concept is flawed and recommended that the LPZ concept be abandoned in siting and in emergency planning.
As an alternative, Kemeny proposed and SOC concurs that a var-iety of possible accidents should be considered during siting, particularly " smaller" For accidents each such which accident, have one a should higher calculate probabili-ty of occurring.
l on a site-specific basis probable levels of radiation releases at a variety of distances to decide the Such kindsprotective of protective ac-actions tion that are necessary and feasible.
may range from evacuation of an grea around the plant, to the distribution of potassium iodide to protect the thyroid gland from radioactive iodine, to a simple instruction to people sev-eral miles from the plant to stay indoors for a specified per-iod of time. Only such a site-specific analysis can predict the true consequences of a radiological accident.
Another major review group formed by the NRC af ter the TMI accident, the Commission's Special Inquiry Group, directed by Mitchell Rogovin, has also suggested that ten miles is For not a
an appropriate cutoff distance for emergency planning.*
further discussion of some of the deficiencies in the logic which went into the choice of a ten mile cutoff distance, see
~
T. Lombardo, T. Perry, " Mitigating the Effects of a Nuclear Ac-cident," Spectrum, July, 1980, page 30.
Neither the Shoreham safety analysis nor the NRC's NEPA review for Shoreham considered Class 9 (very large) accidents.
As a result of the Class 9 accident at TMI-2 and the NRC's Risk Assessment Review Group's conclusion (NUREG/CR-0400) that estimates of the absolute probabilities of accidents inthe WASH-NRC has 1400 (NRC's Reactor Safety Study) are not reliable, no theoretical or practical basis for excluding theShoreham. safety and In environmental assessment of by Class the NRC 9 r.ccidents (Federal at Register, June an interim policy statement 13, 1980), the NRC has decided to approach this problem on a plant-by-plant basis considering the potential However, no consequences Shoreham assess- for large accidents at specific sites. This is ment prior to fuel loading is pr'esently contemplated.
a serious omission.
In the course of preparing WASH-1400, major One ofinsights into the significant the nature of reactor risk were discerned.
- -. - ~ . ..
A Report to
- Rogovin, Mitchell, et al, Three Mile Island:
the Co1=nissioners and to the Public, page 133.
A-3 s
. - , . . . - - _ _ . - . , . - , . - - , . - , , - . . _ y , ,r. , ,_,.._.-
a a i
I conclusions of the study is that the risk to the public from ;
nuclear power reactors arises primarily from core celtdown ac-cidents (Class 9 accidents). In a subsecuent program under-l taken by Brookhaven National Laboratory (3NL), Science Appli-cations, Inc. (SAI), and Battelle Columbus Laboratories (BCL),
the risk contribution of Class 3 to 8 accidents was examined.*
Class 3 to 8 accidents,** incidents that may be expected to occur over the lifetime of a plant and which do not exceed the design bases of the plant, are analyzed in safety analysis re-ports and environmental reports. ,As illustrated in Table A-1, TABLE A-1 COMPARISON OF EXPECTED RELEASES Risk, weighted Curies / reactor-year ***
Normal operational release 1.7 Class 3-8 accidents Actuarial data 0.05 Extrapolated 0.05-0.2 Total 0.1-0.3 Class 9 accidents 540
- R.E. Hall, et al, A Risk Assessment of a Pressurized-Water Reactor for Class 3-8 Accidents, NRC Report NUREG / CR-0603 ,
Brookhaven National Laboratory, NTIS, October, 1979.
- U.S. Atcmic. Energy Commission, Consideration of Accidents in Impleuantation of the National Environmental Policy Act of 1979--Discussion of Accidents in Applicant's Environmental Repqrts: Assumptions, Annex to Appendix D, Title 10, Code of Federal Regulations, Part 50, Federal Register, 36(231):
22851-22854 (Dec. 1, 1971).
- The units of weighted curies used in this table involve a sum-mation of all the important radionuclides weighted by a factor .
to account for the relative effect of that radionuclide to iodine-131 in producing latent fatalities (including fatali-ties from malignant thyroid tumors). Although these weight-ing factors would actually be dependent on the conditions of a specific accident, they provide a convenient approximate means of measuring accident consequences.
A-4
- - ,$- v - * * - . , - - - ~--,,--,,n
..y. , - . - - * ,--v,-,-----
.----,m-~,.-n-wp,w --,,-,,wa- .,n-- --, ..-g, ,,--
i ,
the results of this study indicate that, on the basis of ex-pected value, Class 9 accidents are predicted to have a much greater impact on human health and the environnent than either Class 3 to 5 accidents or normal releases, despite the low prob-ability of core meltdown. While the plant analyted was a PWR, we believe the conclusions are also applicable to the Shoreham BWR.
Using the same weighting factors for the different radio-nuclides as in Table A-1, the totpl radioactivity potentially available for release from the containnent atmosphere at one hour in a core meltdown accident is 108 weighted Ci. The major fraction of the risk calculated in WASH-1400 originated from meltdown accidents in which the containment failed during or shortly after core meltdown, before significant deposition (greater than an order of magnitude reduction) could take place.
In contrast, the amount of radioactivity that would be released to the environment from Class 3 to 8 accident (the Shoreham de-sign basis accidents) would be very small, typically less than 1 weighted Ci. Although these accidents have been predicted to occur with frequency up to as much as four orders of magnitude greater than core meltdown accidents involving large atmos-phere releases of radioactivity, the difference in consequences still would be approximately seven orders of magnitude. Thus the risk, measured by the product of probability and conse-quence, for Class 9 accidents is predicted to be much greater.
Therefore, emergency planning measures must be developed to mitigate the consequences of these releases.
It is also relevant and important to note that the WASH-1400
. Evacuation Model presumes more extensive evacuation Thethan introductory the ten miles presently assumed to be proposed by LILCO.
mention of evacuation in WASH-1400 states: "In the case of a po-tentially serious accidental release, it is assumed that people living within about 25 miles of the plant, and located in the direction of the wind, would be evacuated...."*
Other references to evacuation are in WASH-1400, Appendix VI, where it is explained that credit is taken for evacuation in all directions to five miles and in a downwind 450 section to 25 miles in order to concentrate evacuation facilities where they l
l will do the most good.** The 25-mile limitation precluded the WASH-1400 authors considering evaucation from any U.S. city i
I
- WASH-1400, Main Report, page 51.
- WASH-1400, Appendix VI . page 11-4.
i A-5 1
1
,- , - - - . - - - - . -n. -, - - , . , , , _ , , - . - , - - , , , , - , , . , , . - - -
,e,--,---------- , , . - - .
larger than Cincinnati. (Approximate population of 450,000 in 1970.) However, the WASH-1400 authors did assert that large U.S. cities, such as New Yerk, Boston, Philadelphia, Chicago, and Los Angeles, with reactors at their edges, cannot reason-ably be evacuated in less than one week.*
Further, unlike the Shoreham site-specific condicons, the WASH-1400 authors assumed in the evacuation model that the popu-lation movement is always radially outward from the reactor, i.e., there is no crosswind movep.ent. During the evacuacion, the population is assumed to be unshielded from exposure to air-borne ::adioactive material both externally and through inhala-tion and to be shielded from exposure to ground contamination due to surface roughness and the automobile. If the evacuating population is overtaken by the cloud of radioactive material, it is assumed that people will have moved outside of the contami-nated area within a four-hour period. Beyond 25 miles, the people are assumed to be relocated within seven days if th.e chronic dose due to exposure to ground deposited radionuclides exceeds a specified value. However, if the dose accumulated within the .irst seven days due to exposure to contaminated.
ground exceeds 200 rads, then the people are assumed to be re-located within one day.
Three categories of potential health effects are calculat-ed in WASH-1400: early and continuing somatic effects, late somatic effects (cancers), and genetic effects. Since early-and continuing somatic effects are usually observed after large, acute doses of radiation (e.g., whole-body doses of 100 rads),
they would be limited to persons within 50 miles (not 10 miles) or so of the reactor even for the largest conceivable release.**
Conversely, late somatic and genetic effects may result from very low doses albeit with low incidence. Consequently, these effects may occur at long dist'ances from the reactor.
No figures are presented in WASH-1400 for the conditional probability of a latent death resulting from a BWR release as a function of distance. However, WASH-1400, Figure VI 13-26 shows the conditional probability for an individual of dying from latent cancer as a function of distance from a reactor given the PWR-1A and PWR-1B releases. The probability of latent cancer fatalities is relatively constant out to about 100 miles from the reactor l
!
- WASH-1400, Appendix VI, page 11-5.
- NUREG-0340, Overview of the Reactor Safety Study Consecuence Model, page 16.
A-6 l
1 -
l .
i a 1
. -~
1d ,
.. ~ ..., .
7 .
j '
~
/ .
' \_ines , arma.toi _
l a
1 4 -
t ,o ,
E ,
1 .
1 1 1 i .
,p
- 102 ,$
1 to unn .vem messier ,
FIGURE VI 13-26 Conditional probability of latent cancer death given a PWR-1A or PWR-1B release. (Approximately, absolute mortality probabilities are 10-6 per reactor year times stated ones).
Source: WASH-1400, Appendix VI, page 13-37.
.p+>
" e6 .m ome a t
A-7 f . .
l l
\ .
I beyond which it decreases rapidly. The small difference be-tween the two curves is due With to the different heat rates in the the large heat rate in the PWR-1A and PWR-1B releases.
PWR-1B release, it is less likely for an individual close to the reactor to be subject to significant doses. Therefore, the probability for a latent cancer death would be lower close to the reactor for the hot release case (PWR-1B) than for the cool A similar result would be expected for release case (PWR-1A)'.
a BWR-1 release from Shoreham. Thus, for Shoreham, the number of latent fatalities and health effects is roughly proportional to the population living between 25 and 100 miles from the re-actor and to the associated mitigation measures for this group.
Site-specific consequence analyses as recommended herein have been conducted for the California nuclear plants. The California results confirm that the 10 and 50 mile EPZ's are not appropriately conservative. Following the March 28, 1979, accident at Three Mile Island, Governor Brown formed a task force to study California's emergency preparedness for nuclear power plant accidents. One of the task. force's principle recom-mendations was that site-specific analyses be conducted of the consequences of hypothetical serious nuclear power plant acci-dents. As required by subsequent legislation, a site-specific study of this nature has been conducted by the State Office of Emergency Services (OES). Based on the results of this Study, the OES has recommended Emergency Planning Zones (EPA's) for each of the nuclear power plant sites. The rationale behind the selection of these EPZ's is presented in the following por-tion of this response.
As noted in the OES report, the selection of the ten-mile radius (for the plume exposure EPZ) by the NRC/ FEMA was based primarily on the follawing considerations:
- a. projected doses from the traditional design basis accidents would not exceed Protection Action Guide leve'Is outside the zone;
- b. projected doses from most core melt sequences would not exceed Protective Action Guide levels outside the zone;
.c. _for the worse core melt sequences, immediate life threatening doses would generally not occur cutside the zone;
- d. detailed planning within 10 miles would provide a substantial base for expansion of response efforts in the event that this proved necessary.*
- NUREG-0654, page 9.
A-8 O . .
- - _ . - - - - __--.c - , - -
---c- .v--,---r---- w . - - - - - -- - -
Based upon the site-specific results of the State's study, the OES analyzed the impact of a similar set of hypothesized accidents on the areas about each of the California sites.
EPZ's were developed upon this site-specific basis rather than the generic basis used by the NRC/ FEMA. Based upon the site-specific results, the OES concluded EPZ's should be extended The outer beyond the basic 10 mile NRC/ FEMA requirements.
boundaries of the ext-ended EPZ's presented above for the three large reactor sites in the State with ratings comparable to Shoreham vary in distance from about 18 to 35 miles from the l
reactors, as shown in Table A-2.
TABLE A-2 DOWNUIND DISTANCES FROM REACTOR SITE WHERE PROBABILITY OF EXCEEDING SPECIFIED WHOLEBODY DOSE IS 0.01 (1%) FOR CORE-MELT ACCIDENTS
Diablo Canyon 4.5 miles 20 miles 26 miles Rancho Seco 5 miles 18 miles 28 miles San Onofre 1 8.5 miles 13, miles 18 miles San Onofre 2, 3 4 miles IP miles 35 miles The decision to extend the EPZ boundaries beyond the zone dimensions required by ? 'C/ FEMA was based upon a disagreement in the application of the (u, and (d) considerations above that were used by the NRC/ FEMA in their election of the basic 10 miles radius for the EPZ. As indicated by the dose-distance relation-I ships developed during the study, the results of the State's study of the California site-specific consequences showed basic agreement
- Cunningham, Alex, Emergency Planning Zones for Serious Nuclear Power Plant Accidents, California OES, November, 1980, Table 4.1.
A-9
_ - . _ , , .- _ y . , -- _ - . - - , -
with observations (a) and (b) of the NRC/ FEMA selection con-sideration presented above. That is the State's results in-dicated there was at least a 99.9% probability that the pro-jected doses from the most probably (melt-through) accident sequences would not be expected to exceed Protective Action Guide (PAG) levels beyond a basic 10 mile EPZ boundary. In this regard, the State's results supported the first two of the NRC/ FEMA observations.
However, the OES concluded that prudence dictates that the EPZ's be extended so that advance planning can be per-formed to aid in resolving the potential problems associated with the more severe types of accidents, the penetration leakage and major containment failure classifications. The very severe accidents in these classifications make up about 10 to 20% of all the hypothesized core-meltdown accidents, even for relatively new reactor designs. It did not seem prudent to restrict planning attention to responding only to the potential for incurring immediate life threatening radia-tion doses for such severe accidents. Thus, the extended zones were selected so that the potential for incurring health impacting doses was reduced not only for early fatalities, but also for early injuries and delayed cancer effects as well.
-7.-. f A direct comparison of NRC/ FEMA generic estimates of dose-distance-probability relationships with the State's site-spe-cific results is presented in OES Figure 4-13. The results of the two different sets of CRAC calculations made by the NRC and the State show basic similarities, but also some substan-tial differences. The NRC results appear to have been smooth-ed to follow the general outlines of their CRAC results, uhile the State's dose-distance-probability data shown represent the direct, unmodified output from CRAC (for the Rancho Seco site in the particular example shown). The results presented in the figure (for both the NRC's and the State's data) represent the combined output of the codes for all accident classifications (maj or containment failure, penetration leakage, and melt-through categories). The results have been combined to reflect
.the appropriate weighting factors of the relative probabilities of receiving accidental dose: from each and all of the accident classifications.*
As the NRC observed in NUREG-0396, there is about a 30%
thance- of exceeding the (1 rem) PAG level at the basic 10
- Cunningham, Alex, page 66.
A-10 S
- "'"- ' --m . - . . _ - , _ _ , , _ __ _ _
i I
. 1.0 1
N 1 REM N
% N NN 5'RE! N N
. - . - N h.
0.1 '
9 --
s i N. . I c 9 N
N
\
S b N':50 REM \ \
l c i W w J U \
O U N. -
E < .'
- I g
$ d k
- 0.01 0 h IY \
u c -
K U l u . '
4 b 2 w .
i- .
- .' 1.)
lt M C
- I i' '
5 0.001 5 -
l J
I, 20-) REM 5 R !! 1 RE!!
l
\ i
_.o geol _
100 1000 1 10 DISTANCE (!!ILES ) .
Figure 4-13 WHOLE BODY DOSE ESTI!iATES STATE STUDY NUREG - 0396 A-ll 9
r, , , - , - , _ . , - - .-w., , . - . , . -
y--+---, , . - -
t 4
miles plume exposure EPZ boundary, according to their own da.ta .
The State's data would indicate a somewhat lower probability of about 20% at the same distance. At the same 10 miles dis- -
cance, the NRC results suggest a probability of about 10% of exceeding a 50 rem dose, and about a 3% probability of exceed-ing 200 rem. Bared upon the NRC's and the State's data, adopt-ing extended EPZ boundaries with distances from the reactor of about 18 to 35 miles would reduce the probability of early fatalities by a factor of 0.1 or more (to a probability of about 0.1% of exceeding 200 rem); the probability of early injuries (at 25 rem) would be reduced by a factor of about 0.5 to 0.25 (to a probability of about 7 to 9%); and the probability of ex-ceeding PAG doses would be reduced about a factor of from 0.5 to 0.8 (to a probability of about 15%). Thus, extending the EPZ boundaries results in a prudent reduction in the probabilf~
ties of early health effects and a substantial reduction in c'.se probability of delayed health effects (associated with 0.5 to 1 rem PAG dose limits).
Thus, based on the foregoing, SOC concludes that there is substantial precedent for the prudence of conducting a Shoreham-specific comsequence analysis to develop the appropriate emer-gency planning EPZ's. Further, we believe that such an analysis would indicate that the Plume Exposure EPZ of 10 miles and an Ingestion EPZ of 50 miles is not conservative and does not re-sult in adequate protection of the public health and safety.
The supplemental responses to the specific interrogatories are as follows:
A.1 Site-specific modeling of the impact of population 4
on Claes 9 accident consequences has not been con-ducted'by LILCO or reviewed by the NRC. The NRC has reviewed only the population in the low popula-
~
tion zone as related to emergency planning.* The Shoreham site has exceptionally high projected popu-lation. In the year 2000 it will have a population within 50 miles that puts it in the top 5% of all reactors; within 10 miles it is in the top 20% (see Figures III and IV). If population and the resulting consequences are considered in the NRC decision to t require implementation of Class 9 mitigation tech-niques, Shoreham would be a prime candidate.
