ML20069L832
| ML20069L832 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/15/1982 |
| From: | Hiatt S OHIO CITIZENS FOR RESPONSIBLE ENERGY |
| To: | CLEVELAND ELECTRIC ILLUMINATING CO. |
| References | |
| ISSUANCES-OL, NUDOCS 8211180270 | |
| Download: ML20069L832 (23) | |
Text
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'82 DV 17 Pi:31 UNITED STATES OF AMERICA .
NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of . )
)
CLEVELAND ELECTRIC ILLUMINATING ) Docket Nos. 50-440
) 50-441 C,0MPANY, et al. ) (0L)
(Pehry Nuclear Power Plant, )
Units 1 and 2) )
,- OCHE RESPONSE TO " APPLICANTS' INTERROGATORIES AND RIQUEST FOR PRODUCTION OF DOCUMENTS TO INTERVENOR OEIO CITIZENS FOR RESPONSIBLE ENERGY (SECOND SET)"
Intervenor Ohio Citizens for Responsible Energy ("0CRE")
hereby files its response to Applicants' Second Set of Inter-rogatories, dated September. 22, 1982. In o'rder to conserve its scarce resources, OCRE will not reproduce herein the interrogatories propounded it; the interrogatories are answered in the same sequence and numeration encountered.
OCRE will not produce herewith the documents identified in these responses, as requested by Applicants, since most of these documents are publicly available and the production of same by OCRE would be too great a burden on its limited resour'ces. If Applicants are unable to obtain any document otherwise, OCRE will provide a copy at a cost of $0.10 per paEe plus postage.
ISSUE #8
- 1. No such persons have yet been identified by OCRE.
, _ , . . . - - - wy - - -
';8211180270 821115 -
PDR ADDCK 05000440 3SDS
. . t
- 2. No witnesses have been identified as yet.
- 3. Final Safety Analysis Report for the Perry Nuclear Power Plant, Sections 6.2.5 and 7.3.1.
Letter, dated September 16, 1982, from A. Schwencer, NRC, to D. Davidson, CEI, re Request for Additional Infor-mation Regarding Degraded Core Hydrogen Control for PNPP.
Proposed Rule to 10 CFR Part 50, " Interim Requirements Related to Hydrogen Control" 46 Fed Reg 62281, December
.23, 1981.
\
Final hule to 10 CFR Part 50, " Interim Requirements Related to Hydrogen Control" 46 Fed Reg 58484, December 2, 1981.
SECY-80-107, February 22, 1980, " Proposed Interim Hydrogen Control Requirements for Small Containments" b
SECY-80-107A , April 22, 1980, "Aduitional Information Re Proposed Interim Hydrogen Control Requirements" SECY-80-107B, June 20, 1980, " Additional Information Re Preposed Interim Hydrogen Control Requirements" hegulatory Guide 1.7, Revision 2 (November 1978), " Control of Combustible Gas Concentrations in Containment Fol-lowing a Loss-of-Coolant Accident" Branch Technical Position CSB 6-2, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident" SECY-81-245, April 17, 1981, " Interim amedments to 10 CFR Part 50 Related to HydroEen Control and Certain Degraded Core Considerations" t
o -
" Pressure and Temperature Transients Resulting from Pos-tulated Hydrogen Fires in Mark III Containments" by Mark P. Paulsen and John O. Bradfute, Energy Incorporated, EI 75-4, February 1975.
NUREG/CR-1659 Vol. 4, " Reactor Safety Study Methodology Applications Program: Grand Gulf #1 BWH Power Plant" f
'T October 1981.
NUREG-0626, " Generic Evaluation of Feedwater Transients
,- and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Hear-Term Operating Licensing Applications" February 1980.
NUREG/CR-2540, "A Method for the Analysis of Hydrogen and Steam heleases to Containment During Degraded Core Cooling Accidents" February 1982. .
NED0-10812, " Hydrogen Flammability and Burnirig Characteristics in BWR Containments" General Electric, April 1973.
NUREG/CR-0913, " Generation of rlydrogen During the First Three Hours of the Three Mile Island Accident" July 1979.
