ML20080S452
| ML20080S452 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/14/1983 |
| From: | Cooper W, Krechting J LOWENSTEIN, NEWMAN, REIS, AXELRAD & TOLL |
| To: | |
| Shared Package | |
| ML20080S365 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8310180375 | |
| Download: ML20080S452 (71) | |
Text
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8 UNITED STATES OF AMERICA I l
4 NUCLEAR REGULATORY COMMISSION l 5 !
6 BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD 7
8, '
d ) -
9 In the Matter of: '
10 PACIFIC GAS AND ELECTRIC Docket Nos. 50-275 0.L.
j ) ;
i COMPANY 50-323 0.L.
11 ' . j (Diablo Canyon Nuclear l 12 Power Plant, Units 1 and 2) ;
i 13 l
1 14 i f
15 TESTIMONY ON BEHALF 0F THE INDEPENDENT DESIGN VERIFICATION PROGRAM 16 0F 17 i Mr. John E. Krechting 18 Dr. William E. Cooper i
19 REGARDING 20 CONTENTION 4.a.-l. and o.-u.
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1i! In the Matter of: )
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2 1! PACIFIC GAS AND ELECTRIC ) Docket Nos. 50-275 0.L.
h COMPANY 50-323 0.L.
(Diablo Canyon Nuclear )
4[ Power Plant, Units 1 and 2) )
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51' g ;; TESTIMONY REGARDING CONTENTIONS 4.a.-l. and 4.o.-u.
d 7hINTRODUCTORYTESTIMONY I!
8 Q.1: Please state your name, current position, business n
9 j address and qualifications.
10 l A.1: This information is contained in A.l. of the Testimony a
11 Regarding Contentions 1,2 and 5-8.
F 12 -Q.2: Please describe your participation in the Independent 13 : Design Verification Program (IDVP).
e .
14 A.2: This information is contained in A.2 of the Testimony 15 Regarding Contentions 1,2 and 5-8.
16 ij Q.3: :ibat is the purpose of your testimony?
F:
17 - A.3: Contention 4 alleges that the IDVP " accepted 18 deviations from thd licensing criteria without providing adequate 19 ) engineering justifications" in a number of specific respects.
20 This testimony adcresses every subpart of Contention 4, except 21 Contentions 4.m. and 4.n., which are addressed in the Testimony 22 of the Panel addressing Contentions 3,4.m. and 4.n.
H 23 Q.4: Does every answer in this testimony constitute the 24 j testimony of both members of the panel?
25 A.4: Yes. Since Mr. Krechting had the responsibility for 26 the technical review by the IDVP of each of the subject areas 27 lj u
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Iffcoveredbythetestimony,heismorefamiliarwiththedetailsof .
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. 2 j such review. However. Dr. Cooper had overall responsibility for !
8 [ program management of the IDVP, reviewed and approved the 4 - disposition of the E01s which are referred to in the testimony, 5 and shares the judgments expressed in the testimony, 6i '
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CONTENTION 4.a. i 1[ " Contrary to the requirements of FSAR Section 17.1 regarding i
h compliance of the as-built installation with the design docu-2 i: ments, the IDVP review of the AFWS disclosed that the as-built !
J installation failed to meet the design drawings in that (i) a 8it steam trap on the turbine-driven AFW pump steam supply line is not provided and (ii) there are discrepancies in the arrangement ;
4 of the long-term cooling water supply line."
5 0.1: Does the IDVP know the origin of allegation (i) in i I
6 this contention? i 1 A.1: The IDVP believes that the origin of allegation (i) is 8 E01 8027.
9 Q.2: Please describe the issue there identified.
10 A.2: The IDVP's concern was that a steam trap shown on a 11 piping schematic drawing it was reviewing had not been installed.
I 12 Q.3: How was this concern resolved?