- NUREG-0420, page 2-6. -
A-12 e
---,_-.c_ - - ---._.-._g _.I._, , , , . -_ _ _ _ _ _ _ _ , _ _ , , . , , , , ,, , _ . _ _ _ _ . , _ _ , , ,,_ , . , _ . __
F AGURE Irg
'CO.7ARISON OF POPULATION LEVELS k'ITHI*L50 MILES OF RE
' " " " "" "'""i "' " ' ' * '"*
W 3,g...g ig iig.. go.) i 6 j' " j'" "d
" eQPWta160= we t a'* 4
- to "'L II at til kuCsg ae etant $ef(S -
poa TMt vt am 2000 -
~
we am ocutatis. r toe ore r
roevt atie . t sw am I us.o
- m. ..wwa ueywsato=.tiex
- mu =v= Porssafic= . n ouc ar _
l ShOTeham /
(7 500,000)
./ --
d - -
~
.M _
g 1 . ~
(a -
Ee l -
s r
z=
e- 5 -
y in e.
2z 5 J. -- - . .. .- -
25 O-
.s7 ..
su 3 - .a -
O2 -
u:: . -
% o.
O4 -
s
- g e
4 w
.e
> 4
/
. . Y.
- ,,, [ .
I .
i 1 r. ;!n ,g_,
. t 11,1,:.1:::::t:t g J.1_,31! "1# # # #
e t it_ t "in'a iet i iny L li:t.
l y y ,1, t t tio i.o ei w se pa ws nw t
te a se a l ' '
6 is a so a? so mi m w e ei eLot s i eJ e6 1 2 CUMULATfvE Pf RCINTAGE OF Pt. ANTS WITH SURRCUN0iNG l POPULATIONS EQUAL TO. OR t Ess fl4AN. THE IN0lCAT ED V AL t
Source: NUREG-0348 A-13
PIGURE IV ' '
CO.itPARISON OF POPUI.ATION k'ITHIN TEN,jIII.ES OF FIACTOR SITES _
- i t io. 6.i. ii ..a..o . 66 ii oii au a i u n i i.-
_ i, u.a. t u i .... .j ur. :.e1 ici n e.,u. a l -
e*v. .i .e .. : ... . i . a .i n .
, at ti1 hisCLt as PL thiletil ,.
- 0m test e t am 7n00 me n't AN PCm/L hisO%
- 67.cc ..
u& D. As Po*ut A 7 e04
- 34 ass %.ae. pen POPU. Aficas e . 000 asu.ae per *C*UL ATIO.e
- 793 93 _
~ - ,,
/ ,,,,, -
V
~ '
hr er am /,* ~
(8'2 0C O) 4 -- --- . .. - . ... ]/1 ,--
g _
a _
g .
- - j -
z.
ee -
Em 2"
o' 5b l_ __ _ _ _
3 _
g" -
y -
a:
's _
S::
i o< .
g ..
e
,e - -
t _
_ I _
,,, r i e. i e i, ..!. r . .. ,, en......i .n- . :.uus ;.nu..m' .. ..t iir e u o r'!n i e < ' .' i n
..,..2 .. i : . w a = *w w s . .. .. ..n, n.
CUMULAtikt PERCENTAGE OF Pt ANTS WITH SURROUNotNG s . . . . , POPULATIONS EOt .L 70. OP LESS THAN. THE INOtCAT!O V ALUE Source: -NUREG-03^3 A-14
o . .
A.2: Site-specific modeling of the impact of meteor-ology on the consequences of releases resulting from Class 9 accidents has not been conducted bv ~
LILCO or reviewed by the NRC. As discussed in the preceding, the DBA releases represent less than one millionth the potential releases.
A.3: No site-specific modeling of the effec:s of to-pography on plume dispersicn within the 100' mile radius (approximately). of the Shoreham site have been conducted by LILCO or reviewed by the NRC.
Also, the effect of the Long Island Sound water on deposition velocity of radioactive materials is extremely t:ncertain.
Aj No site-specific modeling of the impact of land use characteristics and local access routes on A5 the accident mitigation measures resulting from releases due to a Class 9 accident have been de-veloped by LILCO or evaluated by the NRC. The NRC has only reviewed the road networks ar.d land use factors (as related to emergency planning) for the LPZ.*
A.6: Local and State (New York, Connecticut, and Rhode Island) emergency plans for even the 50 mile rndius from the plant have yet to be submitted or reviewed by NRC/ FEMA as set forth in the cur-rent regulatory requirements in Appendix E to 10 CFR 50. Thus, coordination of State and local jurisdictions during and following an emergency Also, at Shoreham have not yet been demonstrated.
LILCO has not provided the " implementing" emer-gency procedures as recently set forth in Section V of the emergency planning regulations.
A.7: No"Shoreham-specific analyses of plume lift-off and plume dispersion for releases resulting from Class 9 accidents based on Shoreham-specific re-lease times and release energies have been de-veloped by LILCO or reviewed by the NRC.
- -- - A.E- The Policy Statement has been superceded by the NRC Final Rule on Emergency Planning (45 FR 55,402) of August 19, 1980, which has an effective date of November 3, 1980.
- NUREG-0420, page 2-6.
A-15
..r.._.- ,,,-w.. . - , , - - ~ _ - ~.-- - - ,.. ,.---.,-, , ,r-. .ny. ,w, ,. ,- y .
i
. o l
A.9: See answer to A.8.
A.10: Same as answers to A.1 through A.9.
A.ll The facts and documents on which SOC now re-
""f2.liesaredescribedindetailinthegeneral
- response t'5 this interrogatory. The facts and documents on which SOC expects to rely
- during the Shoreham hearings will include the preceding as supplemented by facts and documents provided by LILCO and the NRC on this docket in the future, and by facts and documents provided by LILCO and the NRC in response to any forthcoming SOC discovery reques t. -
A.13: The documents on which SOC now Since relies the are docu-described in the foregoing.
ments are publically available, no copies have been provided by SOC.
A*
- SOC has not yet decided which witnesses it A. When this de-and ' cision will utilize in the hearings.
is made the response to these interro-6 '
gatories will be supplemented with the re-quested information. To the extent that SOC's present consultants are assisting in review-ing and/or responding to these interrogatories, their resumes are attached hereto.
A.17 SOC first engaged the services of their con-and sultants, MHB Technical Associates, in Decem-A.18: ber, 1979, and has had a continuing discussion with them regarding the contentions in this c as e . However, as of this writing, the con-sultants have not made any reports to SOC re-lated to this contention. General studies or observations that SOC now relies or expects to rely on during the Shoreha= hearing are set Also, see response to forth in the preceding.
A.ll and A.12.
- To the extent they are now known.
A-16 9
~ . - . . - - ~ - . , - - . , . . - . - - ,
, . . . . - - ..- - - - - --.-..-r-- - ~ ,,y.-- . - - - -
rr---
o .
l l
B. SOC CONTENTJON 2 In any melt-thrcugh resulting from a Class 9 accident at Shoreham, Indicactive materials may leach into the groundwater Models to or eventually migrate into the Long Island Sound.
assess the consequences of the possible contamination of water by radioactive releases on health, water supplies, costs, emer- j gency plans, or interdiction techniques have not been adequate-ly addressed generically in WASH,1400 or specifically by IILCO for Shoreham in the FSAR. SOC belf. eves that little or no pre-parations have been made to intercietcontainment the flow of contaminated buildings groundwater from beneath the react:tOmission of the liquid path-should a mcicdown accident occur.
way releasee from consideration in mitigating measures includ-ed in the eventual Shoreham emergency plan appears to be a fundamental deficiency. Therefore, SOC believes a description of potential interdiction systems, including emergency actions should and the potential safety and environmental improvements, be developed as part of the Shoreham safety assessment and EIS.
As set forth by LILCO in the FSAR, Long Island is underlain by soil deposits extending to depths ranging from a few tens of feet below sea level in northwestern Queens County to nearly 2,000 feet beneath the south of Suffolk County. In central Suf-folk County, test wells encountered rock at more than 1,400 feet below sea level. Rock isthe estimated to be approximately Shoreham site. The materials 1,000 feet below sea level at that overlie the bedrock and constitute Long Island's ground-
. water reservoir are unconsolidated deposits of gravel, sand, silt, and clay.* ..
The natural groundwater elevations measured across the area occupied by the Shoreham plant varied from'7 to 10 m1w, but be-cause the site might be flooded-during extremely severe storms, and since the geological formations are very permeable, the de-sign criteria for subsurface loadings included flooding consider-ations. The design of safety related structures has been review- .
ed to ensure they are capable of withstanding hydrostatic pres-sures and uplift due to a stillwater elevation of 26.0 m19 caused by inundation of the site by a storm surg'e. The natural ground-water elevations were temporarily lowered to various elevations How-during construction, as discussed in FSAR Section 2.5.4.6.
l
~-
ever7 tEere are no plans for permanent dewatering during the life of the plant.**
- FSAR, page 2.4-21.
- FSAR, page 2.4-27 B-1
. O 9
,--.,e---
, , - - , . .,,.---.---n,----,-,-,---,---,--,,.mw,--.,
k-
"~
In th2 gansric risk study, the euthors of WASH-1400 in-dicate that "the effects of contamination of water supplies has not been considered in detail"* and assume without a de-tailed analysis that streams and rivers would be contaminated for only a short time."** The inadequate analysis of water contamination is a major flaw in the WASH-1400 presentation.
The financial costs and the societal dislocation due to major l contamination from strontium-90 and other isotopes appear to be potentially very large. They deserve assessment in the Shore-ham safety assessment.
The WASH-1400 authors also assumed that partial core melt always led to complete core melt. They further assumed, be-cause of lack of direct experience, that once a core lost its initial configuration, it would be difficult to contain, i.e.,
even if the containment building above the core were not to fail, the core itself would probably melt through the concrete foundation. It is not clear that the core would necessarily follow this course. It may, instead, melt down but remain " con-tained from below by the foundation due to adequate upward heat transfer. If the molten core has penetrated the containment building through the bottom, its interaction with the sur-rounding soil and water table was not well-defined.*** Fur-thermore, core melt-through was not evaluated with the same comprehensiveness as were atmospheric releases because of the latter's more immediately observable adverse effect.****
Finally, the WASH-1400 report indicates'that airborne re-leases are much more significant contributors to the total risk
- WASH-1400, Main Report, page 76. ..
- Wash-1400, Main' Report, page 134.
- Battelle Columbus Laboratories is continuing the studies that they prepared for WASH-1400 in two main area - release frac-l tions and.the physical proc, esses of a core meltdown. Battelle is developing a new code, MARCH, to describe the thermal and
! mechanical aspects of a core melt. At this time, there ap-l pears to be substantial uncertainty in modeling the core melt penetration.
- Minutes of Meeting Five of the Risk Assessment Review Group,
! page 57.
l l B-2
1 than releases via the liquid pathway. The U.S. Department of Interior did not agree with the notion that the liquid pathway was as insignificant as indicated in the WASH-1400 report and recocmended additional study of the effects of variation of the hydrogeological conditions from site to site.* The NRC accepted the comment and instituted a research program at Sandia. The Sandia study results for a set of generic reactor sites were re-leased in draft form eo the NRC and subsequently to MHB in early 1980.**
Figure 2-A provides a schematic diagram of flow of informa- ;
tion in the basic computer programs used in the Sandia study.
l The general flow of information between the computer programs is indicated in the central column. Other programs either used to generate input data or else whose previously generated output was used as input data are given on the left side of the figure, while input data obtained from literature sources is shown on the right side.
Two sets of programs were used by Sandia as shown in Figure 2-A to describe the releases into the hydrosphere: INTER and WECHSL to provide information about both the appropriate time scales for the various possible releases and the overall com-position of the melt in different environments; and SOURCE to calculate the amounts cf radioactivity which go into each of the releases. The interaction of the melt with the structural ma-terial of the reactor, with the concrete floor of the cot tain-ment and with the soil beneath the containment, were taken as SOURCE described by the models employed in INTER and in WECHSL.
. calculates the amount of every important nuclide contained in each of the three basic types of releases for any given WASH-1400 accident category, and any core inventory. The output of SOURCE consists of the amount of each radionuclide released in-to the groundwater as a function of time.***
Transport of the radioactivity in the hydrosphere was des- GRDWATR des-cribed by two programs: GRDWATR and SRFWATR.
cribes the transport of the radionuclides by groundwater rove-ment from the area beneath the containment to any nearby sur-face water-bodies or wells, SRFWATR describes the subsequent With transport in lakes, estuaries, oceans and river systems.
- NUREG-0410, NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants, January, 1975, Task A-33, page 2.
- Sandia Study (Draft) performed for the U.S. Nuclear Regulatory Commission, "Effect of Liquid Pathways on Consequences of Core Melt Accidents," January, 1980, (Provided to Black Fox service list for construction permit proceeding), Figure 1.2.
- Sandia Study (Draft), pages 10 to 13 of Chapter 1.
B-3 l
l FIGURE 2-A FLOW OF INFORMATION SANDIASTUDk0FHYDROSPHERICCONSEQUEN
" _I -
RELEASE ORIGEN < DATA
-4" SOURCE J INTER (WECHSL) , '
L CORRAL ff GE0 HYDROLOGICAL 1
DATA GRDWATR BIOLOGICAL, y RADIOB2' LOGICAL 4 & SURFACE WATER SRFWATR DATA t
l~
" W POPULATION r ]
ISREM 11 9- & USAGE DATA a P PATHWAY RSS CODE I J (_
EXREM _.
h_
" d= h. ...
9 l
-B-4
% 9 g l -...... , , - - . . . . . _
--- ~ ~ ~--. . ..._ , _ "'-"*W., , , _ _ __
4 i
models have been employed to consider the movemen each waterbody; for the river system, a more ofcomplex The output GRDWATR compart- is the ment amount model has been developed.of each nuclide crossing the aquifer-sur the output of SRFWATR is es-interface as a function of time; sentially the concentrations of the radionuclides in the water and in the sediment as a function of location and time.* '
PATHWAY computes the radiation dose to t'.seEssentially populations at risk, along with the resultant health effects.
standard NRC Regulatory Guide 1.109 dosimetry models have been employed. The health effects models inThe PATHWAY output of PATH- are essential-ly the same as those used in WASH-1400.
WAY consists of the radiation dose received by the population for each appropriate pathway as a function of each of fouraccident categor factors:
time after the accident; The total resultant release; and relative leaching rate. health effects for each pathway a The differences between the radiation doses received h-through atmospheric pathways and through hydrospheric pat ways can be understood by comparing and contracting the path-as de ways themselves. The maj or differences ,
are reviewed in Table 2-A.***
The major conclusions by the Sandia authors in their draft report for plants like Shoreham without specific interdiction barriers were in part as follows:
- a. "In contrast the to the situation for releases to the atmosphere, most probable RSS meltdown cate-releases to the hy-gories result in the largest amounts of radioactivity drosphere. Significant are generally expected to be released to the hy-drosphere during any meltdown accident.
- b. At approximately 507,of all the sites considered, there is estimated to be essentially no radiation dose to the human population as a result of releases
- Sandia Study (Draft), pages 10 to 13 of Chapter 1.
- Sandia Study (Draft), pages 10 to 13 of Chapter 1.
- Sandia Study (Draf t) , Table 1.1.
B-5 e
- _ _ _ , , - . ._,__r.. , , _ _ _ - . . - _ _ , _ _ , , , , _ _ , _ , , , , . _ _ , . . . , _ , _ , _ , _ _ , _ _ . _ _ _ , _ . . _ , _ _ . ,
~
I l TABLE 2-A DIFFERiiNCES llETWEEN ATHOSPilERIC AND llYDROSPilERIC PATilWAYS I
i ,
ATHOSPilERIC REl. EASED llYDROSPilERIC RELEASED i
S0llRCE Atmosphere: Primarily more volatile Melt Debris: Primarily less radionuclides (1, Cs, Te,...). volatile RN's (Ru, Sr, La,...)
Sump water and dypressuriza-tion: Primarily more volatile RN's.
DISPERSAL Population rea'ched rapidly (hours). Population usually reached slow-ly (months to centuries longer).
tn b All RN's mov'e essentially together; Each RN moves through the ground deposition mechanisms differ only at its own rate; each RN moves for the noble gases. through the surface waterbodies 1
wi th i ts own set of int eractions.
PATilWAYS Primarily inhalablonandexternal Primarlly i nge s t. i on (ilrinking (ground). water, aquatic food) and extern-al (shorelines)
- Dominant pathways are relatively Some dominant. pathways are very simple. complex.
\
Populations are straightforward. Popul a t: i ons are not obvious.
I
!!EALTil Acute, latent and chronic. Primarily chronic.
EFFECTS INTERDICTION Source: not possible. Source: often possible Pathway: possible Pathway: possible.
i -
J 4
to the hydrosphere via the hole in the At con- an-tainment basemat formed by the melt.
other 15% of the sites, there is estimated to be a resultant radiation dose of approximately 2 to 3 x 10 6 person-rem, while at the2 remain-ing 35% the potential dose is about to 5 x 10/ person-rem for such releases.
- c. Variations of conditions at actual sites could easily result in the dos,es being either higher or lower than those obtained in this study by Likewise, rea-at least an order of magnitude.
sonable variations in certain portions of the modeling could result in estimated doses which differed from those presented here by an order of magnitude.
- d. The calculations indicate that if interdictive actions are not taken, then the liquid path-ways can perhaps contribute significantly to the risk of core meltdown accident."*
Releases for a core melt accident through the liquid pathway could constitute a major problem for Shoreham because of the site characteristicsThe setimpact forth inofthe FSAR and a core-melt summar-accident ized in the preceding.
through the liquid pathway can conceivably be reduced atThus, Shoreham if certain mitigating measures are employed.
the effects of interdiction, including emergency planning measures, both close to the site of the accident and farther fol-along the pathways to the aquatic and human population, lowing a study using the various Sandia models discussed here-in or equivalent, should be conducted for the Shoreham site.