NUP1G/CR-1575, " Hydrogen Mixing in a Closed Containment Compartment Based on a One-Dimensional Model with Convective Effects" September 1980.
HUREG/CR-1561, "The Behavior of Hydrogen 'During Accidents in Light Water Reactors" August 1980.
NUREG/CR.-2017, " Proceedings of the Workshop on the Impact of Hydrogen on Water Reactor Safety" Volumes 1-4, September 1981.
NUhEG/Ch-1250, "Tnree Mile Island: A Report to the Com-
_4_
. \ .
missioners and to the fuelic" Volume II, .by the NRC u
Special Inq'iry Group, M. hogovin, Director.
" Containment of a Reactor Meltdown" by Jan Beyea and Frank Von Hippel, The Bulletin of the Atomic Scientists, Vol. 38, No. 7, August / September 198.2, pp. 52-59.
Transcript of May 27, 1982 meeting to discuss concerns of
' John Humphrey re Mark III containment, pp. 174-177, t
- e 181-185, 202-205, 257-260.
IE Bulletin 79-08, April 14, 1979, " Events Relevant to Boiling Water Powe'r Reactors Iden41fied During Three Mile Island Accident"
' IE Bulletin 79-05, April 2,1979, and 79-05A, April 5, l'979,
" Nuclear Incident at Three Mila Island"
- 4. Potentially any or all of the documents identified above o
may be offered as exhi-bits or used during cross-examination in support of Is' sue #8.
- 5. To the extent that this. interrogatory requires OCRE to define the accident scenario which will Eovern the liti-gation of Issue #8, OCRE objects to this interrogatory, as this is not OCRE's responsibility. The Appeal Board stated that "(i)t is the Licensing Board's function to determine what a TMI-2 type accident is, insofar as the Perry facility is concerned." ALAB-675, slip op. at 19, footnote 13.
To the extent that this interrogatory requires OCRE to perfonn a detailed time-domain analysis of a TMI-2 type accident scenario specific to Ferry,'OCRE objects to this
' -5 interrogatory, as OCRI does not have-the resources for running the computer simulation models necessary to deter-mine the rate and quantity of hydrogen production during l
the accident.
Furthermore, it should*oe noted that the TMI-2 accident does not " represent a unique scenario by which lar6e amounts
/
of hydrogen with eventual core cooling could be acheived.
It is not possible to define a unique scenario since there .
are nnmerous ways in which the same end results could be obtained." NUREG/CR-2540 at 3. An examination of Table 5-4 of NUREG/CR-1659, Vol. 4 reveals that for the BWR/6-Mark III, containment failure from hydrogen burn is a likely result of 18 of the 36 accident sequences analyzed, which included large oreaks, small breaks, an'd transients. ,
The IMI-2 accident involved a feedwater transient,. stuck-open PORV, and operator error, with the consequences being E damaged core, an estimated 35-50% metal-water reaction, and the detonation of some of the hydr 6 gen thus pr.oduced in the containment. An equivalent scenario for the.BWH/6 would involve generally similar events. For the purpose of responding to the subsequent interrogatories, OCRE will
~~
ass'ume that the T23PQE-Y sequence analyzed at pp. 6-12 to 6-14 and C-6 to C-10 of HURED/CR-1659 Vol. 4 is an equiv-alent to the TMI-2 type accident. This accident involves:
Tc3 - loss of feedwater transient P - failure of a safety / relief valve to resent
= _ _ _
~
\
Q - failure of tne pow'er conversion system to to provide makeup water
- E - failure of the ECCS
- k
.}/-containmentfailureduetohydrogenburn According to the NUREG/CR-1659 analysis (specific to Gran Gulf), core melting begins at 71 minutes. At pp. C-21 to .
- C-26 of NUREG/CR-1659 some of the parameters calculated with the MARCH code for the T 23PQE sequence are presented graphically as a function of time. These graphs do not include any information on the nate and quantity of hydro -
gen production.