13 '
A.3. Af ter issuance of E01 8027, the IDVP determined that i
14 y although the design originally did not call for the steam trap, a H
15! l!! design change had been subsequently initiated adding the steam 16 y trap which appears in a piping schematic drawing reviewed by the 3 h 17 IDVP. However, the design change document was not signed by 18 h General Construction (G.C.) to authorize installation because it 0
19 r was subsequently determined by start-up testing that the trap was 20 i not required. Therefore, the design change adding the steam trap 21 was never officially approved. G.C. wrote a design change 1
22 ,' document superseding the original and revising the piping i
23
{ schematic.
24! Thus, the IDVP determined that the as-built condition (with-1 25 i out the steam trap) :orresponds to the approved design. The IDVP 26 review is reported in ITR-22.
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Q.4: Does the IDVP know the origin of allegation (ii)?
2 A.4: The IDVP believes that the origin of allegation (ii) ,
8 1 is E01 8048.
li 40n Q.5: Please describe the issue there identified, A.5: The IDVP's concern was that a check valve not shown on 5]
6l a piping schematic drawing it was reviewing had been installed in 7 ll the long-term cooling water supply line.
8d Q.6: How was this concern resolved?
E y A.6: Af ter E01 8048 was issued, the IDVP determined that 9 'l h
101 the long-term cooling water supply line had a check valve, as the i
11 l original design required. A design change to that supply line 12 l had been issued which did not require removal of the check valve.
13 0 However, a draf tsman misinterpreted a Xerox copy of the design 14 change document and incorrectly removed the check valve from the 15 piping schematic drawing. This error on the drawing was L
16 [ corrected.
17 h Thus, the IDVP verified that the as-built installation cor-b 18 ; ~; responds to the approved aesign. The IDVP review is reported in 19 , ITR-22.
20 i Q.7: Did the IDVP accept any deviations from licensing n
21;! criteria relating to the as-built installation of the AFWS?
22 + A.7: No. No deviation from the licensing criteria exists.
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p CONTENTION 4.b.
" Contrary to FSAR Section 8.3.3, the electrical design does 1h;not fully comply with the connitments regarding separation and 2 j color coding."
u 8j Q.1: Does the IDVP know the origin of this conter, tion?
4 A.1: The IDVP believes that the origins are E01s 8055 and 5j8059.
6[ Q.2: Please describe the issue identified in E01 8055.
- 6 A.2: In E01 3055, the IDVP's concern was that two pressure 7.[!
8 j indicators (P153A and P153B) on the main control board did not i
9 f' meet the separation criteria of FSAR, Section 8.3.3, in that the 10 indicators are less than five inches apart.
11 Q.3: How was this concern resolved?
12 A.3: Af ter issuance of E0I 8055, PGandE stated that the in-of the Section 8.3.3 separation criteria is to provide 13[ tent 14 adequate isolation and insulation between exposed current-15 l carrying partions of mutually redundant power control devices, e
16 e.g., transfer switches. However, these separation criteria were 17 g not intended to apply to low energy instrumentation signal n
18 devices such as the pressure indicators which were addressed in 19 ] E018055. The IDVP acc%d PGandE's interpretation of the FSAR, 20 ![ Section 8.3.3, as reasonable and consistent with the underlying 21 basis for this section, subject to PGandE's connitment to revise 22 ; Section 8.3.3 to clarify the requirements for separation of i
23 { mutually redundant indicating devices on the main control boaro.
9 24[TheIDVPreviewisreportedinITR-27.
25 L Q.4: Please describe the issue identified in E0I 8059.
h 26 ;ji A.4: The IDVP interpreted FSAR Section 8.3.3 to provide for 27 li the color-coding of safety-related cables only. Since the IDVP 28 i!
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ll 1[found that some non-safety-related cable was color-coded, it 2 d issued E01 8059 to assure that the matter would be clarified.
8ll Q.5: How was this concern resolved?
4 j A.5: After issuance of E01 8059, PGandE explained that the t
5: intent of FSAR Section 8.3.3 was to ensure that safety-related t-6 cable was identified but that it did not preclude color-coding of 7 ] some non-safety-related cable. The IDVP accepted PGandE's 1
8 jl interpretation of the FSAR, Section 8.3.3 as reasonable and con-
- l 9 sistent with the underlying basis for this section, subject to 10 0 PGandE's comitment to revise Section 8.3.3 to clarify the color-i!