The supplemental responses to the specific interrogatories are as follows:
B.1: Refer to general description in the preced-ing for a general discussion of the path fol-lowed by liquids and dissolved radioactive contaminants released from a nuclear plant during an accident. In general, radioactivity can be released'directly to the hydrosphere after a core-melt accident as a result of
- -~ _ _. __
- Sandia Study (Draft), pages 6-16 and 6-17.
B-7
, , , , ---r, ,--, ,w,-., , - - - , - - - - ,-,v.,- , - , , , - - - , ,,,--,e-,-, --- -w ,- -,y-~,, , ., t- w-w
leaching of the core-malt debris, escape of sumpwater, and depressurization Most of the long-of the containment atmosphere.
lived isotopes, including the actinides, are expected to be found primarily in the melt debris. Therefore, even though this release occurs relatively slowly, the im-pact for a soil site such as Shoreham can still be significant.
B.2: The emergency action lpvel guidelines des-cribed in NUREG-0654 for the Ingestion Pathway, including the guidance on size of the EPZ appear to be derived from WASH-1400 and, thus, results in inadequate assessment of radioactive releases through the liquid pathway.*
B.3: Yes.
B.4: The NRC regulations which require considera-tion of hypothetical Shoreham releases through the liquid pathway include 10 CFR 10 50.33, CFR 50, Appendix E, 10 CFR 50.34(b),
10 CFR 50.47, and 10 CFR 50.54 for the basis as described in the preceding.
B.5 lies are described in detail in theThe general
""f.ThefactsanddocumentsonwhichS0C response to this interrogatory.
and documents on which SOC expects to rely **
facts '..
will include
- during the Shoreham hearings,d by facts and the preceding as supplemente documents provided by LILCO and the NRC on this docket in the future, and by facts and documents provided by LILCO and the NRC in response to any forthcoming SOC discovery request and/or interrogatories.
B.7: The documents on which SOC now Since relies theare docu-described in the foregoing.
ments are publically available, no copies have been provided by SOC.
B.8, SOC has not yet decided which witnesses it will B.9, utilize in the hearings. When this decision and is made the response to these interrogatories B.10:
- NUREG-0654, pages 13~to 18 and Appendix 1.
- *To the extent t ey h are now known.
B-8 m
e
. .me. e .. . . .
,-ee., ,- ,--,-->,r---,,----~+--,e-- - , - p-e.,<. , , - . , - e - , . , e-,.v- er r,,-a -w-,w-~-y-w ,-, ye. e ,--.--em-
will be supplemented with the requested information. To the extent that SOC's present consultants are assisting in re-viewing and/or responding to these inter-rogatories, their resumes are attached
. hereto. _
B.ll SOC first engaged the services of their and consultants, MHB Techn'ical Associates, in B.12: December, 1979, and has had a continuing discussion with them regarding the conten-tions in this case. However, as of this writing, the consultants have not made any reports to SOC related to this contention.
General studies or observations that SOC now relies or expects to rely on during the Shoreham hearing are-set forth in the preceding. Also, see response to B.5 and B.6.
l ..
I I
l .. __. .. . _.
l l
B-9 l . . . .
l . . _ - -
_ _ _ . _ _ _ . . . . . _ . . _ _ - _ . . . = , . . . _. - _ . _ . _ . . . _ . _ _ . . . _ . _ . . . . _ . _ . . . . , _ . . _ _ _ .
I C. SOC CONTENTION 3 Regulatory Guide 1. 9 7, Re vis ien 2 was cre a te d as a res ult of the TMI-2 accident mad describes several new or revised measurements to assess the conditions in sad around the olant envir6ns during or'following an accident. The Regulatory Guide also includes consideration of ins truments for monitoring de-graded core cooling. ,
The regulations that Reg Guide 1.97 is responsive to are 10 CFR 50, Appendix A', Crite ria 13, 19 and 54 which de al with the subjects of Instrumentation and Control, Control Room, and Monitoring radioactivity releases.
The FSAR on Shoreham does not include information responsive to Re g Guide 1. 9 7 , Re v. 2. In the SER, the NRC has identified Reg Guide 1.9 7, Rev. 2 as a possible condition of the operating license if it is not met before the license is issued (SER, page 1-9). The specific language regarding 1.9 7 is as follows :
"The Applicant will be expected to upgrade pos t-accident ronitoring ins trumentation in accordance with Revision 2 to Regulatory Guide 1.9 7, ' Instrumentation for Light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Canditions During and Following an Ac ci dent . ' " (SER, page 7-13)
Reg Guide 1. 9 7, Rev. 2 has a table of variables to be measured ;
for BWR's (Table 3) including ranges 'of operation and conditions for qualification. Several of the key items are alr a included l in NUREG-0737 as requirements. and they ~ also have required sche-dules for implementation. It is SOC's position that pos t-accident monitoring ins trunentation (defined in Reg Guide 1.9 7 zad also in NUREG-0737) and the inadequate or degraded core cooling measure-ments (discussed in Reg Guide 1.9 7 mad NUREG-0737) should be implemented before fuel load mad included in the SER. Spe cific responses are discussed below:
C.1: This information is not clear at this time because LILCO has not responded to earlier informal inter-rogatories regarding the PAM devices they plan to install. Examples of ins truments not presently ,
installed include the in-core thermocouples for :
core temperature measurement and the high-range continuous Iodine monitor.
C-1 l
, , , , - , -- ,w, ~ .-.,w-. - - - - ---,
C.2: The s cope and rangt s of the measurements in Table 3 o f Re g Guide 7 D / , Rev. 2, are consicered suf-
, ficient to encompass a minimum set of instruments necess ary to meet the requirements and regulations lis ted above .
C.3 C.4: SOC believes the instruments responsive to Reg Guide 1.97, Rev. 2 and also NUREG-0737 should be imole-mented before fuel load to ensure safe operati.on of the plant and to protect workers and the public from accidental radiation exposure.
C.5 C.6: In addition to the information des cribed above , there will be responses to formal and informal interroga-tories It whichis also SOC has or will that anticipated submit LILCO to LILCO and the will submit NRC.
a response to the TMI issues in NUREG-0737 and this
' should contain information related to this contention.
C.7: The documents mentioned above are all in the public d'o main or were created by LILCO so they are not included as attachments .
C.8 C.9 C.10: SOC has not yet decided on When which thiswitnesses decision isit made, will utilize in the hearings .
the response to these interrogatories will be supple-mented with the requested information. To the extent that SOC's present consultants are assis ting in re-viewing and/or responding to these interrogatories SOC first engaged ,
their resumes are attached hereto.
the services of their consultants , MHB Technical Associates, in December, 1979 and has had a con-ti~nuing discussion with them regarding the contentions in this case . However, as of this writing, the con-sultants have not made any reports to SOC related to this contention.
C.11: No.
' ~
~ ~C .17 : N/A.
C-2 l
. . - - - - - , .-~ . - ,y,w, -,-. , - , . . ,.,~ ---- , . ,.-, .-, ,,w-,,,,--- .--,--,,,._,.,.,.,,---.---.---,-,,w,,,,----,w,-.m-nw,,---,m._v-- - , - - , - - , - - ,
I C.13 C.14: SOC believes that the prasent d2 sign of Shoreham does not include the necess ary instrumentation to con: ply with the requirements of General Design criteria 13,19 and 54.
C.15 that the design of the instrumentation C.16: To the extent system does not include the necessarv instruments to meet the requirements of GDC 13,19 "'and 54, then it demons trate d" may the also safety be said of the thatplan:1, LILCO has i notts sys tems and its ins tru-mentation.
e O * , em 9
C-3 1 y
O 9 -
y.- - - - - - - - - - ,, -
,w -- , .,.,=,,---~-,,,,,---ee-w-w-r, ---r- .-v-=-- -
-3+ - *1 ---w , - , -
I D. SOC CONTENTION 6 (a)(i)
Contention 6, as discussed herein, is narrowly directed towards providing answers to the numerous questions which have arisen over the adecuacy of implementation of Quality Assur-ance (QA) measures during t'ne design and construction of the Shoreham proj ect. At the outset it should be noted that no measure of the effecsiveness of implementation of the Shoreham QA progra= is either described by the Applicant in the FSAR
- or evaluated by the NRC in the SER. For example, the Shoreham SER is totally void of any reference to either the LILCO con-struction QA program or its implementation. Rather, the SER only addressed the proposed operating QA program.* Further, the ability of SOC to provide definitive and specific respon-ses to the interrogatories propounded by LILCO is severely hampered because of both LILCO's and NRC's failure to provide responses to informal discovery requests submitted by Suffolk County in February and March, 1981.**
The Quality Assurance (QA)*** program mandated by the NRC for commercial nuclear power programs by 10 CFR 50, Appendix B is intended to assure that the safety and reliability features specified by the regulations and the system designers are in fact implemented by the many different organizations involved.
Successful implementation of the QA program is essential to achieving expected levels of safety, yet there have been numer-ous allegations that an adequate implementation of QA programs is not presently being attained at Shoreham. Contention 6 is intended to address the root causes of these allegations.
One SOC concern addressed the timing of the QA program.
In 1970, the AEC, predecessor to the NRC', added 10 CFR 50, Ap-pendix B to the existing regulations. Appendix B contains the general _ requirements of a QA program for the design, construc-tion, and operation of a nuclear power plant. Nuclear stations such as Shoreham, which by 1970 were well along in their design and procurement activities, were required by the NRC to meet the intent of Appendix B in their remaining design / construction activities in so far as practical. Thus, the degree of com-pliance of the Shoreham QA program for design / construction with current QA requirements is an area of potential uncertainty which appears to require further clarification and specificity.****
~~ ~
- NUREG-0420, pages 17-1 to 17-6.
SOC chose not to submit a duplicative and redundant discov-l ery request.
- QA as used herein, includes both the functions of quality assurance and quality control.
- "** Shoreham CP application was docketed on May 15, 1968 D-1 5
, , , . , - - , , , . , -,e- .,--...-----,.-----,...,,..g,- - - . - - . . w-,.-v.,,-c. ,-e.-.,,w,.,m--.-.,e ...,.,-o.--,.-~~y ,,-- --,,,-.m .. .- - - ,
c - .-
I A second SOC concern addresses the adequacy of the NRC's Inspection and Enforcement (I&E) program. The NRC's I&E pro-verification that gram is intended to provide an independent the Shoreham facilities' structures, systems, and components are designed, manufactured, installed, and operated in strict accordance with the applicable quality assurance requirements.
In the past, the I&E program has not fulfilled this intended function. For example, the "after the fact" discovery by the NRC of quality deficiencies at the North Anna,* Browns Ferry **
and TMI-2*** plants raises serious questions about the ade-quacy of the whole I&E program. In particular, questions
> need to be answered about the NRC policy of relying on build-ers for primary inspections with NRC officials serving as only auditors.
The potential deficiencies in the I&E program are not new concerns. In partial response to the numerous criticisma of the NRC I&E practices, in May, 1976, the NRC provided Sandia Laboratories of Albuquerque, New Mexico, with over a quarter of a million dollars of funding to conduct a comprehensive, inde-pendent assessment of the NRC activities related to the review, com-approval, and inspection of quality assurance programs at mercial nuclear power plants. Specifically, the study assessed the effectiveness of the overall philosophy of the NRC QA pro-gram and the relative strengths and weaknesses of the practices employed to assure a high standard of quality assurance for nu-clear reactors.****
The Sandia study's final report was released as a NUREG series document in September, 1977.***** While the 16
~
- EMD-77-30, Allegations of Poor Construction Practices on r the North Anna 9uclear Powerplants, U.S. General Accounting 1977.
- Office, Washington, D.C., June 2,
- " Browns Ferry Nuclear Plant Fire," hearings before 16, 1975.
the Joint Committee on Atomic Energy, September The Legacy
- Kemeny, John G., et al, The Need for Change:
of TMI, Report of the Prisioent's Commission on the Acci=
dent at Three Mile Island, Vol. IV by the Technical Assess-i ment Task Force entitled, " Quality Assurance," Washington, D.C., October, 1979.
_ ~
- _ NRC Press Belease No.76-122, entitled " Independent Assess-ment of NRC Quality Assurance Activities Planned," May 25, 1976.
- NUREG-0321, A Study of the Nuclear Regulatory Commission Quality Assurance Program, U.S. Nuclear Regulatory Commis-l sion , Washin ; ton, D. C. , August , 1977.
1 i
D-2 i
vo , n wa --
-y n ~ ~ - - - - , - - , - , - ~ , - , . . - . . .,,,,-en..en,,>. --,,ev, -,wg w,~,---4,,-w-m- ,*w.we. . , - .,,~ - - - , - ,->,-4--, n,, < , --w,,,~m,----
recommendations of the study group were carefully worded in a positive manner so as not to imply that the existing NRC I&E l program is inadequate, the message is still clear. The re-port states in the summary that " based on the results of our survey and the stringent demands for reactor safety, we con-clude that further improvements are warranted in both industry l quality assurance programs and NRC regulations of these pro-grams." Specifically, the study concludes that " routine direct NRC inspection and testing of hardware be increased, and that data pertinent to quality decisio,ns made in the construction and operation of a plant be evaluated by the NRC on a routine basis. This includes the evaluation, for example, of radio-graphic and ultrasonic test da:a."
In 1978, a study, including Shoreham as an example, con-ducted by the General Accounting Office (GAO) described the following weaknesses in the NRC's I&E program during nuclear power plant construction which may result in QA deficiencies going undetected:*
"Although the Nuclear Regulatory Commission is responsible for assuring that nuclear power plants are constructed safely, it has not been independently testing the quality of con-struction work. The Commission should do this, plus
-- improve its inspection and reporting practices,
-- use the inspector's time and talents more ef-ficiently, and
-- better document its inspection findings.
The Commission is aware of the need for improve-ments and L,s made some changes, one of which is the assignment of resident inspectors to select-ed reactors under construction."
Other studies conducted by and for the NRC have identifiad opportunities for improvement in assuring QA program effective-ness.** Since a key factor in assessing the potential risk of a nuclear plant is the assumption of a disciplined, thorough
- EMD-78-80, The Nuclear Regulatory Commission Needs to Aggres-sively Monitor and Independently Evaluate Nuclear Power Plant l ~~ ConstrQction, U.S. General Accounting Office, Washington, D.C.,
l September 7, 1978.
- Examples include:
- a. NUREG-0397, Revised Inspection Program for Nuclear Pcwer Plants, U.S. Nuclear Regulatory L mission, Washington, D.C.,
March, 1978.
- b. NUREG-0425, NRC Inspection Alternatives, U.S. Nuclear Regula-tory Commission, Washington, D.C., February, 1978.
D-3
-- - -- - -, _ _ - . - - - - - . . . _ _ _ - . , , _ _ - ,, - , p.. , -
,-4 m.,-, , , - . _ . , - . + _ _ , , .
quality assurance program, any inadequacies in the NRC's I&E oversight of LILCO's quality assurance program may allow de-ficiencies in the Shoreham progra= implemen:ation which will pose a significant hazard to the public health and safety.
A recent and thorough investigation of the adequacy of QA program implementation was conducted by the President's Commission on the Accident at Three Mile Island. Some of their key findings related to quality assurance implementa-tion deficiencies were as follows:*
- a. TMI-2 went critical in March, 1978, and to commercial operation on December 30, 1978.
On February 8, 1979, GPU published on inter-office memo (TMI-II 7306); the subject was
' Incomplete work items TMI-II.' These were all the remaining incomplete work and en-gineering items at turn-over from GPU to Met Ed. Some of the items dated back to 1977 or longer. The estimated cost to com-plete these items was $2.1 million.
- b. The TMI administrative procedures require --
and the NRC review indicates -- that QA is generally involved in the review of mainte-nance and repair procedures on safety-relat-ed hardware. Met Ed had chosen not to use QA on other hardware -- a policy that re-sulted in maintenance on many critical plant systems not being covered by quality control.
- c. This diverse assignment of conf'iguration control responsibility, with minimum QA involvement, resulted in an inadequate understanding of hardware configuration at the site. This mr.y have been a direct factor in th! cause of the accident. A significant part of the TMI configuration control problem was the state of the as-built drawing files. NRC discussions with Met Ed personnel and the I&E audit of TMI document that TMI did not :urrently meet
. __. _ . . _ , J:RC requirements. .The arenitects were as-signed responsibility for maintaining up-to-date drawings. During the latter stages
- Inskeep, G.W., "The Cause and Effect at Three Mile Island,"
Quality, June,'1980, pages 42-45.
D-4 9
I of construction and start-up, a large ECM backlog resulted in updated draw-ings not being available for months.
In these same interviews, two !!:I em-playees indicated that no group was currently assigned responsibility to up-
--- date equipment drawings . The master
' transparent -drawings were to be onsite but the manager of generation engineer-ing, and others at There the sitearewere threeunsure places of their status.
where site personnel must go to deter-mine what a given system is supposed to look like:
e aperture card index e outstanding ECM file e TMI completed change modification packages
- d. Audit reports were found to be well-written and met requirements. The
- -- follow-up system was being used, al-though extensions were being granted too frequently. In many cases, the TMI auditor did not accept the corrective action proposals and held the finding open. As of June 5, 1979, 48 nuclear findings were still open; six had been open since 1977. Audit reports and station information are widely distri-Kowever, it buted to Met Ed management.
was not evident that Met Ed management was acting on the information.
~
- e. The TMI QA supervisor indicated that he was unable to cover all maintenance and repair procedures requiring verification. He also stated that in many cases QA personnel are not available to cover hold points in main-tenance operations. They They attempt to follow re.lew the up during the operation.
work request to assure that all data is
- -- -- - complete and all steps were conducted. ATMI QA QC personnel work only the day shift.
representative is on call for second shift maintenance and surveillance procedures; but, D-5
t reviews are usually conducted the next day.