The. dominant contributor sequences to T23PQE are listed on'p.'D-4 of NUREG/CR-1659.
The only operator error assumed for some of the sequences is the failure of the operator 'to manually initiate the ADS.
b 6.
i
' OCRE believes that.any accident which is physically possible is credible. -
Credibility does not depend on predictability or probability. The history of the nuclear industry has shown that accidents can begin and proceed in the most unlikely and unpredictable ways, e.g., Browns Ferry fire, TMI-2, Browns Ferry 3 partial scram failure, Fermi breeder accident, etc.
In addition, Appendix B of NUREG-0626 indicates that feed-water transients and safety / relief valve failures occur with moderate frequency in BWRs.
, 7.
OCRE does not have the resources for determining, through sophisticated computer models, the off-site radiation doses associated with the T 23PQE S[ sequence.
Even if OCRE were able to run such computer models, it:should be noted that y-
such analyses are. subject to large uncertainties _and are dependent upon various assumptions,.such as meteorological conditions and the effectiveness of protective actions.
NU.tEG/CR-1659Vol.4considerstheT23QE-Tsequenceto-P fall 1nto release category BWR-3 and to have a probability of 2.7 x 10-7 per reactor-year. Using Fi Eure 5.4 of the
, / Perry FES, NURIG-0884, and assuming that the T23 PQE"(
I sequence does occur at PNPP, it can be expected'that at least"100 persons will receive whole body radiation doses
' in excess of 200 rems; that at least 20,000 persons will receive thyroid doses in excess of 300 rems; and that at least 100,000 persons will receive whole body doses in excess of 25 rems.
- 8. 10 CFR 100.11(a)(2) states that a person located at the low-population-zone boundary should not receive a dose in excess of 25 rems whole body or 300 rems to the thyroid. PNPP's LPZ has a radius of 2.5 miles. About 4225 residents live within the LPZ, which also has a peak transient population of 1575. Perry SER, NUREG.-OsS7, Section 2.1.3. Thus the worst-case total LPZ population would be 5800 persons.
The estimates given in the response to Interrogatory 7, above, indicate that the number of persons receiving at least the 10 CFR 100 doses vastly exceeds the LPZ population.
Thus, it is reasonable to assume that a person located at the LPZ boundary would receive radiation doses greater than the 10 CFR 100 values.
- 9. (a) All documents relied upon were identified in the responses.
- ~
1
~6-
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(b) No such persons have yet been identified.
- 10. (a) OCRE has no documents specifically concerning igniters; t
nowever, some of the documents identified in the response to Interrogatory 3, above, dontain some applicable info'rmation.
. (b) No persons have provided OCHE with information, t
- expert advice or knowledge.
(c) OCRE believes that igniters will not safely control i
hydrogen produced by the T23PQE scenario. Figure s
q C-ll, p. C-23 of NUREG/CR-1659 Vol.- 4 indicates that approximately 55% of the fuel rod cladding has reacted to liberate hydrogen by 105 minutes into the accident. Assuming that all of this hydro-b gen is released into the wetwell (and ignoring any other sources of hydrogen, e.g., radiolysis of water),
Figure 2 of NUREG/CR-1561 indicates that in a Mark III, a 55% metal-water reaction will result in a hydrogen concentration in,the containment _of about 22 vol-%.
Since the Mark III containment is not inerted, l oxygen ls present to support combustion. Igniters l
Ere designed to induce hydrogen combustion at con-l
~~ ~
centrations of 4-8 vol-% (Applicants' Answer to OCRE Interrogatory 5-14). These igniters, as with i
l all parts of the PNPP hydrogen control system, are initiated manually. There may be a delay of up to e
(
60 minutes before the hydrogen analyzers, the first I
component to be activated, are in use.
PNPP FSAR Section 6.2.5. Of course, there is also the possibilit]
l
-g.
that, as with any manual action, the hydrogen control system will not be actuated at all, especial-ly during the busy and stressful time of an accident.