11'{ coding requirements. The IDVP review is reported in ITRs-27 and 12 ll -28.
13 Q.6: Did the IDVP accept any deviations from licensing 14 L criteria relating to separation of electrical circuits and 15 [l devices or color-coding?
i 16 q - A.6: No. There were no deviations from licensing require-17 ments.
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- i CONTENTION 4.c.
10 " Contrary to the single failure criterion of Appendix A to j! 10 CFR Part 50, a single failure may cause loss of redundant 2 h power divisions because redundant electric power division trains 8 ll4a are electrically single interconi.
power transfer acted through two circuit breakers and switch."
18 4d 0.1: Does the IDVP know the origin of this contention? i 5C A.1: The IDVP believes that the origin is E01 8041.
6 :; Q.2: Please describe the issue identified in E01 8041.
7h A.2. The IDVP was concerned with the possibility of an im-8l! proper transfer from normal to alternate sources of electric 9 h power tnrough use of a switch common to both sources at a loca-10fition in the CRVP system. Failure of the switch could cause 11 damage to two non-mutually redundant safety-related circuits.
12 Q.3: How was that concern resolved?
13 A.3. Af ter issuance of E0I 8041, PGandE demonstrated that 14 ; its standard operating practice for transfer switches allows con-15 nection of only one of the two sources at any time. PGandE 16 l ' issued a formal operating order for DCNPP-1 which specifies its 17 l standard transfer procedure. Therefore, the highly improbable 18 . failure of the switch will not affect both non-mutually redundant 19 circuits.
20 Q.4: Did the IDVP accept any deviations from the single 21 failure criterion with respect to a power transfer switch?
22 A.4. No. There is no deviation from licensing e
23 d requirements. Operator action ensures proper switch / breaker 24 operation. The IDVP review is reported in ITR-26.
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l! CONTENTION 4.d.
1; " Contrary to GDC 57 of Appendix A, valve operators for the .
i: isolation valves which provide the steam supply to the turbine-2 !: driven auxiliary feed pump from two of the main steam generators t have not been classified and procured as safety-related compo- ,
8.i nents." .
U-4 ': Q.1: Can the IDVP address this centention? ;
h 5 !! A.1: No.
d 6 !! Q.2: Why not?
7[ A.2: Review of the valves in question with respect to GDC 8 57 was not within the scope of the IDVP sample as described in 9 ;, the Phase II Program Plan.
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P CONTENTION 4.e. !
ll l;i "The single failure of an auxiliary relay would prevent '
n automatic' closure of the redundant steam generator blowdown iso-lation valves on automatic iritiation of the AFWS contrary to a .
2 q] Westinghouse interf ace requirement and FSAR Figure 7.2-1."
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Q.1: Does the IDVP know the origin of this contention?
4[.
5!!a A.1: The IDVP believes that the origins are E01 8047 and
. 6 SER Supplement 18, p. C.4-12.
7 j Q.2: Please describe the issue identified in E01 8047.
8l A.2. The IDVP was concerned that fatlure of a specific non-4 9 h safety grade relay (3 AFWP) would prevent the automatic closure 10 U of the steam generator blowdown valves. The IDVP questioned c
b 11il whether continued blowdown from the four steam generators during 12 a postulated accident requiring auxiliary feedwater system opera-13 j tion had been considered in the accident analysis in FSAR Chapter 14 15. ;
11 I 15 } Q.3: How was this concern resolved?
16 A.3: The ID'!P subsequently determined that various analyses .
17 i by Westinghouse of accidents requiring operation of the AFW f i 18 system had assumed that the blowdown valves are isolated.