Frequently performed standing maintenance, such as the repacking of valves, is csually i verified after the fact. l In summary, the potential impact of any QA deficiencies on safazy may be significant, and the misinformationThus, that is a result' of this uncertainty is equally cf concern.
SOC believes it is essential that LILCO demonstrate that the Shoreham QA program was adequately implemented. SOC, there-fore, endorses the concept of an independent third party de-tailed reinspection and retest of two complete electrical and mechanical systems as proposed by Suffolk County in the draft County /LILCO settlement proposal. Such a reinspection and retesting of equipment should provide a significant mea-sure of the actual quality achieved at the Shoreham plant.
The following responses will be supplemented by SOC fol-lowing the review of the requested discovery documents. The supplemental responses to the specific interrogatories at this time are as follows:
D.1: " Root" can be defined as meaning "something that is an origin or source (as of a condi-tion or a quality)."* The wording is con-sistent with the NRC order on the South Texas Project that Houston Lighting and Pow-er Company should " develop and implement a more effective system to provide for the identification and correction of the root causes of the nonconformances which occur."**
(emphasis added) -
D.2: Criterion 2 of Appendix B to 10 CFR Part 50 requires that LILCO' establish a quality as-surance program at the earliest practicable time. SOC believes that this means that the QA program should be established for each work element prior to the conduct of the work element.
D.3: Applicant's contractors, agents, or consul-tants are defined in Criterion 1 to Appendix
-' -~~- -"
B to 10 CFR Part 50. In general, SOC refers
- See Webster's New Collegiate Dictionary.
- Federal Register, Vol. 45, No. 92, May 9, 1980, at 30756.
D-6 e
-.-r -- - ., . - - - - . . , - - - - ,
, , , - - - , , , - - - , - , - - , .-y ,, y_ -,c- ,-
to those parties involved in on-site activi-ties. Therefore, SOC does not intend to gen-erally include equipment manufacturers in this classification D.4: SOC believes that persons or organizations
-- completed york elements prior to the estab-lishment of QA program which violates Appendix B to 10 CFR 50 and Criterion 1 of Appendix A to 10 CFR 5'O.
D.5: The requirements for independence from cost and schedule considerations meant by SOC is set forth in Criterion 1 to Appendix B to 10 CFR 50.
D.6: Identification of the persons and organiza-tions which lacked sufficient independence .
from cost and schedule is not yet complete as a result of the discovery process. For examples, see worker allegations 14, 21, and 30.
D.7: Independence is lacking at times during the construction because of shared responsibili-ties by a single individual for the conduct of the work element and for the QA of the work element. Such individuals lack the necessary organizational freedom.
D.8: The requirements for stop work authority are set forth in Criterion 1 to Appendix B to 10 CFR 50.
D.9: Same as response to' question D.6.
D.10: Same as response to question D.7.
D*11 Qualification and training requirements of
^"d on-site QA personnel are set forth in Cri-j
- yo,
terion 2 to Appendix B to 10 CFR 50. Worker l allegations 7, 17, 19, 20, 21, and 29 provide, in part, the basis for this issue.
D.13 Requirements for theincontrol forth of special Criterion 9 to pro-Ap-cesses are set pendix B to 10 CFR 50. Part D of this to D.18:
(
D-7
' ' ~ - - - - - - - - - - - , _ _ _ _ . . . . _ _ __ _
contention is in part based on worker allega-tions 2, 10, 17, 19, 20, 21, 23, and 29. SOC also relies on I&E inspection reports of on-Shoreham as part of the basis site activity at of this contention.
D.11 Criterion 15 to Appendix B to 10 CFR 50 sets and - forth the QA program requirements for noncon-Part E D.20: forming materials . parts, or components.
of this contention is in part based on worker allegations 1, 4, 5, 6, 8, 12, 15, 22, 23, 26, and 30 as well as I&E reports which Other address exampleson-site activities would include at IEShoreham.
Bulletins No. 74-02, 79-10, and 80-02.
D.21 Criterion 16 to Appendix B to 10 CFR 50 sets to forth the QA program requirements for correc- ~
D.23: tive action and provides the meaning for the terms utilized by SOC in Part F of this con-tention.
Part F of this contention is in part based on worker allegations 2 and 22 and Viola-tion A as well as by I&E reports which address on-site activities at Shoreham.
D.24 Criterion 18 to Appendix B to 10 CFR 50 sets and forth the QA program requirements for audits D.25: and provides the meanin6 for the terms util-The ized by SOC in Part G or this Contention.
contention (Part G) is in part based on worker allegations 1, 5, 6, and 14 as well as by I&E reports which address on-site activities at Shoreham.
D.26: SOC believes that I&E Investigative Report 50-322/79-24 addresses the symptoms rather th'an the root causes of the worker allega-j tions. For the preceding reason, SOC be-lieves the I&E report is deficient.
I D.27: The deficiencies are set forth in Parts A through G of this contention.
! _ , . . . _ . - forth The basis for the deficiencies-are set l
D.28:
I in responses D.1 through D.25.
l l
D-8 l
l l .
l
I D.29 The facts and documents on which SOC now and relies are described in detail in the gen- ,
D.30: eral response to this interrogatory. The facts and documents on which SOC expects to rely *during the Shoreham hearings will include the preceding as supplemented by j facts and documents provided by LILCO and '
- the NRC on this docket in the future, and by facts and documents provided by LILCO and the NRC in response to SOC discovery requests and/or interrogatories.
D.31: The documents on which SOC now relies Since the are described in the foregoing.
documents are publically available, no copies have been provided by SOC.
D.32, SOC has not yet decided which witnesses it D.33, will utilize in the hearings. When this and decision is made the response to these in-D.34: terrogatories will be supplemented with the requested information. To the extent that SOC's present consultants are as-sisting in reviewing and/or responding to these interrogatories, their resumes are attached hereto.
D.35 SOC first engaged the services of their and consultants, MHB Technical Associates, in D.36: December, 1979, and has had a continuing discussion with them regarding the conten-tions in this case. However*,' as of this writing, the consultants have not made any reports to SOC related to this contention. SOC General studies or observations that now relies or expects to rely on during the Shoreham hearing are set forth in the preceding. Also, see response to D.29 and D.30.
D.37: Yes.
_..D 38.:
SOC's investigation is incomplete at this Examples of potential deficiencies
~
I time.
are set forth in responses D.1 through D.25.
As discussed in the preamble to this inter-rogatory, SOC also expects to rely on the
- To the extent they are now known.
D-9
- --- - - , . - , , , - - -.m - - - - _ . - , , , ------.w.-.,,--,-m,,.- -
- m. .n, . , , , . . - - - - - ,vg-, - - - . . -- -- -,~- g. - , , ,
I results-of the independent reinspection being proposed by Suffolk County in the draft LILCO/ County settlement agreement.
D.39: Yes.
-D.40: See preamble to this interrogatory as well as re'sponse D.38.
e G e 4
O i
a m; e . ..
{
d 0
8 i
l-i D-10 e
b f
o m y , , . . g,,,.,m--...m._%-._7,_~,..,.. ., ,, ,p. w ww. c r,__, . yy--.. .,,.m.ew,Sv.,-w,,.emw_#,c my.mm_wwm,.m,..,. _-..,me +, m m y .w ,_
l F. SOC CONTENTION 8 General Design Criteria 13 requires , among other things ,
that instrumentation be provided to monitor variables and sys-tems that can effect the integrity of the reactor core and these meas uremen ts cover anticipated operational occurrences and the accident conditions . In designing the protective systems ,
the fuel temperature regulaflons require analyses to show that will remain below 2200cF during design basis accidents (10 CFR 50.46 and Appendix K) . However, 'the experience at T>E-2 has caused the re-eveluation of the adequacy of present systems to properly detect the onset or approach of inadequate core cooling con ditions . The TMI-2 accident exceeded the design basis acci-den t , exceeded the calculated fuel temperatures , and resulted in loss of integrity of the reactor core.*
In order to detect the onset of inadequate core cooling at Shoreham, the fuel cladding temperature and integrity should be monitored by a direct measurement. Presently, the Shoreham design ** does not include such a measurement, relying ins tead on variables such as coolant level and temperature to infer or estimate an indirect indication of the temperature of the fuel-and the integrity of the core.
At least one other.BWR of the Shoreham type (the LaSalle plant in Illinois) has been required to LILCO installhas thermocouples not committed for incore temperature measurement.
to an incore temperature measurement technique but has s tated an opposition to installing incore thermocouples.
SOC believes there is a need to provide a to direct the integrity measure-ment of the of theand fuel variable indicative the reactor core (i of.e.
the, the treat fuel temperature) in orde~r to adequately protect the public health and safety and to meet the regulations set forth in GDC 13.
A Reoort to
- The problems encountered are described in TMI:
the Commissicners and to the Public, Vol. II, Part 2, directed by Mitchell Rogovin, U. S . NRC , J anuary , 19 80, and T32 ,
- --Report of the Presidents Co=4 ssion en the Accident at directed by John Kemeny, October,19 79, Main Report.
- As documented in the Shoreham FSAR.
F-1 e ,, w-e - - n--- ,-,-e e,.w,-~.-.-,,,,..-..e.-,e- w + ~ ** Sw~--
Supplemental responses are as follows.
F.1: See the above statement.
F.2: SOC does not propose any particular design or type of instruments, only that there be a direct means of measuring the fuel temperature to detectOne theexample threat to the integrity of the fuel cladding.
of such a system is the use of incore thermocouples to measure fuel cladding temperature.
F.3: The regulations in 10 CFR 50.46 do not dis tinguish between BWR fuel and PWR fuel in their requirements.
Also the cladding for both types of fuels is basically a zirconium alloy and s ubj ect to oxidation and degra-dation under high temperature conditions and s team environment. Voiding and inadequate cooling on a BWR can create the same types of fuel failures as occurred at TMI.
F.4 F.5: As indicated above, SOC will rely upon the FSAR for Shoreham and TMI-2 documents such as the Kemeny Report, the Rogovin Report, NUREG-05 78, NUREG-0660, NUREG-0737, and any other published NRC or public reports related to this subject which may be issued be cween now and the hearings . SOC may also rely upon answers to interrogatories which have 'been or may be sent to LILCO mad the NRC.
F.6: Since all the above documents are 1) in the public.
domain, 2) filed on this docket, or 3) prepared by LILCO, they are not included in this submittal.
F.7 F .,8 F.9:
SOC has not yet decided which witnesses it will uti-lize in the hearings. When this decision is made , the response to these interrogotories will be supplemented with the requested informar' an . To the extent that assisting in reviewing SOC's present consultants a.
and/or responding to these in~.errogatories SOC firs t
, their engaged the
-- - -. - resumes are attached hereto.
services of their consultants , MHB Technical Asso-ciates, in December, 1979 and has had a continuing dis cussion with them regarding the cententions in this case . However, as of this writing, the cons ul-tants have not made any reports to SOC related to this contention.
F-2
. ,_ - _. g - . _ . - . -.
I i I
F.10: No , none at this time .
F.ll: N/A.
F.12: See above .
F.13: ,
See above . -
F.14: To the extent that the Shoreham design does not include a means for direceiv measurine fuel te:rera-ture , the Applicant has not' " demonstrated" the '
safety of the plant and its ability to comply with GDC 13.
F.15: See above .
't i .
4
- M. e op. ee.
e i
9 F-3 4
- . . , - , -c-,..m,- -
- n. -
-,-,,-.--.-----,,.--.e .,c-ms .-r- *=wm,,.. w , ~ ~ .enmev .rw e- . --,-. r- w w-ew- we e-e----
t G. SOC COMTENTION 9 TMI-2 showed the s afe ty i=p act of inadequate indication to the ope rators o f the s tatus o f s afe ty sy s tems . The operators at TMI-2 were unaware of the status of the Auriliarv Feedwater Sys tems at a time when they anticipated automatic initiation of their functions. _
Similar sys tem problens can occur at Shoreham. First, there has not been a complete re-evalua' tion of systens and components important to safety at Shoreham. Thus , there is no assurance ~
that all the appropriate necessary sys tems and components have been provided with the necessary bypass indication. Second, there is no trans fer of bypass or disabled conditions to the remo te shutdown panel. Thus , in an accident con dition , the operator may not be able to know if his necessary systens are operable , bypassed, or dis abled. Finally, for sys tems impor-tant to safety but not presently provided with bypass indica-tion, there is no indication that operating procedures will be adequate to provide the intended safety. protection function required in 10 CFR 50.55a(h) .
The Shoreham FSAR lis ts several sys tems where Reg Guide 1.47 has apparently been applied.* However, several sys tems required for s afety , safety-related, or required for safeThe shut-down** are not shown as having Reg Guide 1.47 applied.
words used to describe the bypass design basis are insufficient to draw conclusions on the adequacy of the approach:
" Manual bypass of instrumentation and control equipnent components is also ender the control oc the main control room operator. If the ability to t-ip some essential part of the system has been bypassed, this fact is continu-ously indicated in the main control room."***
(emphasis adde d) l l
Reg Guide 1. 47 is shown as applicable to only bypass inputs ,
manual switch inputs , and trip logic actuator outputs .****
I - * - - -FS AR , Figure 7.1.1-2,- Codes and S tandars Applicability Matrh , Rev. 16, April 39 79 (subsequent to TMI).
- FSAR, Figure 7.1.1-1, Plant Ins trumentation & Control Systems - Classification , Re v. 16 , Ap ril, 19 79.
- FSAR, Item P , page 7.1-11, Item P , p age 7.1-13.
- FS AR, Figures 7.1.1-4, 7. 2. 2- 5 , and 7.1.1-6 .
l G-1
, ,- --r _ .-. _ . - .
r * -- r- - - + - nv -- -n -
Supplemental responses are es follows :
G.1: Deliberately disabled may be interpreted in the words of Reg Guide 1.47 as " bypassed or made inoperable ."
G.2 G.3: Informal interrogatories related to the operator interfacas and human factors design were submitted to LILCO on March 19, 1981 and Ap ril 6 , 19 81. Only a fraction of the firs t setSince and LILCO none of the second has broken off set has been answered.
the informal dis covery process , the remaining inter-rogatories will have to be submitted as formal inter-rogatories and their responses awaited prior to Also, until a responding to this interrogatory.
thorough review is made to de terrine a revised lis t of systens and components important to s afe When ty , it is difficult to answer this interrogatory.
such a review is completed and responses to inter-r:gatories are obtained, this response will be s upplemen te d.
G.4: See above .
f G.5: In addition to the FSAR, SER, Kemeny Commission Re-port, Rogivin Committee Raport , NUREG-05 7 8, 0660, and 0737, there are responses to interrogatories which may provide informative data. Also, any reports or documents published between now and the time testinony is prepared may be used.
G.6: Since all the listed documents are in the public domain or are filed on this docket or are prepared by LILCO, they are not included here.
G.7 G.8 .
G.9: SOC has not yet decided which witnesses it will When this decision is made, utilize in the hearings.
the responses to these interrogatories will be supple-mented with the requested data. SOC's present con-s ultan ts (MHB), to some extent have contributed to the response and review of these interrogatories.
- -- -- -- Therefore , their resumes are attached. SOC first engaged the services of their consultants in De-cember, 1979 and has maintained a continuing dis-cussion with them regardirg the contentions in this case. However, as of this writing, the consultants have not made any reports to SOC related to this con ten tirn.
G-2 T em .e 9
- ----,-,-,,.-,--e-.,,-,-,-w--.--.,- . , , . - , , . , - . , . - - .-
,m . - - - . . . . ..,., , . - . - . . - , -m,.-,-.,.--,..,-.e, y.---- - - , -
G.10 this time, nor are any foreseen other G.ll: No, not at than preparation of tescimony.
G.12 G 13: See the preceding.
G.14 the systems do not meet Re g G.15: To the extent that Guide 1.47 and 10 CF3 50.55a(h), the Applicant has not demonstrated the safe:y of the plant, its systems and its instrumenration.
O O4 e
)
4
@ e em, me G-3 S *g9 $
-, - , , ,, ..,,r-r--- -,+-r-,, e.,---w.., , - - - - - , - - .--e, r--lm ,, e,, --,,-m , - , , ,,--n,e,----v,.--, --r,ww_--, -,w-n-,---< - - --
I H. SOC CONTENTION 12 (P ar: Two)
The generic design deficiencies of :he Mark II type con-tainment utilized at the Shoreham f acility were firs t identified in 1974 and have been the subject of anThe extensive design research, features and tes t, and analysis program since that time.
problets have been wet 1 documented in numerous industry and NRC reports , but the April,1981 Shoreham Safe:y Evaluation Report (SER) contains in Section 6, perhaps the most recent condense d summary of the Mark II program. The cause of1)the failure Mark II con-to account tainment deficiencies is basically twofold:
adequately for all hydrodynamic loads the containment may experi-ence during the course of a loss-of-coolant' accident, and 2) f ail-ure to adequately consider loads generated in the suppression pool and transmitted through the plant structures and components by the blowdown forces from s team released bv the s a fe ty-relie f valves (S RV's) into the pool. The seriou'sness of these omissions were realized in early 1975 when a generic program (LTP) was assembled by the affected utilities and the General Electric The LTP Company, with assis tance from numerous consultants programs . and associa:ed
-- has grown extensively as the included tes analyses raised additiet al issues , and the completion date of the LTP has been unable to keep pace with the cons Consequently, truction schedules design at several of the plants , Shoreham included.
and cons truction decisions were made at Shoreham in advance of confirmatory LTP completion, and a number of open items remain at this time.
' Section 1 (1. 7) of the SER lis ts the remainingIt open is interes-items identified at this time by the NRC for Shoreham.
ting to note that more than one-fourth of the 61 open items are directly related to containment issues , giving some indication of the as yet incomplete state of review.
LILCO ha.s proposed for Shoreham that some of the Mark II '
load definitions will be in compliance with the final LTP results rather than committing to the more conservative lead plant cri-teria.