Figure C-ll of NUREG/CR-1659 indicates that the
~
metal-water reaction begins at about 68 minutes into the accident; the reaction rate gradually ir?.reases
/
} until 105 minutes, at which time there is a rapid increase in reaction rate. Therefore, even if one
- a' assumes that.the igniters can control hydrogen at the lower generation rates, at 105 minutes the rate increases such that there is a dangerous rise in hydrogen concentration. Specifically
- prior to t = 105 minutes, about 18% of the fuel rod Adding has reacted. The resultant hydrogen is assumed to be removed by the hydrogen control system (recombiners-igniters). After t = 105 minutes, an additional 37%
of the claddfng reacts rapidly; according to.the NUREG/CR-1561 data, 57% metal-water reaction corres-ponds to a 15%'.concentr ation by volume of hydrogen in the containment. According to Figure 3 of NUREG/CR-1561, adiabatic combustion of hydrogen, with ainitial
~~
' conditions of atmospheric pressure, temperature of 2500, and air saturated with water vapor, will result in a pressure of 5.7 atmospheres (84 psia) and a temperature of 17000 K (26000 F), for an initial hy-droGen concentration of 15 vol-%. Compare these values with the Mark III design maxima of 15 psig
' I and 185 F. Containment failu,re is certainly probable.
The igniters would probably be the source of ignition.
Assuming the most limiting condition, i.e., that the hydroEen control system is not effective at all at the lower rates and concentrations, combustion of the full 22vol-% of hydrogen would yield a pressure of 7 atmospheres (103 psia) and a temperature of 2400 K (3860 F), again from FiEure 3 of NUREG/CR-1561.
Figure C-14 of NUREG/CR-1659 Vol. 4 indicates that a hydrogen burn is expected at tA 100 minutes, with Jt he pressure risinE to 175 psia.
It should be noted that for other accident sequences, e.g. , AE and SE (large and small pipe breaks with failure of the'ECCS), the rate of hydrogen generation may 'oe even greater. Figure C-25 of NUREG/CR-1659 Vol. 4 indicates that for these sequences, a 50%
metal-water reaction occurs within 50 minutes.
Tne safety of using igniters to control hydrogen has been questioned by'many; e.g.:
- General Electric: in view-graphs presented to the NRC
( and contained in SECY-80-107A, GE considered burning i
.. .to be " impractical for significant rates and all sources" and identified several other problems with l ignition, e.g., how to remove the heat of combustion and the difficulty of ensuring prompt ignition.
Interestingly, GE also considers other hydrogen l
l control methods (inerting, recombiners, venting) to be impractical as well.
- NRC: "the Commission believe,s that control methods that do not involve ourning provide protection for a wider spectrum of accidents than do those t, hat involve burning" 46 FR 62282, December 23, 1981.
dA
- Hermann L. Jahn,.Battelle, Frankfurt, Germany:
, " Enforced burning e.g. by spark plugs may be I
/ expected to be of questionable value and perhaps dangerous." NUREG/CR-2017, Vol. 3, p. 273.
/ L. Berlad, et al. , Brookhaven National Labora-
- A.
- tory: "In assessing the advantages and disadvantages of various hydrogen control approaches, we have strongly favored methods which eliminate the com .
bustion of accident-released hydrogen. " NUREG/CR-2017, Vol. 4, p. 168.
- 11. Pre-accident inerting of the containment will prevent hydrogen combustion. The effectiveness of post-accident inerting depend' s , on 'the rapidity with which it is initiated.
Containment inerting, however, will not prevent failure of-the containment by overpressurization by hydrogen and other noncondensible Eases, with no combustion.
NUREG/CR-1659 Vol. 4 at p. 6-14 indicates that there is an equal
__ chance of containment failure in the T23 PRE sequence by either hydrogen burn or overpressurization; the latter mode of containment failure, however, results in a less severe release category. General Electric, in SECY 107A, states that for a 100% metal-water reaction, the hydrogen pressure alone could exceed the containment design pressure, and thus inerting may not prevent con-
4
. \
tainment failu'e.
r In NUR3G/CR-201,7, -Vol. 4, p. 162, it is stated that "of all the hydrogen control approaches considered, a strategy of continuous inerting of the con-tainment building is the only one which clearly eliminates the combustion hazard, does not involve adverse environ-mental effects, and succeeds in a way that is independent of the accident scenario." Post-accident inerting (i.e.,
inerting during the accident period) has the disadvantage of increasing containment pressure through the addition of inert gas during a time when the c'ontainment atmosphere cannot be vented. NUREG/CR-2017, Vol. 4, p. 179.