Two of the SER open items apply directly te downcomer loads (Steam Condensation Downcomer Laterial Loads and Steam Condensation Os cillation and Chugging Loads) . A third open Downcomer Fatigue Analysis , also applies dire ctly to l item,
--downcomer adequacy. It is not likely that the LTP final results
) will drama:ically change the containment requirements but it is l
dis concerting to learn that the loadings are so close to margin that LILCO must wait for the LTP results rather chan go with the lead plant criteria. This same conce rn exis ts for the other remaining open itecs.
I I
H-1 .
l I _ _ .
I The original s urce of co:.:ern on this contention area was learning in June of 1979 that Commenwealth Edison was forced to of their icoose an approximate 12-mon th delay in the comple tit:
LaSalle plant in order to incorporate a massive dcwntomer re-s traint sys tem.
This decision was reportedly forced by a change in the submerge d s tructure load de finition The, the sameof details load as change this one that remains open in <he Shoreham SER.
were communicated verbally to one of SOC's consultants on June 29, 1979. Additional concern was, develeped on March 5,1981 when it was learned from LILCO in a Shoreham Mark II meeting in Bethes da that the downcomer restraint system had been modified to remove all downcomer res traint ties to the pe des tal and suppression pool wall. This was apparently required to uncouple the pe des tal load input , but causes amplification of the down-comer assembly from the seismic input.
It is therefore our position that Shoreham's Mark II sys tem s till has not been demonstrated to be in compliance with50.
16 and the The requirements of 10 CFR 50, Appendix A and GDC 4, NRC'.; Mark II SER final approved loads are not scheduled until is scheduled for the mid-summer andfinal LILCO's the Design Mark II Assessment SER SupplerentReport will not be avail-fall.
able until November,1981 at the earliest so it is impossible to ascertain at this time if the sys tem is accep table.
The following specific answers supplement the above para-graphs in replyir g to LILCO's interrogatories :
H.1: See discussion in introdd'ctory paragraphs.
H.2: We have not been able to determine this at this time as the final " loads" are not yet available nor is the final version of the DAR.
H.3: Failure of the downcomers through overstress or fatigue can cause the s te am re le ase d during a LOCA to bypass the suppression pool and cause over-pressurization of the primary containment, le akThe
- o r rup ture , and subsequent offsite releases.
specific violations are :
- a. The containment and other hardware important to safety are no't designed to " accommodate the effects of and to be compatible with the environrental conditions associated with. . . .
postulated accidents . . . . ." as required by GDC-4.
H-2 ,
O e
H.3: b. The containment design does not as s ure that conditions important to safety "are not ex-cseded for as long as postulated accident conditions require" as required by GDC-16.
- c. The containment design basis does not assure that che sys tem "can acco=modate , without ex-cee ding the design le akage rate . . . . . " condi-tions resulting from any loss-of-coolant accident as required by GDC-50.
H.4: s. See introductory paragraphs and responses to 2 and 3 above ,
- b. SOC intends to review DAR revisions as they become available and file interrogatories on LILCO and the NRC. .
H.5: Documents that SOC now and during the hearings intends to rely on include the Shoreham FSAR and SER, including future revisions and supplements ,
the Design Assessment Report and subsequent re-visions , all Mark II LTP repcres by the Owners Group and the NRC companion reports. SOC will also rely on documents received in responses to interrogatories that may not be included in the above grouping.
H.6: All documents identified in 5 above are or will be in the public domain and" are or will be in LILCO's possession.
H.7 H.8 -
H.9: SOC has not yet decided which witnesses it will utilize in the hearings. When this decision is ma de , the responses to these interrogatories will be supplemented with the reques ted data. SOC's present consultants (MHB), to some extent have contributed to the response and review of these inte rrogatories . There fore , their resumes are attache d. SOC firs t engaged the services of their
-- -- - consultants in December, 19 79 mad has maintained a continuing discussion with them regarding the contentions in this case. However, as of this writing, the censultents have not made any reports to SOC related to this contention.
H-3 G
, ,, .-,y.., , , . . . , - - . ,-m- - _- -, ,
,, ,.y,---. , - , --+-~,
H .10 . a an d b : No.
H.ll: a en d b : N/A.
H.12: Yes.
H.13: a and b .
See answer to ques tion 3 above .
H.14: Yes. 9 H .15 : a - d: See introductory paragraphs and ques tion 3 response.
amen e me , me e
l l
l l -
HL 4
I. SOC CONTENTION 12 (Part Three)
This contention deals with the issues of hydrogen generation and control and the possible exceedance of 10 CFR 100.11(a)(2) exposure guidelines that could result from a fission product release associated with a degraded core accident at Shoreham.
The basis for this concern is the TNI-2 accident where core damage occurred due to inadequate cooling were hydrogen generation and release from the primary coolant system far in excess of the quantities specified in 10 CFR 50.44 This problem is widely reported in NRC and other nuclear indus try literature and is presently planned to be considered under a rulemaking proceeding as noticed in the Federal Register on October 2,1980 (pages 65247 & 8 and 65466-477) . The actual s chedule and conclusion of this proceeding is not known at this time but the .'act remains that serious doub t exis ts regarding the adequacy r f the Shoreham plant to protect the public from the radioactive releases that could result from such an event.
The issues of concern are 1) the probability of a loss-of-coolant accident resulting in excessive hydrogen quantities ; 2) the lack of commitment by LILCO to provide and operate the plant with adequate hydrogen control features as discussed in the NRC SECY-80-107 document dated February 22,1980 (Proposed Interim Hydrogen Control Requirements for Small Containments) : 3) no indication on LILCO's part to upgrade the hydrogen reco=biner capabilities at Shoreham so as to improve the post-accident response time in reducing and controlling the quantity of free hydrogen that may exis t within the containment, and 4) no de=onstration on the part of LILCO to analyze an/or upgrade the primary containment so as to be able to., withstand pressures beyond the containment design conditiens. Similarly, and di-rectly to one of the points contained in the ASLB Order of July 2,1980, LILCO has presented no information on system or hardware modification intended 'at Shoreham that would reduce or prevent the " likelihood of an operator interfering with ECCS operation."' Neither has LILCO dire ctly responded to the new Board question: "Why is inerting for the Shoreham containment not reco= mended as a result of the TMI-2 accident while inerting is recommended for later plants of similar design?"
It is perhaps premature to attempt to respond specifically to the detailed interrogatories of LILCO since this contentien
~hai' H6t Pet been particularized. However, the following answers will supplement the above paragraphs with regard to LILCO's specific ques tions :
1-1 9
,-c- , , , - - - . - - , - - , - -
- - , . - - - . . - - - - , - , - nn .-. -- - -- -
I I.1: Yes.
I.2: SOC intends to provide future contention responses in accordance with the proceeding schedule which remains to be established by agreement between the Board and.the parties.
I.3: Yes. ,
I.4: Notwiths tanding the Board's apparent decision te not waiver the provisions of 10 CFR 50.44, the Shoreham ple.nt must s till comply with the reautre-ments of 10 CFR 50, Appendix A, General Desi o.
Criteria (GDC) . Our belief is that the Shoreham f acility will not meet the requirements of GDC 4, 16 and 50 because subs tantial .quantia.ies of hydrogen may be generated in excess of the 0.41gn basis 'of 10 CFR 50.44 I.5: As indicated in the introductory paragraphs above,
- deficiencies in the facility's provisions for hydrogen generation and control may cause offsite releases in excess of the 10 CFR 100 guideline values.
I.6: a. SOC relies on the fact that the TMI-2 accident demonstrated the high likelihood of release of larger quantities of hydrogen than are currently specified in 10 CFR 50.44, and on the facts , as provided in the Shoreham. FSAR, that the facility is not designed to handle such quantities. See the introductory paragraphs of this response for additional references to summary documents and include d f acts ,
SOC intends to file interrogatories on LILCO and
~
b.
the NRC on this issue . Facts thus ob tained and also obtained through future research will be '
utilized in the hearing.
I.7: a. In addition to the documents identified in the
-- - -- introductory paragraphs , SOC, ,of course , relics l on the FSAR, SER, and other design documents specific to the Shoreham docket. We additionally rely on the TMI-2 reviews conducted by and for the NRC, specifically the Kemeny Report and App en dices , the Rogovin Report and Appendices ,
NUREG-05 78, NUREG-0660, and NUREG-0 737.
l I-2 l
l -
I
- b. 50C will relv on documents expected to be I.7: received in response to interrogatories :o be filed on LILCO and the "RC as well as toWe documents idcntified by future research.
will also review future LILCO submittals to the NRC, specifically those pertaining to TMI and Mark 11 ques tions , for applicable information. .
I.8:
All documents identified in response 7 are in the public domain and can reasonably beNone expected have been to already be in LILCO's possession.
destroyed to our knowledge.
I.9 1.10 decided which witnesses itmade will uti-1.11: SOC has not yet When this decision is ,
lize in the hearings.
the response to these interrogatories will Tobe thesupple-extent mented with the reques ted information.
~~ that SOC's present consultants are assis ting in re-viewing and/or responding to these interrogatories , SOC first engag their the services resumesofare accached their consultantshereto. , MHB Technical Associates , in December, 1979 and has had a continu-ous discussion with them regarding the contentions in this case. However, as of this writing, the consul-tants have not made any reports to SOC related to this contention. ..
I .12 : a. No.
- b. We do not know at this time if any will be made.
I.13: .N/A.
I.14. Yes.
- a. & b. As described in the introductory paragraphs ,
I.15: we believe the Shoreham facility does not comply with the requirements of 10 Spe CFR 50, :
cifically l '
Appendix A, GDC 4, 16 and 50.
(1) The containment and other hardware impor-tant to safety are not designed to "acco=ne-date the effects of and to be comoatible with the environmental conditions asso'ciated with
. . . . .pos tulate d accidents . . . . . " as require d by GDC-4 I-3
.----p
I design does not assure that I.15: (2) Tne containment to safe ty "are not d conditions important exceeded for as long as postulated acci ent condition = require" as required by GDC-16.
assure (3) The that containment design basis does notthe s "
exceeding the design leakage rate . . . . . co accident as required by GDC-50.
I.16: Yes.
See response to question 15 a and b 1.17: a. through d. and information presented in intro-ductory paragraphs.
V
- **e. , ,,
9 I-4 e
- , , _ - , _ _ - - . - . . - - - . . p- _.,,,y_. 9.. ,_ ,,_7., y, , , ., . . ,, y. ,, c. m,-.,.m. ,#4 _ _ . . , , , . , , _
J. SOC CONTENTION 15 NRC Bulletin 79-26 notified current and future operators of BWR's of a newly observed failure mode of reactor centrol blades heretofore not considered in the accident analyses con-ducted in licensing reviews. This failure mode involves the stress corrosion cracking of the control blade tubing which
~
contains the neutron adsorbing boron-carbide (B 4 C) powder which provides reactor core control and shutdown. As s ta te d in the Shoreham Safety Evaluation Report , stress corrosion cracking of control blade tubing is another example of a f ailure phenomenon that has been observed in a BWR fuel system component that is not an integral part of the fuel assembly e
its'lf. In this case , hot cell examinations of both foreign and domestic control blades revealed cracks in some of the stainless steel tubing and a loss of boron (B 4C) from some of the tubes . The safety significance of boron loss is its impact on shutdown capability and scram reactivity. Although shutdown capability is demons trated by shutdown margin tes ts after re-fueling, the calculated control blade worths used in the safety analysis are based on the assumption that no boron loss has o ccurre d.
Based on the hot cell observation that there is no boron loss until 50% local B-10 depletion (burnup) is attained (NEDP-24226), GE assessed the potential effect of the boron loss on shutdown capability, CPR reduction, and the consequences of control rod drop accidents and concluded that a) control rod drop accidents are not sufficiently sensitive to reductions in scram reactivity to be affected by small boron losses before the end of the blade's design life; b).there is a negligible effect on transient CPR reduction and MCPR limits for small boron lot;es ; but c) if any control blades have experienced more than 10% reduction in projected worth, the shutdown margin should be domonstrated (by testing) to be adequate .
The NRC' has reviewed the bases for GE's ccnclusions , in-cluding the hot cell examinations and calculational assu=ptions ,
and decided that the relationship between boron loss and B-10 depletion was sufficiently well understood tocertain justify actions BWR were operation on an interim basis provided that taken by the licensees. Those actions , which include further
-- analyses , shutdown margin tes ts . and destructive examinations ,
are described in detail in IE Bulletin No. 79-26, Re vision 1, and written responses are required of all operating BWR plants.
Unfortunately , Bulletin 79-26, Rev. 1 does not yet apply to Shoreham since the plant does not have an operating license. ,
1-1 O
__g-.-- .- , -y , . , _ - . . .----.4 ---.~,-,--r . , - - - --r-ry---ry-.-,,-,,...-.%-.,-,-..,,cy%--.--.,,..c.., , wy 9y .,w m..-- , --y--., ,-- - -
It has been assumed by the NRC in their licensing review : hat Shoreham and tha: assurance will these actions wil'. be taken atcon:rol blade reactivi:y will not be signifi-be main:ained : hat LILCO cantly degraded by control blade stress corrosion cracking.
has made no documented commi: rent on either the interim or long
- erm actions specified in Bulletin 79-26 so there is no assurance that Shoreham procedures will reflect ene appropriate actions nor that any icng term solution to the problem is likely or even The concern is exacerbated by LILCO's apparent con template d.
refusal to make imp rovements to the Shoreham backup shutdown capabilities tha: would be provided by NRC's recommended ATWS fix, Alternative 3A.
Despite the relative newness of the communication of thisIt this is not a new is s ue .
concern by was brought to thethe NRC (1979 Bulletin) attention of the NRC, the ACRS , and the t Joint (Congress ional) Committee on Atomic Energy in tes timony presented by Bridenbaugh, Hubbard and Minor on February 18,of the heari 1976. That tes timony , a trans cript industry and NRC responses are published in a two-volume docu-ment by the 94th Congress entitled:
"Inves tigation of Charges Relating to Nuclear Reactor Safe ty ," Volumes 1 & 2, February 18, 23 and 24 and March 2 and 4, 1976.
Pages 752-754 contain the NRC's responses to concerns expressed over BWR centrol rod life and the (as then) uninves tigated The NRC there problem of tube cracking and boron leaching.
s tate d :
"The s taff (NRC) now requir'e's current generation plants and operating reactors to consider the consequences of a complete f ailure of all rods to 1.nsert upon signal. . . . . ." and, " Plants under
. cons truction must incorporate appropriate design features to make the potential consequences of ATWS acceptable."
While boron depletion through tube cracking is in itself unlikely to initiate an ATWS- type event , it does degrade the margins of s afe ty. This , coupled with LILCO's resis tance to the implemen-committed to
-- tation -of the NRC's ATWS fix the U.S. Congress by the NRC in 1976) (which was in effectgives rise to subs tantial concern over the maintenance of transient and long term shutdown capability at Shoreham.
1-2 E
y w ... ,~e - - ,m -wgm~n ,-w-w ~w-m- -,
w - - --
i.
I Supplementary information in response to the detailed ques tions contained in LILCO interrogatories follows J.1: Yes.
J.2: Criterion 4 requires that "coLeonents important to safety shall be designed to accomodate the effects of and to be 'co=patible with the environ-mental conditions associated with normal operation,
....." The tubes centaining the Bt C in control blades do not appear to meet this requirement.
Criteria 21 and 22 require that reactor protection sys tems be designe d fer high reliability and in-se rvi ce tes tability and for indeoendence from the effects of normal operating conditions. The common-mode f ailure of the control blade tubes resulting from exposure of the tubes. co the normal operating environment jeopardites both of these Criteria in the Shoreham reactor. The control blades must be considered a common stension of the reactor pro-tection system. If the blades are non-functional, the protection sys tem is non-functional. A reliable ,
in-service testing feature is not provided. The only way that blade cracks can be detected is by (remote) visual inspection, limited to times when the reactor is shutdown for refueling and mainte-nance. Rod worth tes ting does not provide indica-tion of trouble until after. the failure has occurred.
See introductory paragraphs above. LILCO has not J.3:
committed to any action , interim or long term, so there is no way to judge adequacy.
J.4: 'See_ response to 3.
J.5: a. Facts on which SOC now relies are contained in the introductory paragraphs.
- b. SOC expects to ob tain further information through the interrogatory process from both LILCO and the NRC.
J.6: a. Documents now include - .
(1) Shoreham FS AR, Volumes 1 6 7.
(2) Shoreham SER, Section 4.
J-3
-- - _-w.-.-,. - - - . - - , , - . - y--, , , - ,-- - - . - , - - e-, m--,----e--, - , - , , , -
. . l I j l
l
- (3) NRC Bulletin 79-26 and revisions.
(4) JCAE Hearing Record on Inves tigation of charsc. Relating to Nuclear Reactor Safety, Appendix 5 and Appendix 11.
- b. These documents have not yet been identified.
J.7: All documents identidied in All 6 aare
& breadily should available already be in LILCO's possession.
in the public domain. A copy of the title page of the JCAE Hearing Record is attached for reference purposes in the event that document is not now in LILCO's hands . ( Attachment J-1) To our knowledge ,
none of them have been destroyed.
J.8 J.9 SOC has not yet vccided which witnesses it will uti-J.10: When this decision is made ,
li:e in the hearini.
' the response to these interrogatories will be supple-mented with the requested information. To the extent that SOC's present consultants are assis ting in re-viewing and/or responding to these interrogatories SOC first engaged ,
their resumes are attached hereto.
the services of their consultants , MHB Technical Associates , in December, 1979 and has had a continu-ing discussion with them regarding the contentions in this case.
However, as of this writing, the consul-cants have not made any reports to SOC related to this contention.
J.ll: No.
J.12: N/A.
J.13: Yes.
J.14: See response to ques tion 2.
J .15 : Yes.
J.16: See introductory paragraphs above and response to ques tion 2. .
3-4 e
,- - - - - - - -..n., , - - - - _ , - , , , - . , , a . ,--, -_. .. , -
ATTACH!ENT J-l INVESTIGATION OF CHARGES RELATING TO NUCLEAR REACTOR SAFETY
_,--------,-7 - - - - - - . _,,__n, -
HEARINGS IICFolt! Tile JOINT CO.UllTTEE ON ATOMIC ENERGY CONGRESS OF THE UNITEI) STATES
~
NINETY FOl'IiTII CONGI ESS SIXONLt 81:5910N F1 ltitt*AltY 13. 03. ANil .*4. ANI ilal:c!! :: ANi> 4.19 0 Volume 1: Hearings and Appendises 1-11 P.inted for the use of the Joint Corumittee on Atorriic E:ier:y '
_. -.~. --
1.'.5 COVERN3t ENT Ph!NTING OFFICC C C- C!.9 W ASHINGTON : 19*6 Tor sale h.e the Superintendent of Documetees. t' ?. r . rument Prtatter Of':ee
% a hus;te.n. D C. ::'e str.* - I'rie r ?'a !.ee J-5
' ' ~
9
?00R [R8NAL
K. SOC CONTESTION li The proble= of clad swell or rupture during conditions of high temperature, high differential At the ECCS pressures, rulemaking and/or accidents hearings in has been known for years.
1973, fuel cladding embrittlement and rupture was discussed and the validity of a simple fuel temperature limit was chal-lenge d. However, the Appendix K evaluation models were adopted with the following requirements for acceptability.
"Each evaluation model shall include a provision for predicting cladding swelling and rupture from consideration of the axial temoerature dis tribution of the cladding and from the difference in pressure between the inside and outside of the cladding, both as functions of time . To be acceptable , the swelling and rupture calculations shall be based on acclicable data in such a way th at the decree of swelling and incidence of, rup ture are not un de rp re di c te d . " x (emphasis added)
Unfortunately, the accepted model used by GE and other reactor vendors was based on limited data and in f act may underpredict the likelihood of swell or rupture of fuel cladding under acci-dent conditions.
In an effort to resolve some of the uncertainty in the pre-diction of fuel behavior, the NRC's Office of Nuclear Regulatory Research (NRR) established a research program on zircalloy fuel behavior. TheThe draftresults resultsshowed of theirthat effort were documented in the behavior of zircalloy NURE G- 0 6 30. **
fuel at high stress conditions was not the same as predicted in the models used by GE and others ; in fact, the models under-predicted in several cases.
line The GE mddel for clad rupture was a simplified straight This approximation with a slight increase at about 9 2 5 oF. ***
was 'relatively satis factory for an assumed slow ramp but sub-stantially underpredicts the results for a f as t ramp condition.
. ._10 _CFR 50 , App en dix K , Se ction I . B .
- MUREG-0630, Figures 55 & 56.
K-1 m- - .,ee- . - - - , . - - - - - , .- - , - . _ --
, --- ,-- 4 , , , , , , , ,. % -,,--. ---,--,-...,,-r-,-.--..v,.,
I The implicatiens are that the fuel could experience core swelling, de formation, and rupture due to f as t-ramp or ace'-
dent condition and thus change the fuel configuration. De gra-dation of the fuel geometry could adversely i= pact the ability of the ECCS to cool the core during an accident. This issue is listed as an open item in the SER,* requiring additional calculations of E'CCS performance assuming the models in NUREG-0630. As it presentty s tands , the Shoreha= plant does not appear to comply with the requirements of 10 Cf n ~0. 46 an d -
Appendix K.
Supplemental responses are as follows :
K.1: SOC presently has no evidence that models other than the models identified in the FSAR (Section 6.3.3.4) and the SER (Section 6.3.2) were used by the Applicant in analyzing the ECCS performance .
K.2: See general discussion above.
K.3: See general discussion above.
K.4: Se'e general discussion above .
K.5: Not at this time .
K.6: There are additional analyses required in the SER and there will be interrogatories on this subject.
K.7: N/A.
K.8: See above .
K.9 K.10 : See above .
K.ll K.12 K.13: SOC has not yet decided When which witnesses it will uti-this decision is made , the lize in the hearings.
response to these interrogatories will be supplemented with the reques ted information. To the extent that
_. SOC's present consultants are assis ting in reviewing and/or responding to these interrogatories SOC first engaged , their the resumes are attached hereto.
- SER, NUREG-0420, page 1-7, Item 11.
K-2
_ _ _ , . , _ _ _ - - - - , . , . , 7 .., , . . . _ . . , _- _ ,,_,_,,my .,_m,__.- . , , _ - . . _ _ . _ _ _ _ _ _ _ . . _ _ _ _ , _ . . - . ,
l services of their consultants , MHB Technical' Asso-ciates , in Dece=ber,. 1979 and har had a continuing discussion with them regarding tha contentions in this case.
However, as of this vriting. the consul-
' tants have not made any reports to SCC related to this contention.
K.14: No.
K.15: N/A.
K.16 K.17: See general discussion above.
K.18 K.19: See general discussion above . Also , to the extent that the present Shoreham design does not meet the requirements of 10 CFR 40.46 and Appendix K,Section I .B , and there is an open item in the SER on this issue, the Applicant has not " demons trated" the safety of the Shoreham design.
O.
- = m e my e, K-3 m e
- . - _ _ _ _ _ , ' - ~ ' " - ~ * ' + - - .%_,,_ m ,,, _ _
.L. SOC CONTENTION 17 The TMI-2 accident involved opera:cr actions which resul:ed in shutting off safety systems while they were still needed to perform their function. This has called further a::ention to a regulation which is not fully complied with on the Shoreham de-sign. The Regulation is 10 CFR 50.55a(h) which calls for plants with Construction Permits issued after January 1, 1971 (Shore-ham CP was issued April 14, 1973), to comply with IEEE-279* for protection system functions.
The Shoreham FSAR, Section 7.1.1, claims compliance with the 1971 version of IEEE-279, including the manual switches, by-passes, and logic for the reactor scram system, ECCS, NSSSS and several other systems. However, there is a require =ent in IEEE-279 that says that the " protective system shall be so de-signed that, once initiated, a protective sys:em action shall go to completion."** ,
The safety function of many of the systems listed as com-plying with IEEE-279 is either manually- initiated, as in the case of the standby liquid control, or can be manually inter-rupted as in the case of ECCS pu=ps , va'ves , etc. It is this very ac: ion tha: prevented the - *: rect operation of the safety functions at TMI-2. The same proolem is a potential occurrence at Shoreham.
The FSAR states that the N: JS is designed to go to com-pletion but allows for the operator to reverse it:
"The system has been designed so that, once initiated, automatic isolation action goes to co=pletion. Return to normal operation after isolation action requires deliberate operator action."***
The ECCS.section is even less definitive and makes no state-ment of a safety design basis which requires the system function to go to completion.* *
- The regulations changed within the last few years to call for the version of IEEE-279 in effect at the docketing date.
_ .I.or_.Shoreham, being a very old design, this precedes the is-suance of IEEE-279 by 3 months (docket date is May 15, 1968, and IEEE-279 effective date is Aug. 30, 1968). Given tha:
almost five years elapsed from docketing to CP ther6 is no l with IEEE-279.
justification for Shoreham not comply .ng
- IEEE-279-1971, section 4.16.
- FSAR, Section 7.1.2.1.2.
- FSAR, Section 7.1.2.1.3.
L-1 l
,. , ,- -. - -- - - , -- - - - - , - - - , .,-vv.
The Shoreham design should be thoroughly reviewed to docu-ment the extent to which safety functions are not required to go to completion. To the extent that these safety functions do not go to completion or are' reversible by operator action, the design does not comply with the intent of 10 CFR 50.55a(h)'.
Supplemental responses are as follows .-
L.1: See the above.
L 2: See the above.
L.3: This refers to the safety function of a particu-lar system being allowed to be completed and not being interruptible or prevented by operator ac-tion once initiated.
L.4 Modifications as required to correct deficiencies and in complianca with 10 CFR 50.55a(h).
L.5:
L.6: See the general discussion above.
L.7: See the general discussion above.
L.8, At this time, SOC has not decided which witnesses it L.9, will utilize in the hearings. When this decision is L.10, made, the response to these interrogatories will be and supplemented with the requested information. To the L.ll: extent that SOC's present consultants, MHB Technical Associates, are assisting in reviewing and/or respond-ing to these interrogatories, their resumes are at-tached. SOC first engaged the services of their con-sultants in December, 1979, and has had a continuing
. discussion with them regarding the contentions in this case. However, as of this writing, the consul-tants have not made any reports to SOC related to this contention.
L.12: No.
L.13: N/A 3
L.14: There is insufficient documentation in the FSAR to draw conclusions regarding specific systems.
- There is a need for a thorough review of the sys-tems before such a determination can be made.
L-2 9
-,- ,, r-- -,-----,r--,w - - - - , - , - - , - m- e- - , - - - -,-v - ,. - , - - , - - , - w n- + ~
M. SOC CON ~INTION 19 The Atomic Safety and Licensing Board (Board) has not yet ruled on the scope of proposed Contention 19 as submitted by
. SOC to the Board on March 18, 1981. Pending the Board's ruling and the resumption of SOC discovery related to this contention.
SOC respectfully declines to specifically respond to this in-
! terrogatory at this tJme.
During 1980 and 981 a series of meetings was held be-tween SOC, LILCO, and the NRC in a joint effort to particular-ice the technical issues encompassed by Contention 19.* Sub-scantial progress was made in the informal discovery process in reducing the issues in Contention 19 as noted by SOC at Page 3 of the March 18 motion as follows:
". . .of the 145 Regulatory Guides subj ect to review and informal Discovery by the parties, approximate-
- ly 100 of those Regulatory Guides have been resolv-ed to the satisfaction of each party. SOC fully expects that additional Regulatory Guides from the 46 that remain will be reviewed and resolved prior
.. to hearings in the informal discovery process on this issue. Thus, discussion among the parties has already substantially contributed to answering the Board's question of whether the standards or goals of recent Regulatory Guides have been met."
It should also be noted that the informal discovery process on Contention 19 was terminated at LILCO's request. SOC under-stood that LILCO wished the termination to enable the Board to define the scope of the contention. Since the Board decision is still pending, and since SOC has been willing to hold additional discovery in abeyance pending the. Board's ruling,- SOC believes that response to LILCO's fifteen interrogatories is not appro-priate or warranted at this time. Accordingly, SOC objects to the interrogatorie.s submitted by the Applicant regarding SOC's proposed contention 19 on the ground that they are premature.
These and future interrogatories submitted by Applicant or other parties will be responded to in timely fashion af ter the Board has ruled on SOC's March 18, 1981 Motion.
- For example, see LILCO's minutes of meeting's held on July 17
- and September 12, 1980, and January 22, 1981, to particularite the issues in Contention 19.
M-1 9
-- -. - - . ~ . . . , . , - - . ...,~.,.-.-,7.-w..-..-. -,---....--..---y- -,-..--,_-.% , ..y-....-n._,m,
- . - , ,c._,-v...,,,,..c--,-,.y,-. - -
_ w. __,, m e,a- .A-- A - -- 1 ._ - . - m a,-w a -n_-_., , - ,,_,,., .- - ~ , sn- -- _+ ,
i em 1
6 4
1 1
1 i
6 1
- e 9
e i.
f t.
1 i
l.
@ h ee S
I
. ..e 1
l I
l
~
---rer-*+-=r- --.v,. , , _ . . . _,, ._ _ , __ _
'M *Or 'W~ -w~.-map-,yw,_,w_., , , _
- "O'""""*"TW"v4'we-*W-my y.
. 1; PROFESSIONAL QUALIFICATIONS OF DALE G. B RI DENB AUGH DALE G. B RI DENB AUGH 1723 Hamilton Avenue ,
Suite K ,
San Jose, CA 95125 (408) 266-2716 EXPE RIENCE : .
19 76 - P RESENT Prssident - MHB Technical Associates, San Jose, California.
CoS founder and partner of technical consulting firm. S p e c ia lis't s" in energy consulting to governmental and other groups interested in evaluation of nuclear plant safety and licensing. Consultant in this capacity to sr. ate agencies in California, New York, Illi-nois, New Jersey, Pennsylvania, Oklahoma and Minnesota and to the Norwegian Nuclear Power Ceemittee, Swedish Nuclear Inspectorate, and various other organizatters and environmental groups. Per-formed extensive safety analysis for Swedish Energy Commission and contributed to the Union of Concerned Scientist's Review of W ASH-14 00. Consultant to the U.S. NRC - LWR S af ety Improvement Program, performed Cost Analysis of Spent Fuel Disposal for the Natural Resources Def ense Council, and contributed to the Depart ,-
ment of Energy LWR S af ety Improvement Program for Sandia Labora-tories. Served as expert witness in NRC
and state utility commission hearings.
1976 - ( FEB RUA RY -
AUGUU T)
Consultant, Project Survival, Palo Alto, California. _
Volun teer wo rk on Nucl. ear S a f eguard s Initiative campaigns in l
! California, Oregon, W a s h in g t on , Arizona, and Cclorado. Numerous presentations on nucicer power and a-lternative energy options to civic, government, and callege groups. Also resource person for public service presentatioes on radio and television.
1977 - T976 -
Manager, Performance Evaluation and Improvement, Ceneral Electric Company - Nuclear Energy Division. San Jose, California.
F l
Managed seventeen technical and seven clerical personnel with
! responsibility for establishment and management of systems to monitor and measure Boiling Water Reactor equipment and system operational performance. Integrated General Electric resources in customer plant modifications, c oo rd in a te d correction of causes of forced outages and of efforts to improve reliability.and per-formance of BWR systems. -
,~ .-,... - ,-.- .. --,, , --- . . .-- . - - . . , . - , . - - . - _ . - . . - - - - - - - --
I 1973 -
1976 ( Con t d)
Responsible for development of Division Master Performance Improvement Plan as well as for numerous Staff special assign-ments on long-range studies. Was on special assignment for the management of two different ad hoc projects formed to resolve unique technical problems.
1972 - 1973 Manager, Product Service, General
- Electric Company - Nuclear Energy Division, S an Jose, California.
Managed group of twenty-one technical and f our clerical personnel.
Prime responsibility was to direct interface and liaison personnel involved in corrective actions required under contract warranties.
Also in charge of refueling and service planning, performance analysis, and service communication functions supporting all com-~
pleted commercial' nuclear power reactors supplied by General Electric, both domestic and overseas (Spain, Germany, Italy, Japan, India, and Switzerland) .
1968 - 1972 Manager, Product Service, General Electric Company - Nuclear Energy
- Division, S an Jose , California.
Managed sixteen technical and six clerical personnel with the r e s pon s ib ility for all customer contact, plann in g and execution of work required after the customer acceptance of department-supplied plants and/or equipment. This included quota tion , sale and delivery of spare and renewal parts... S ales volume of parts
- increased from $1,000,000 in , 968 to over $3,000,000 in 1972.
1966 - 1968 Manager, Complaint and Warranty S e rvic e , General Electric Company -
Nuclear Energy Division, S an Jose, California.
Managed group of six persons with the responsibility for customer contacts, planning and execution of work required after customer acceptance of department-supplied plants and/or equipment--both domestic and overseas.
1963 - 1966 Field Engineering Supervisor, General Electric Company, Installation and Service En z ine e rin g Department, Los Angeles, California.
Supervised approximately eight field representatives with responsi-bility f or General Electric s team and gas turbine installation and main tenance work in S outhern Calif ornia, Arizona, and Southern Nevada. During this period was res pons ible for the installation of eight different central station steam tu rb in e gene rator units , plus much =aintenance activity. Work included customer contact, prepa-ration of quotations, and contract negotiations.
- * - - ' ~ ~ I - ,- -
y , , , _ _ _
1956 - 1963 Field En g inee r. General Electric Company, Installation and Service Engineering Department, Chicago, Illinois.
Supervised installation and main ten anc e of steam turbines of all s ize s . Supervised crews of from ten to more than one hundred men, depending on the job. **orked
- primarily with large utilities but had significant work with steel, petroleum and other process industries. Had four years of experience at construction, startup, trouble-shooting and refueling of the first large-scale commercial nuclear power unit.
1955 - 1956 En g in e e r in g Training Program, General Electric Company, Erie, Pennsylvania, and Schenectady, New York. .
Training assignments in plant facilities design and in steam turbine testing at two General Electric Factory locations.
1953 - 1955 United S tates Army - Ordnance School, Aberdeen, Maryland.
Instructor - Heavy Artillery Repair. Taught classroom and shop disassembly of artillery pieces.
1953 Engineering Training Program, General El'e'etric Company, Evendale, Ohio.
Training assignment with Aircraft Gas Turbine Department.
EDUCATION & AFFILI ATIONS :
BSME - 1953, South Dakota S chool of Mines and Technology, Rapid City, South Dakota, Upper k of class.
Erofzas.ional Nuclear Engineer - California. Certificate No. 0973.
Member - American Nuclear Society.
Various Company Training Courses during career including P ro f es-s ional Bus ine s s Mana gement , Kepner Tregoe Decision Making, Effective Presentation, and numerous technical seminars.
9
. - - , - - - . . , _ . --..-.n , , _ . , - - . . , . - - - - ,,__ ,. - --n. - , . , , , ,
HONORS & Ak' A RDS :
Sigma Tau - Honorary Engineering Fraternity.
General Managers Award, General Electric Company.
Born November 20, 1931, Miller. South Dakota.