ISSUE #9 '
- 12. No such persons-have yet been identified.
- 13. No witnesses have been ' identified as yet.
- 14. NUREG/CH-2156, " Radiation-Thermal Degradation of PE and PVC: Mechanism of synergism and Dose Rate Effects,"
Roger L. Clough and Kenneth T. Gillen, Sandia National Laboratories, June 1981.
NUREG/CR-2157, " Occurrence and Implications of Radiation Dose-Rate Effects for Material Aging Studies," Kenneth
_. T. Gillen and Roger L. Clough, Sandia National Labora-tories, June 1981.
Final Safety Analysis Heport.for PNPP, Section 3.11.
Propos'd e rule to 10 CFR 50, " Environmental Qualification of Electric Equipment for Nuclear Power Plants" 47 FR 2876, January 20, 1982.
Memorandum for Raymond F. Fraley, Execut'ive Director, ACES, from Robert B. Minogue, Director, NRC-RES, dated April
16, 1982, re final rule 10 CPR 50.49 and analysis of public connents on proposed rule.
Final Rule, 10 CPR 50, " Environmental Qualification of
, Electric Equipment" 47 FR 28363, June 30, 1982.
'DE Bulletin 79-01, " Environmental qualification of' Class lE Equipmen't" February 8, 1979.
/ IE Bulletin 79-OlA, June 6, 1979.
T IE Bulletin 79-OlB, Supplements 2 and'3, Sept. 30, 1980
/ and'Oct. 24, 1980.
,/
CLI-80-21, 11 NRC 707 (May 27, 1980) Memorandum and Order, in the matter of Petition for Emer'gency and Remedial Action.
IE Information Notice 81-20, " Test Failures of Electrical Penetration Assemblies" July 13,.1981.
IE Information Notice 81-29, "Eqnipment Tes ting Experience" September 24, 1981. -
IE Information Notice 82-03, " Environmental. Tests of Electrical Termi'nal Blocks" March 4, 1982.
Documents pertaining to specifications for electrical
, caoles used at PNPP, listed in the attached letter of November 2,1982 from Ronald Wiley, CEI to Susan Hiatt, OCRE.
- 15. Potentially any or all of the documents identified above
,, may be, offered as exhibits or used during cross-examination in support of Issue #9.
- 16. Based on research performed at Sandia National Labora-tories, documented in NUREG/CR-2156 and NUREG/CR-2157, OChE-believes that the following polymers degrade more rapidly' when exposed to lower levels of radiation for longer periods of time (i.e. , normal service conditions) than when exposed to high levels for shorter periods.
\
(i.e., integrated-dose accelerated aging qualification tests):
polyvinyl enloride (PVC) polyethyl'ene (PE) crosslinkedpolynkefin(CLPO) ethylene propylene rubber (EPR) chloroprene rubber (CP) chlorosulfonnted polyethylene (CSPE) -
- 17. OCRE believes that any components or equipment using these i
polymers in a radiation environment may suffer from degrada-tion. Applicants have stated that safety-related electrical cable and wire used in a radiation environment use cross-linked polyolefin, crosslinked polyethylene, and ethylene h propylene rubber as insulation. (Applicants' answer to OCRE Interrogatory 3-4) Therefore, OCRE believes that the caole types identified below, when used in the locations identified in the answers to Interrogatories 18, 20, and 21, below, may be subject to dangerous degradation.
Particular circuits will be identified later.
Cable type Use Polymer EKB- (xx) control CLPO, hypalon jacket EKC-71 to instrumentation CLP0 EKC-107 EKF- (xx) thermocouple CLPE, hypalon jacket EXC-1 to instrumentation CLPE, CSPE jacket EKC-50 EKA-61 to power CLPO, hypalon jacket EKA-110; EKA-140 to to EKA-206
Cable type Use Po1ymer EKA-2 to power EPR, CSPE jacket EKA-58; EKA-lll to EKA-138 ;
EKA-151 to EKA-159 (This information was obtained from documents provided by Applicants. for, inspection at FNPP; copies of same
,/ p'rovided to OCRE are listed in the November 2, 1982 T
letter from R. Wiley, CEI, to S. Hiatt, OCRE. 0CRE is unsure whether hypalon is the same as CSPE.)
- 18. This interrogatory is somewhat un'elear as'to whether the locations to be identified are those made dangerous.by a possible accident caused or aggravated by the polymer degradation or those in which the radiation levels or ot'her environmental conditions are such that this degrada-tion is likely. Assuming the latter, these locations are generally identified in the answers to Interrogatories.
20.and 21, below.,
- Of particular concern to OCRE are those locations where cables serving redundant and/or diverse safety systems or components may be subject to similar degradation effects i due to similt.r radiation environments. E.g., FSAR Figures
- 8.3-14 to 8.3-17 indicate that some principal cable routes for both Division 1 and Division 2 Class 12 circuits are located within the same radiation zones.
NUREG/CR-2156 indicates that such " dangerous locations" miE ht be very localized and specific, more specific than the zones identified below. A cable (the discovery of which prompted the Sandia studies) which had been in service 1
l
t .
in a nuclear fa'cility and which. exhibited severe embrittle- -
- s ment in certain locations exhibited no deterioration in adjacent areas along the cable. This suggests that detailed dosimetry mappinE would oe necessary in order to pinpoint J
the exact locations of concern. OCRE suspects that such I dosimetry mapping would require sophisticated computer analyses and possible could not be performed except by
~
experiment. Since such techniques are beyond OCRE's resources, this Intervenor cannot identify the " dangerous
.: locations" with.any greater specifibity than the . radiation jf zones described below.
4 19. It was estimated in NUREG/CR-2156 (p. 8) that the maximum dose rate experienced by the severely ' degraded portions of i
g the electrical cable (the discovery of which prompted-the
- ' Sandia research) wasa' bout 25 rad /hr. This cable was in an unsafe condition, since the insulation fell off the-wire when bent. . Based on thisLinformation, OCRE would' therefore conservatively estimate that an average dose rate greater than or equal to 10 rad /hr might cause dangerous deterioration. (Should future research implicate even lower levels, OCRE will revise these responses accordingly.)
-- 20. The-radiation zones identified herein are the same as the
- environmental zones described in FSAR Table 3.11-1. The maximum normal Eamma radiation doses given for these zones are'inteErated over 40 years. Since 40 years is equivalent to 3.5 x 10 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, the average dose rate was obtained 5
by dividing the FSAR values by 3.5 x 10 . Usin6 data from FSAR Tables 3.11-2 through 3.11-8, the following zones 4
_ _ , . _ = ., _ , . ._ __e _
have dose rates during normal operation of at least 10 rad /hr:
Zone Dose Rate, rad /hr g//-l
- Df 80 Dh-2 . 129 -
4 DW-3 5.7 (This zone was
! included because a large neutron
/ !. fluence, 2.9 x 108 Ntn/cm2/hr, is also present.)
f (7 DW-4 80 DW-5 80 CT-5 514 AB-8 22 FB-6 27
- 21. The radiation zones identified herein are the same as the environmental zones identified in FSAR Table 3.11-1.