Married, three children 6'2", 190 lbs., health - excellent Honorable discharge f rom United S tates Army Hobbies: Skiiing, hiking, work with Cub and Boy Scout Groups.
P UB LIC ATION S & TESTIMONY :
- 1. Operating and Maintenance Experience, presented at Twelfth Annual Seminar for Electric Utility Executives, Pebble Beach, California, October 1972, published in General Electric NEDC-10697, December 1972.
- 2. Maintenance and In-Service Inspec tion , presented at IAEA Symposium on Experience From Operating and Fueling of Nuclear Power Plan ts , B r.idenbaugh, Lloyd & Turner, Vienna, Austria, October, 1973. ..
- 3. Operating and Maintenance Experience, presented at Thirteenth Annual Seminar for Electric Utility Executives, Pebble Beach, California, November, 1973,' published in General Electric NEDO-20222, January. 1974
- 4. Improving Plant Availab ility , presented at Thirteenth Annual S eminar f or Electric Utility Executives, Pebble Beach, Cali-f ornia, November 1973, published in General Electric NEDO-r 20222, January, 1974
- 5. Application of Plant Outage Experience to Improve Plant Per-
~~
~ fo'rmance, B ridenbaugh and Burds all, American Power Conference, Chicago, Illinois, April 14, 1974 i
- 6. Nuclear Valve Testing Cuts Cost, T,i m e , Electrical World.
October, 15, 1974.
l The Risks of Nuclear Power Reactoes: A Review of the NRC i
l 7.
Reactor S af ety Study VASH-1400, Kendall, Hubbard, Minor &
Bridenbaugh, et al, for the Union of Concerned Scientists, August, 1977.
l
(
l
Swedish Reac tor ,S af e tv Study: Barseba'ck Risk Assessment, 8.
MHB Technical Associates, January, 1978. (Published by the Swedish Department of Industry as Document Ds1 1978:1)
- 9. Testimony of D.G. Bridenbaugh. R.B. Hubbard, C.C. Minor to the Calif ornia S tate As sembly Committee on Resources, Land Use, and Energy, March 8, 1976.
- 10. Testimony of D.C. Bridenbaugh, R.B. Hubbard, and G.C. Minor before the United S ta tes Congress, Joint Committee on A t om i e.
Energy, February 18, 1976, Washington, DC (Published by the Union of Concerned S cientis ts , Cambridge, Massachusetts.)
- 11. Testimony by D.G. Bridenbaugh before the California Energy Commission, entitled, Initiation of Catastrophic Accidents at Diablo Canyon, Hearings on Emergency Planning, Avila Beach, California, November 4, 1976.
- 12. Testimony by D.G. Bridenbaugh before the U.S. Nuclear Regula-tory Commission, subject: Diablo Canyon Nuclear Plant Perfor-mance, Atomic S af ety and Licensing Board Hearings, December, 1976.
- 13. Testimony by D.G. '. cide nbaugh be f o re the California Energy Commission, subject: Interim Spent Fuel S torale Considerations, March 10, 1977.
- 14. Testimony by D.G. Bridenbaugh before the New York S tate Public Service Commission Siting Board Hearings concerning the James-port Nuclear Power S tation , subject: Cost Effect of Technical and and Reliability, Safety Deficiencies on Nuclear Plan _t April, 1977.
- 15. Testimony by D.G. Bridenbaugh before the Calif ornia S tate Energy Commission, subject: Decommissioning of Pres surized Water Reactors, Sundesert Nuclear Plan t Hearings, June 9, 1977.
- 16. Testimony by D.G. Bridenbaugh before the Calif ornia S tate Energy Commission, subject: Economic Relationships of Decommissiot.ing, Sundesert Nuclear Plant, for the Natural Resources Defense Council, July 15, 1977.
- 17. Test mony by D.C. B ridenbaugh bef ore the Vermont State Board of Health, subj ect : Operation of Vermont Yankee Nuclear Plant and It s Impac t on Public Health and Safety, October 6, 1977.
c 18. Testimony by D.C. Bridenbaugh bef ore the U.S. Nuclear Regula-tory Commission, Atomic S af ety and Licens ing B oard , su bj ec t :
Deficiencies in Safety Evaluation of Non-Seismic Issues, Lack of a Definitive Finding of Safety, Diablo Canyon Nuclear Units October 18, 1977, Avila Be ach, California.
' - I . . . . .
l l
I Testimony by D.G. Bridenbaugh before the Norwegian Commission 19.
on Nuclear Power, subject: Reactor S af etv /Ris k . October 26, 1977.
Testincny by D.C. Bridenbaugh before the Louisiana S tate 20.
Legislature Committee on Natural Resources, subject: Nuclear Power Plant Deficiencies Impacting on Safety & Reliability, j
' Baton Rouge, Louisiana, February 13, 1978.
3ridenbaugh
- 21. Cpent Fuel Dispos,al Costs,Defense report prepared by 2.0, Council (NRDC), August 31, for the Natural Resources t
1978.
- 22. Testimony by D.G. B ridenbaugh, G.C. Minor, and R.B. Hubbard before the Atomic Saf ety and Licensing Board, in the matter of the Black Fox Nucleat Power S tation Construction Permit Hearings, September 25, 1978, Tulsa, Oklahoma,
- 23. Testimony of D.G. B ridenbaugh and R.B. Hubbard before the Louisiana Public Service Commission, Nuclear Plant Louisiada7 and Poser Generation Costs, November 19, 197 8, B aton Rouge, Testimony by D.C. Bridenbaugh before the City Council and 24 Electric Utility Commission of Austin, Texas, Sesign. Con-struction, and Operating Experience of Nuclear Generating Facilities, December 5, 1978, Austin, Texas.
- 25. Testimony by D.G. Bridenbaugh for the Commonwealth of Massachusetts, Department of Public Utilities, Impact of Unresolved Safety _ Issues. Generic Deficiencies,_and_Three i
l . Mile Island-Initiated Modifications en Power June Generation 8, 1979, Cost at the P roposed Pilgrim-2 Nuclear Plant,
- 26. Improving the 3afety o f LW R -P ower P l an t s , MHB Technical
' Associates, prepared for U.S. Dept. of Energy, Sandia Laboratories, September 28, 1979.
- 27. BWR Pipe and Nozzle Cracks, MHB Te chnical As so cia tes , for the Swedish Nuclear Power Inspectorate (SKI), October, 1979.
- 28. Testimony of D.G. Bridenbaugh an d C . C . Minor before the Atomic Safety and Licen s in g B oard , in the matter of i ~' ~ Sarramen to Municipal Ut'ility Dis tric t , Rancho Seco Nuclear Generating S ta tion f ollowin g TMI-2 accident, subject:
Operator T r a in in g and Human Factors En g in e e r in g . for the California Energy Commission, February 11, 1980.
Italian Reactor S af ety Study: Caorso. Risk Assessment. MHB 29.
Technical As sociates , for Friends of the Earth, Italy, March, 1980.
- 30. Decontamination of Krvpton-85 from Three Mile Island Nuclear, B r id en b au gh , et al, R. Pollard, & D.G.
Plant, H. Kendall, to the G'overnor The Union of Con cerned S cien tis ts , delive red of P ennsylvania , May 15, 1980.
F
,-,,,,w - -, , - . - -
-n---l ,ne,-v. ,,.---.-n---rnw-,--, , _ . , , , - , - ,--_m-.------,--.---- ,~-..,-.,--,,r-,,,m-,,..,, ,r--- .--------or
- 31. Testimony by D'.G . B ridenb augh b ef o re the NPublic ew Jersey Board of Advocate's Public Utilities, on behalf of New Jersey An a ly s is of 1979 Salem-1 O f f :.c e , Pivision of Rate Counsel, Refueling Outage, August, 1980.
P osition S tatement , Proposed Rulemaking on the Storage and Dispos _al o f Nuclear Was te, Joint Cross-S to tement of Position 3 7. .
of the New Englans Coalition on Nuclear Pollution and the Natural Resources Defense Council, September, 1980.
C. Minor, before
- 33. Testimony by D.G. Bridenbaugh and Gregory Commission, In the Matter the New Island York SLightingtate Public Service Rate Case, prepared of Long Company Tempo rary Coalition, September 22,'1980, for the Shoreham Opponents Shoreham Nuclear Plant Cons truction S chedule.
D.G. Bridenbaugh before the New 34 Supplemental Testimony by on behalf of New Jersey Jersey Boad of Public Utilities, Division of Rate Counsel, An aly s is Public Advocate's Office, of 1979 Salem-1 Refueling Outage, December, 1980.
. e 9
= * . .. . ,,,
d I
., e --e-- - - , , , - . ~w r, -e,---r , .,,,-,v -e a g--- - - - - - r~~~ .-m ,
- 9
! i PROFESSION AL QUALIFICATIONS OF RI CH ARD B . H UB B ARD RICHARD 3. H UBB A RD MHB Technical Associates 1723 Hamilton Avenue Suite K -
San Jose, California 95125 (408) 266-2716 E XP E RI EN CE :
9 / 76 - P RESENT Vice-President - MHB Technical Associates, San Jose, California.
Founder, and Vice-President of technical consulting firm. Special-is ts in independent energy assessments for government agencies, particularly technical and economic evaluation of nuclear power facilities. Consultant in this capacity to Oklahoma and Illinois Attorney Generals, Minnesota Pollution Control Agency, German Ministry f or Research and Technology, Governor of Colorado, Swedish Ener;y Commission, Swedish Nuclear Ins pectorate, and the U.S.
Department of Energy. Also provided studies and testimony for various public interest groups including the Center for Law in the Public Interest, Los Angeles; Pu blic Law Utility Group, Baton Rouge, Louisiana; Friends of the Earth (F0E), Italy; and the Union of Concerned S cien tis ts , C a mb.r i d g e , Massachusetts.
Provided testimony to the U.S. Senate / House Joint Committee on Atomic Energy, the U.S. House Committee on In te rior and Insular Affairs, the Calif o rnia Assembly, Land Use, and Energy Committee, the Advisory Committee on Reac tor S af e guards, and the Atomic S af ety e and Licensing Board. Performed comprehensive risk analysis of the accident p rob (b ilit ie s and consequences at the B ars eback Nuclear Plant for the Swedish Energy Commission and edited, as well as contributed to, the Union of Concerned Scientist's technical review of the N RC's Reactor S af ety Study (WASH-1400) .
l 2/76_- 9/76 _
Consultant, Project Survival, Palo Alto, California.
Volunteer vork on Nuclear Saf eguards I ni t ia t i ve campaigns in Cali-fornia, Oregon, Washington, Arizona, and Colorado. Numerous I presentations on nuclear power and alternative energy options to
! civic, government, and college groups. Also resource person for public service presentations on radio and television.
I
.. ,- , . , - - , - , . , - . , . , ., - , - - - - , , , , . . ~
l 5/75 - 1/76 i
Manager - Quality Assurance Section, Nuclear Energy Control and instrumentation Department, General Electric Company, San Jose, California.
Report to the Department General Manager. Develop and implement quality plans, programs, methods, and equipment which assure that products produced by the Department meet quality requirements as defined in NRC regulation 10 CF,R 50, Appendix 3, ASME Boiler and Pressure %ssel Code , customer contracts, and CE Corporate policies and procedures. Product areas include radiation sensors, reactor vessel internals, fuel handling and servicing tools, nuclear plant control and p ro t ec t io n instrumentation systems, and nuclear s team supply and Balance of Plant control room panels. Responsible for approximately 45 exempt personnel, 22 non-exempt personnel, and 129 hourly personnel with an expense budget of nearly 4 million dollars and equipment investment budget of approximately 1.2 million dollars.
11/71 - 5/75 Manager - Quality Assurance Subsection, Manufacturing S ection of Atomic Power Equipmen t Department, General Electric Company, San Jose, Calif o rnia .
Report to the Manager of. Manufacturing. Same functional and product responsibilities as in Engcgement #1, except at a lower organizational report level. Developed a quality system which received NRC certification inand 1975. The system was also success-
"NPT" symbol authorization in 1972 fully surveyed f o r AS ME "N" a'nd 1975, plus AS ME "U" and "S" symbol. authorizations in 1975.
Responsible for from 23 to 39 exempt personnel, 7 to 14 non-exempt personnel, and 53 to 97 hourly personnel.
3/70 11/71 Manager - Application Engineering Subsection, Nuclear Instrumen-tation Department, General Electric Company, San Jose, California.
Responsible for the post order technical in ter f ace with architect engineers and power plant owners to define and schedule the instru-mentation and con t ro l sy s ter.s for the Nuclear S team Supply and
-- B a lam c e- o f P l a n t portion of nuclear power generating stations.
Re s p on s ib ilit ie s included preparation of the plant instrument list with approximate location, review of interface drawings to define functional design requirements, and release of functional require-ments for detailed equipment designs. P e rs onnel supervised inc lud e d 17 engineers and 5 non-exempt parsonnel.
S n+ , - . - - - e. s,--, -p- - , - - - - - ,,-~w ,
-wa ,e.- --p w--r - -
12/69 - 3/70 Chairman - Equipment Room Task Force, Nucle ar Ins trumen tation Department, General Electric Company, San Jose, California.
Responsible for a special task force reporting to the Department General Manager to define methods to improve the quality and reduce the installation time and cos t of nuclear power plant control rooms. S tudy resulted in the conception of a factory-f ab ricated control room consis ting of signal conditioning and operator control panels mounted on modular floor sections which are completely assembled in the factory and thoroughly Personnel tested s up e rv is ed for proper operation of interacting devices.
included 10 exempt personnel.
12/65 - 12/69 Manager - Proposal Engineering Subsection, Nuclear Instrumentation Department, General Electric Company, San Jose, California.
Responsible for the application of instrumentation systems for nuclear power reactors during the proposal and pre-order period.
Responsible for technical review of bid specifications, p r ep a ra tio n of technical bid clarifications and exceptions, definition of material list for cost estimating, and the "as sold" review of Personnel contracts prior to turnover to Application Enginee ring.
supervised varied from 2 to 9 engineers.
8/64 - 12/65 Sales Engineer, Nuclear Electronics B us ine s s Section of Atomic Power Equipment Department, General Electric Company, San Jose, California.
Responsible for the bid review, contract negotiation, and sale of ins trumen a tion systems and components f or nuclear power plants, test r e a c t o r,s , and radiation hot cells. Also r e s po n s ib le for indus trial sales of radiation sensing systems for measurement of chemical properties, level, and density.
10/61 - 8/64
--- A pp lic at i o n Engineer, Low. Voltage Switchgear Department, General Electric Company, Philadelphia, Pennsylvania.
Responsible for the application and design of advanced diode and s il ic o n- c o n t r o lled rectifier constant voltage DC power systems and variable voltage DC power systems f o r indus trial ap plications .
Designed, followed manuf acturing and personally tested an advanced SCR power supply fo r product introduction at the Iron and S teel Show.
Proiect Engineer for a DC power system for an aluminu= pot line so ld to Anaconda beginning at the 161KV switchyard and encompassing all the equipment to convert the power to 700 volts DC at 160,000 amperes
_3_
,n ,, - - , , - - , .~- ,,e
1
. . l 1
l 1
J 9/60 - 10/61 Program CE Rotational Training the GE Ro ta tional Training P rogram Four 3-month ass ignments on for college technical graduates as follows: '
Detroit, Michigan.
Installation and Service Eng.
a.
Installation and startup testing of the world's largest automated hot strip steel mill.
- b. Tes te r - Indus try Con t rol - Ro anoke , Vi r g in i a .
Factory testing of control panels for control of steel, paper, pulp, and utility mills and power plants.
- Johnson
- c. Engineer - Light Military E le c t ronics -
City, New York.
Design of ground su'pport equipment for testing the auto pilots on the F-105, S ales En gineer - Mo rrison, I l l in o i s_.
d.
Sale of appliance controls including range timers and refrigerator cold controls.
University of Arizona, B achelor o f S cience Electrical Enginee ring, 1960.
Administration, University of Santa Clara, 1969.
Mas ter o f Busine ss P RO FES SION AL AFFILI ATION :
State of California.
Registered Quality Engineer, License No. QU805 the Nuclear Power Engineering Committee Member of S ubcommit t ee 8 of Power Engineering Society responsible for the prepara-o f--the-IEEE Standards:
tion and revision of the following 4 national Q . A.
Supplementary Requirements
- a. IEEE 498 (ANSI N 4 5. 2.16) : and Test for the Calibration andConstructior, Control of Measuring and Maintenanco of Equipment used in the Nuclear Power Generating S tations .
t G
.._ , . , , _ . .+ - , , - , - , - - . . ._, , , , - , - -
I PROFESSIONAL AFFILI ATION : ( Con td)
- b. IEEE 336 (ANSI N45.2.4): Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment during the Cons truction o f Nuclear Power Generating S ta tio ns ,
- c. IEEE 467 4 ANSI 45.2.14): Quality Assurance Program a Requirements for the Design and Manufacture of Class IE Instrumentation and electric Equipment for Nuclear Power Genera ting S ta tio ns .
- d. IEEE Draft: Requirements for Replacement Parts for Class IE Equipment Replac emen t ' P ar ts for Nuclear Power Genera ting S ta tions .
Birth Date: 7/08/37 Married; three children Health: Excellent P UB LI C ATIONS AND TESTIMONY:
- 1. In-Core System Provides Continuous Flux Map of Reactor Cores, R .B . Hubbard and C.E. Foreman, Power, November, 1967.
~
- 2. Quality Assurance: Providing It, Proving It, R .B . Hubbard, Power, May, 1972.
- 3. Testimony of R .B . Hubba[d, D.G. B ri d enba u gh , and G.C. Minor before the United States Congress, Joint Committee on Atomic E n e r g y', February 18, 1976, Washington, DC. (Published by the Union of Concerned S cientis ts , Cambridge, Massachusetts.)