The gamma radiation doses given for these zones for accident conditions are inteErated over 6' months. Since 6 months is equivalent to 4380 hours0.0507 days <br />1.217 hours <br />0.00724 weeks <br />0.00167 months <br />, the average gamma dose rate was obtained by ' dividing the FSAR values by 4380. In addition, the beta dose rates are calculated to be 67 krad/hr for all drywell zones and 25 krad/hr for all containment zones (see Notes 2 and 3 to T' ables 3.11-2 to 3.11-8). The total averaSe dose rate is thus the sum of the summa and beta values. Accident conditions are those identified in the FSAR tables. Using data from FSAR Tables 1
3.11-2 through 3.11-8, the following zones have dose rates during accident conditions of at least 10 rad /hr:
. t ,
Zone . Dose Rate, rad /g D#-l 128 x 103 DW-2 128 x 103 DW-3 128 x 103 DW-4 . 128 x 10 3 DW-5 128 x 103 CT-1 35 x 10 3 CT-2 29.x 103 CT-3 29 x 103 CT-4 \29 x 103 CT-5 29 x 103
]
CT-6 -
29 x 103 CT-7 29 x 103 29 x 10 3 g CT-8 -
AB-2
~
6.8 x 103 AB-3 4.3 x 100 3
AB-4 ,
9.4 x 10 AB-5 2.5 x 10 3 i AB-7 3.9 x 10 3 i
AB-8 2.2 x 10 3 FB-6 1.2 x 105
- 22. (a) All documents were identified in the response.
The calculktions made in unswering Interrogatories
- 20 and 21 used data in Amendment 9 of tne FSAR.
Should a later amendment provide di.fferent data,
the responses will be amended' accordingly.
(b) No such persons have been identified.
23-26. The information requested by these Interroghtories is provided by the attached affidavit. OCRE objects to the portion of Interrogator; 26 dealing with the location f where the document search was conducted as this in-T formation is not relevant 'to either Issues 8 or 9 and
~~'" '
- infringes too closely upon the work-product doctrine.
F.R.C.P. 26 (b) (3).
/
, 27. The following statements pertain to Issue #8:
(a) Separate Views of Commissioners Gilinsky and Bradford and Separate opinion of 'Commissione~r.'BrkAford, to Duke Power Company (Wm. B. McGuire Nuclear Station, Units 1 and 2) CLI-81-15, 14 NRC l (1981); these statements concern the litigation of hydrogen control under 10 CFR'Part 100.
(b) Separate Views of Commissioner'Gilinsky to Final Rule to 10 CFR Parts 2 and'50: " Licensing.Requirecents for Pending Construction Permit and Manufacturing License Applications" 47 Fed Reg 2286 at 2300 I
(January 15, 1982), in which the Mark III is c'harac-(
terized as a weak containment which should be required to be stronger.
Respectfully submitted, e
i Susan L. Hiatt OCHE Representative 8275 Munson Rd.
Mentor, OH 44060 -
.A' -
i T H E C L E VP ERRY, Ei.v.W D EL E CTkO C~ 1e L T EL EPHON E (216) 256-37 37 L U !.O W AE
' ADDRESS-10 CENTER RO AD OHIO 44081 s
' ' P.O. BOX 97 s Senk T$ lit Lpcfug in the Nation N'e'vember2,1982lh,.h,...3fte J 74;. y Ij ERl$cy VICE k
I
- i
' Ohio Citizens for Responsible Energy l c/o Ms. Susan Hiatt
- 8275 Munson Road i Mentor, Ohio 44060 l
Dear Ms. Hiatt:
i On behalf of Ohio Citizens f or Responsible Energy, you requested copies of documents relative to issue #9 which you inspected at PNPP on October 24, 1982.