~
Excerpts from testimony published in Quote Without Comment, Chemtech, May, 1976.
- 4. Testimony of R .B . Hubbard, D.G. B ridenb augh , and G.C. Minor
__ _ __,to.the California S tate Assembly Committee on Resources, Land Use, and Energy, Sacramento, California, March 8, 1976.
- 5. Testimony of R. B. Hubbard and G.C. Minor before California S tate Senate Committee en Public Utilities, Transit, and Energy Sacramento, California, March 23, 1976.
- 6. Testimony or R.B. Hubbard and G.C. Minor, Judicial Hearings Regarding Grafenrheinfeld Nuclear Plant, March 16 & 17, 1977, Wurzburg, Germany.
e l I
( Con td) ;
PUBLICATIONS AND TESTIMONY:
7.
Testimony of R .B . Hubbard to United '*ates House of .
Rep r es en ta tiv es , Subco=cittee on Energy and the Environ-ment, June 30, 19 7 7, W ashin g ton , DC, entitled, Effectiveness of NRC Regula tions - Modifications to Diab lo Canyon Nuclear Units.
Testimony of R.B.
~
Hubbard to the Advisory Committee on Reactor 8.
Safeguards, August 12, 1977, Washington, DC, entitled, Risk Uncertainty Due to Deficiencies in Diablo Canvon Ouality Assurance Program and Fa ilur e to Implement Current NRC Practices.
- 9. The Risks of Nuclear Power Reactors: A Review of the NRC R.B.
Reactor Safety Study WASH-1400, Kendall, et al, edited by Hubbard and G.C. Minor for the Union of Concerned S cientis ts ,
August, 1977.
- 10. Swedish Reactor Safety Study: Barsebick Risk Assessment, MHB Technical Associates, January 1978 (Published by Swedish Depart-ment of Industry as Document DSI 1.978:1).
- 11. T es t imony o f R.B. Hubbard before the Energy Facility S iting Council, March 31, 1978, in the matter Pebble of Pebble Springs Springs Nuclear Nuclear Plant, Power _ Plant, Risk Assessment:
Portland, Oregon.
- 12. Presentation by R.B. Hubbard before the Federal Minis try for1978, Research and Technology ( B MFT) , August 31 and S ep temb er 1, Meeting on Reactor Safety Research, Risk Analysis, Bonn, Germany.
Mino r T es timony by R .B . Hubbard, D.G. Bridenbaugh, and G.C.
13.
before the Atomic Safety and Licensing Board, S ep t emb er 25, 1978,.
in the matter of the Black Fox Nuclear Power Station Constructinn Permit hearings, Tulsa, Oklahoma.
- 14. Testimony of R.B. Hubbard before the Atomic Safety and Licensing j
Board, November 17, 1978, in the matter of Diablo Canyon Nuclear Power Plant Operating License Hearings, Operating Basis Earth-quake and Seismic Reanalvsis of Structures, Systems, and Com-
- p o n e n t s_ , Avila Beach, California.
_15., ,L'ouisiana of R.B.
TestimonyPublic Hubbard and D.G. B rid e nb a u gh before the j Service Commis s ion, November 19, 1978, Nuclear l
i Plant and Power Generation Costs, Baton Rouge, Louisiana.
- 16. Testimony of R.B. Hubbard bef c re the California Legislature, Subco=mittee on Energy, Los Angeles, April 12, 1979.
_f_
,. _ . , . . -, -r -- -- ,. y - - , - - - ,
, , sI#
/
f PUBLICATIONS AND TESTIMONY: (Contd)
- 17. Testimony of R.B. Hubbard and G.C. Mino r befo re the Federal Trade Commission, on behalf of the Union of Concerned S cientis ts , Standards and Certification Proposed Rule 16 CFR Part 457, May 18, 1979.
~
- 18. ALO-62, Improv1nt the Safety o f LWR Power Plants , MHB Technical Associates, prepared for U .S . Department of Energy, Sandia National Laboratories, September, 1979, available from NTIS.
- 19. Testimony by R.B. Hubbard before the Arizona State L e g is la tu r e ,
Special Interim House Committee on Atomic Energy, Overview of Nuclear Safety, Phoenix, AZ, September 20, 1979.
- 20. "The Role of the Technical Consultant," Practising Law Insti-tute. program on " Nuclear Litigation," New Yo rk City .and Ch'icago ,
November, 1979. Available from PLI, New York City.
- 21. Uncertainty in Nuclear Risk Assessmen t Me tho do lo gy , MHB Technica:
As s o cia tes , January, 1980, prepared for and available from the Swedish Nuclear Power Inspectorate, Stockholm, Sweden.
Italian Reactor Safety Study: Caorso Risk Assessment, MHB 22.
Technical Associates, March, 1980, prepared for and available from Friends of the Earth, Rome, Italy.
- 23. Development of Study Plans: Safety Assessment of Mon ticello and Prairie Island Nuclear Stations, MHB Technical Associates, August, 1980, prepared for and a v,a ila b le from the Minnesota Pollution Control Agency.
- 24. Affidavit of Richard B. Hubbard and Gregory C. Minor before the Illinois Commerce Commission, In the Matter of an Investi-gation of the Plant Construction Program of the Commonwealth Edison Company, prepared for the League of Woman Voters of Rockford, Illino is , November 12, 1980, ICC Case No. 78-0646.
- 25. Systems Interaction and Single Failure Criterion, MHB Tech-nical Associates, November, 1980, prepared for and available from the Swedish Nuclear Power Inspectorate, Stockholm, Sweden.
l I
i
! I 1
- - - - , n.,. y, , - .---,--g,m.c g-- - , . . , ,
.i PROFESSIONAL OUALIFICATIONS OF GREGORY C. MINOR G RE GO RY C. MINOR MHS Technical Associates ~
172 3 H amilton Avenue Suite K San Jose, California 95125 (408) 266-2716 E XP E RI EN CE :
1976 - PRESENT V i c e -P r e s id e n t - MHB Technical Associates, San Jose, California.
Engineering and energy consultant to state, federal, and private organizations and individusals. Major activities include studies
- of safety and risk involved in energy generation, providing tech-n i c a l c or. . u ?. t in g to legislative, regulatory,-public and private groups and expert witness in be'ialf of state organizations and citizens' t,roups. Was co-editor of a critique of the Reactor S af ety S tudy (W ASH - 14 0 0) f or the Union of Concerned S cien tis ts and co-author of a risk analysis of Swedish reactors for the Swedish Ene rgy Commission. Served on the Peer Review Group of the N RC/TMI Special Inquiry Group (Rogovin Committee). Actively involved in the h1 clear P ower Plan t standards Committee work for the Ins t rumen t Society of America (ISA).
1972 - 1976 -
Manager, Advanced Control and Ins trumen tation Engineering, General Electric Company, Nuclear Energy Division, San Jose, California.
Managed a design and development group of thirty-four engineers and support personnel designing systems for use in the measurement, control and operation of nuc1 car reactors. Involved coordination with other reactor design organizations, the Nuclear Regulatory
~~Camnisrton, and c u s t o m e r s ', both overseas and domestic. Responsi-bilities included coordinating'and managing the design and development of con trol systems, safety systems, an d new control concepts for use on the next generation of reactors. The position included responsibility for standards applicable to control and instrumentation, as well as the design of short-term solutions to field problems. The disciplines involved included electrical and mechanical engineering, seismic design and process computer control /
programming.
9
- . . - - , . - , , . . - . - - - .. - -. . . . - .,.,- . ,, . - , . . , - , _m,__.-.,,m..v-m--,---,,y4- ~ - . . ., - . , - --
l l
I l
1970 - 197' Manager, reactor Control Systems Design, General Electric Company, Nuclear Energy Di v i s ion _,_ ji a_n _ J o_sh California.
Managed a group of seven engineers and two support personnel in the design and preparation of the detailed system drawings and control documents re la t in g to safety and emergency syste=s for nuclear reactors. Responsibility required coordination with other design organizations and interaction with the customer's engineering personnel, as well' as regulatory personnel.
1963 - 1970 Design Engineer, General Electric Company, Nuclear Energy Division, San Jose, California.
Responsible for the design of specific control and in s t rum en t a't ion systems for nuclear reactors. Lead design responsibility for various subsystems of instrumentation used to measure neutron fluxPerformed in the reactor during startup and in t ermedia~t e powe r operation.
lead system design function in the design of a major system for measuring the power generated in nuclear reactors. Other responsi-bilities included on-site checkout and testing of a complete reactor control system at an experimental reactor in the Southwest. Received patent for Nuclear Power Monitoring System.
1960 - 1963 Advanced Engineering Program, General Electric Company; Assignments in W ashin g ton , California, and Arizona.
Rotating assignments in a variety of disciplines:
- En g in e e r , reacto r main t en an ce and instrument design, KE and D, reactors, H an f ord , Washington , circuit design and equipment ma in t en an c e coordination.
- Design engineer, Microwave Department, Palo Alto, Cali-f orn ia. Worked on design of cavity couplers fo r TWT's.
- Design engineer, Computer Department, Phoenix, Arizona.
-- - De s Lgn of core d rivin g. c ircuit ry .
- Design engineer, Atomic Power Equipment Department, San Jose, California. Circuit design and analysis.
- Design engineer, Space Systems Department, Santa B a rb a ra, Calif o rnia. Prepared control po r tion o f satellite proposal.
9
- - - - -,- , . - - - - - , - - , -, .,, , - - , , , , , ,-,--,-,-n-
)
1
, .= .
- Technical Staff - Technical Military Planning Operation.
(TEMPO), Santa Barbara, California. Prepare analysis of missile exchanges.
During this period, completed three-year General Electric program of extensive educa tion in advanced engineering p rinc ip l es of h i gh -
er mathematics, probability and analysis. Also completed courses in Kepner-Tregoe, E f f ective Pres enta tion, Management Training Pro- -
gram, and various technical seminars.
EDUCATION University of California at B e rk eley , BSEE, 1960.
Advanced Course in Engineering - three-year curriculum, General Electric Company, 1963.
Stanford University, MSEE, 1966.
HONORS AND AS S O CI ATIONS
- Tau Beta Pi Engineering H ono rary S ocie ty .
- Co-holder of U.S. Patent No. 3,565-760, " Nuclear Reactor Power Monitoring System," February, 1971.
- Member: American Association (or Advance of Science.
- Member: Nuclear Power Plant Standards Committee, Instru-ment Society of America.
P E RS ON AL DATA Born: June 7, 1937 Married, three children Res id en ce : San Jose, California
. -. ..~. --
0 9
- - . - , , , - , , , - , - , - - , . - - . , ----,--w. , - , - -,v---..,-,,----,,---,-,,m---, -,,.m-- - , , , , , . . - , , , - + . . . - - , , , , - , , ~ ~ - , .--
e 4 e P U3 LI C ATIO N S AND TESTIMONY Mdore,
" Control Rod Signal Multiplexing,"
- 1. G.C. M in o r , S.E. Vol. NS-19, February, IEEE Transactions on Nuclear Science, 1972.
Milam, "An Integrated Con trol Room S ys tem
- 2. G.C. Minor, W.G.
Nuclear Power Plant," N E DO -lC presented at In-for a Fair6 and 5 8, Technical Meetings, ternational Nuclear Industries
- O c t ob e r , 1972, Basle, Sdit:erland.
3.
The above article was also published in th e German Technical Magazine, NT, March, 1973.
H ubba r d Mino r , D.G. B r id e nb a u gh , and R.3.
- 4. T es timony o f G .C. Atomic Energy, Eearings held before February the18, Joint Committee 1976, on and published by the Union of Concerned S cien tis ts , Camb r id ge , Massachusetts.
D.C. B ridenbaugh, and R.3. Hubbard
- 5. Testimony of G.C. Minor, before the California State Assembly Committee on D.e s o u r c e s ,
Land Use, and Energy, March 8, 1976.
the Cali-
- 6. Tes timony of G.C. Minor and R.Bon. Hubbard bef ore Public Utilities, Transit, f ornia S tate Senate Co mmit t e e and Energy, March 23, 1976.
7.
Tes timo ny o f G .C . Minor regarding the GrafGermany enrheinf eld Nu-clear Plant, March 16-17, 19 7 7, W urzburg ,
- 8. Testimony of G.C. Mino r bef ore the Cluf f Lake Board of In-quiry, Regina, Saskatchewan, Canada, S ep t emb er 21, 19 A Review of theKendall, NRC 9 '. The Risks of Nuclear Safety Power Reactors:
Study WASH-1400 (NUREq-75/0140), _
H.
et al, _ edited by G.C. Minor and R .B . Hubbard for the Un io n Reactor of Conc ern ed S cientis ts , August, 1977.
3arseb'dck Risk Assessment, 10.
Swedish Reac tor S af ety Study: (Publish ed by MH3 Technical Associates, January, 1978.
S w ed is h Department of Industry as Document Sd1 1978:1)
Testimony by G.C. Minor bef ore the Wisconsin Public Service 11.
Co mmi s s ion , February 13, 1978, Loss of Coolant Accidents:
Their P rob ab ility and Consequence _.
Testimony by G.C.
Minor bef ore the Cali ~ornia Legislature 12.
As s e mb ly Committee on Resources, Land Use, and Energy, California.
3108, April 26, 1978, Sacr.nento,
_4
i I
P UB LI C ATION S AND TESTIMONY
- 13. Presentation by G.C. Minor before the Fede ral Minis try for Research and Technology ( B MFT) , Mee tin g on Reactor Safety Research, Man /Mr. chin e In terf a c e in Nuclear Reactors, August 21, and September 1, 1978, Bonn, Germany.
- 14. Testimony by G.C. Minor, D.G. Bridenbaugh, and R.B. Hubbard, before the Atomic Safety and Licensing Board, September 25, 1978, in the matter of the Black Fox Nuclear Power S tation Construction Permit Hearings, Tulsa, Oklahoma. -
- 15. Testimony of G.C. Minor, ASLB Hearings Related to TMI-2 Accident, Ra
- cho S e co P ower P lan t ,- on behalf of Friends of the Earth, September 13, 1979.
- 16. Testimony of G.C. Minor before the Michigan State Legisla-ture, Special Joint Conmittae on Nuclear Energy, Impliedtions of Three M i l e_,1,s,l a n d Accident for Nuclear Power Plants in Michigan, 10/15/79
- 17. A Critical View of Reactor Safety, by G.C. Minor, paper presented to the American As s o c i a t ion for the Advancement 7, of Science, Symposium on Nuclear Reactor S a f e ty , January 1980, San Francisc.o, California.
- 18. Th2 Effects of Aging on S af ety of Nuclear Power Plan ts ,
paper presented at Forum on Swedish Nuclear Referendum, S t oc kh olm, Sweden, March 1, 1980,
- 19. Minnesota Nuclear _?lants Gaseous Emissions S t u d y_ , MHB Te chnical As so cia tes , S ep temb e r, 1980, prepared for the Minnesota Pollutioa Control Agency, Roseville, MN.
- 20. Testimony of G.C. Minor and D.G. Bridenbaugh before the New York State Public ,S erv ic e Commission, Shoreham Nuclear Plant Construction Schedule, in the matter of Long Island Lighting Company Temporary Rate Case, September 22, 1980.
Bridenbaugh before the
- 21. Testimony of G.C. Minor and D.G.
New Jersey Board of Public Utilities, Oyster Creek 1980 Refueling Outage Investigation, in the matter of Jersey
~ - ~ - Tentral Power and Light Rate Case, February 19, 1981.
O
,-s--c , , , , , ----,..---...,-,c --.
UNITED STATES OF AMERICA i NUCLEAR REGULATORY COMMISSION l BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of '
)
)
LONG ISLAND LIGHTING COMPANY ) Docke t No. 50-322
)
(Shoreham Nuclear Power )
Station, Unit 1) )
g FIDAVIT RE RESPONSE TO APPLICANT'S INTERROGATORIES DATED MAY 21, 1981 I am duly authorized and did assist in the preparation of Shoreham Opponent's Coalition (SOC's) response to Applicant's Interrogatories A through M dated May 21, 1981. The answers are true and correct to the best of my knowledge and belief.
' 2.D h J +/7' Stephe'n B. Le thani t-Sworn to before me this s
9th day of June, 1981.
, 1 j . j ? ,/
.- m % -$-) ' re f , . . /,1) =--*
JANICE M. OLSEN NOTARY PUBLtC. State of New Ycrk No. 524527177, Suffolk Ceanty Term Expires March 30,19 l' /
I e *#
. 9
- - - - - _ _ - --...,v -, --- ,.__- , _ . _ _ _ _ _ . - .
. ,-. , _ , _ . _. , , . . _ , _ ,_,-,,._y., . _ . . _, -- ,,--
I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
l In The Matter Of -
)
)
LONG ISLAND LIGHTING COMPANY )
) Docke t No. 50-322 (Shoreham Nuclear Power )
Station, Unit 1) )
AFFIDAVIT RE RESPONSE TO APPLICANT'S INTERROGATORIES DATED MAY 21, 1981 We are duly authorized and did jointly assist in the prepa-ration of Shoreham Opponent's Coalition (SOC's) response to Applicant's Interrogatories A through M dated May 21, 1981. The answers are true and correct to the best of our knowledge and belief. ..
t '--,
DALE G. BRIDEI[BAUGH /
.0x4J A 4(('%f RICHARD B. HUBBARD
- dN(
G'/EGOIE C. MINOR Subscribed and sworn to before me thi,s 5 th day o f June , 1981 ,_-- -- - - -
' d OFFICIAL SEAL '
/ ./ # / '
lINDA L ROBUtSCN NOTMY PV9 TIC
- CAlff 0RMW
' O SANTA CUuta courm Notary Public 4
pr e .em aus a. ns2 [
j/ <
My Commission ~ expires 8' d f- J_3 0
. - , _ -_-