4 Copies of_ each of the following documents are enclosed:
l From the file entitled, " Cable - Misc. Saf ety Related Instrumentation l Specification SP-793-01":
Bill of Material Sheets: EKB 51, 61, 71 & 81 EKC-71 through EKC-107 j h RSS-6-104, RSS-6-105, RSS-6 110, Rockbestos Co. Spec Sheets:
RSS-6-ll2, RSS-6-116, RSS-6 200, RSS-6-207 From the file entitled, " Cable Thermocouple Extension - Class lE Specification SP-567',' Conf ormed Spec. pp.2 7-11 Bill of Material Sheets: EKF-1 through EKF-35 (except those which are deleted)
From the file entitled, " Cable Instrumentation Class lE Specifica-tion SP-561 Auditable Material" Conformed Spec. pp. 7-11 Bill of Material Sheets: ERC-1 through EKC-50 From the file entitled, "Cabic. Class lE Small Power 6 Control Cable SP-560 Auditable Material'," Conf ormed Spec. pp. 7-11 ERA-61 through EKA-206 Bill of Material Shee ts: EKB-l' through EKB-100 From the file entitled, " Cable Class lE Medium Voltage Power Cable Specification SP-559 Auditable Material',' Conf ormed Spec. pp. 7-13; 16-18 EKA-2 through EKA 159 Bill of Material Sheets:
7"'
r' '
Ohio Citizens f or Responsible Energy November 2, 1982 From the binder entitled, " Project Design Criteria":
pp. 2-48, 2-48a and op. 2 68 through 2-79 From the file entitled, " Cable Class lE Medium Voltage Power Cable -
AEIC-Standards"
- p. 8 from document entitled: ."Specifications for Polyethylene
/ and Cross-linked Pclyethylene" T p. 9 from document entitled: " Specifications for Ethylene Propylene Rubber" Please sign below and return one copy of this letter to indicate that you have received these documents.
Applicants' cost of duplication for these documents 0 $0.10 per p' age is $30.30.
Please remit a check in the above amount payable to THE CLEVELAND ELECTRIC ILLUMINATING COMPANY, and address to:
Ronald G. Wiley c/o Perry Nuclear Power Plant P. O. Box 97 Perry, Oh'o 44081 i
' Sincerely,
. THE CLEVELAND ELECTRIC ILLUMINATING CO.
Ronald G. Wiley Requested documents received by:
$9. b. YrW. Al0J.k /952.
Date Susan L. Hiatt Ohio Citizens for Responsible Energy RGW: dip cc: Jay E. Silberg
, , , _ - , , ,7- . - - -
t DCLMETED AFFIDAVIT #
I, Susan L. Hiatt, being duly sworn depose $ d t t I am responsible for the answers given in OCRE's R g g e[g{(g'{
Applicants'InterrogatoriesandRequestforProductioINk'
- ' Documents to Intervenor Ohio Citizens for Responsible Energy (Second Set) and that these answers are true to the best of my knowledge and belief.
f Susan L. Hiatt
\
Sworn to and subscribed before me this / day of November, 1982.
AA WA Notary Pub ftAND1 B.JENKINS, Attorney.at Law Notary PuMc, Sta'.e of Ohio
- My Commission Hes No Expiration Date Section 147.03 R. C.
I
+
CEhTIFICATE OF SERVICE 00H ETED
'JMc This is to certify that copies of the foregoing OCRE RESPONSE TO " APPLICANTS' INTERROGATORIES AND REQlWBTN P3Q 31 DUCTION OF DOCUMENTS TO INTERVEN0d OHIO CITIZENS FOR ES ONSIBLv ENERGY (SECOND S.rT)" were served by deposit in the,U.S. Mail, first class,postageprepaidthis15thdayofNgiemberfl9p2Ato
- ""y those on the service list below. pCf j VICE 6lu VW Susan L. Hiatt T SERVICE LIST Peter B. Bloch, Chairman Daniel D. Villt, Esq.
Atomic Safety & Licensing Board U.S. Nuclear Regulatory Comm'n P.O. Box 08159 Cleveland, OH 44108
' Washington, D.C. 20555 Dr. Jerry R. Kline Atomic Safety & Licensing Board U.S. Nuclear Regulatory Comm'n Washington, D.C. 20555 Frederick J. Shen Atomic Safety & Licensing Board U.S. Nuclear Regulatory Comm'n '
Washington, D.C. 20555 -
Decketing & Service Section Office of the-Secretary U.S. Nuclear Regulatory Comm'n Washington, D.C. 20555 James M. Cutchin, t:IV, Esq.
Office.of the Executive Legal Director U.S. Nuclear Regulatory Comm'n Jay Silberg, Esq.
1800 M Street, N.W. '
Washington, D.C.
20036 Atomic Safety. & Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D. C. 20555 e
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