'l 19 Results of these analyses were forwarded to the NRC (letter of 20 h October 9, 1980, Crane of PGandE to Schwencer of NRC). The 21 ! assumption of blowdown isolation is used in the analysis which e
22 ] supports the conclusions of FSAR Sections 15.2.8 and 15.2.9.
il 23,: However, PGandE satisf actorily demonstrated that, for the acci-24 !> dents described in FSAR Sections 15.2.8 and 15.2.9, if protection ii 25 i: systems do not initiate a diverse signal to trip safety-grade 26 j blowdown valves, adequate auxiliary feedwater flow exists assum-L 27 j ing both a single failure of one AFW train and blowdown valves U
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1 unisolated. In other Chapter 15 accident scenarios requiring 2 ;i - auxiliary feedwater, the IDVP verified that the blowdown valves t ,
U j would be tripped closed by safety-grade trip signals from diverse ,
a 4 4 sources (such as. safety injection). In these cases, the blowdown 5 . valves would receive diverse safety-grade trip signals and close 6 in accordance with the Westinghouse accident analysis assump-7htions.
11 8j 0.4: Did the IDVP accept deviations from applicable licens-9!.ingcriteriarelatingtothefailureofanauxiliaryrelay?
l 10 !e ! A.4: No. As stated in ITR-27, the IDVP concluded that no II 11[i safety limits or licensing commitments have been violated with 12 regard to the ability to mitigate accidents or remove decay heat i
13 h and cool down the plant described in Chapter 15 of the FSAR.
0 14 " FSAR Figure 7.2-1 is a functional drawing which depicts the 0
15 jl signals that close the blowdown valves; in the IDVP's opinion it n
16 does not reflect a Westinghouse interface requirement for 17 _ redundant relays. This interpretation has been confirmed by a 18 F letter from Westinghouse to PGandE dated September 6,1983. In 19 0 the opinion of the IDVP, the Westinghouse interface requirements 20 are satisfied.
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pCONTENTION4.f. i
- 1 16 " Contrary to NUREG 0588 regarding environmental qualifica- '
h tions, flow transmitter FT-78 and flow control valve FCV-95 are 2 d located in a harsh environnment but were not listed as such in .
ll the PG&E Environmental Qualification Report air:ed September 1981, 8 y and are not yet environmentally qualified."
4j Q.1: Does the IDVP know the origin of this contention?
6 A.1: The IDVP believes that the origin is E01 8052.
6l Q.2: Please describe the issue identified in E01 3052.
I.
A.2: The IDVP expressed concern that the specified 7]
8 ij transmitter and valve were not qualified for a harsh environment.
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Q.3: How was this concern resolved?
10 ! A.3: PGandE responded that the identification tag for FT-78 11 g was changed to FT-200. FT-200 is listed in the Environmental n
12 Qualification (EQ) Report. PGandE also indicated that this flow v
13 n transmitter is qualified for a harsh environment based on the 14 vendor's report 'which justifies operation pending completion of j 15 i the qualification progrsT.. PGandE stated that the item is in the ,
16 l vendor's on-going qualification program and that qualification i
17]documentationwillbeaddedwhencomplete.
i 1 18 h In addition, PGandE responded that the EQ Report fails to >
c I 19 d identify FCV-95 as being in a harsh environment but that, in 20 f act, it has been qualified for a harsh environment. PGandE pro-21 vided its component evaluation report to document its approval of 22 c the vendor's qualification testing.
ij j 23 j The NRC's DCNPP-1 SER, Supplement No. 15, states that it l 24hperformed a 100% review of PGandE's equipment qualification 25 program and found that it meets regulatory requirements. The y.
l 26 U SuDplement also records that NRC acknowledges and accepts the 28 Y 99 4-11 I"
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Il fact that equipment can be conditionally qualified or that addi-2 ;j tional information may be needed to complete qualification.
8: Q.4: Did the IDVP accept deviations from the licensing 4
4jcriteria i relating to the environmental qualification of this i:
5 transmitter and valve?
6: A.4: No. On the basis of the ICV?'s review (reported in 7 .nl ITR-27) and the NRC SER assessment, the IDVP concludes that there 8 are no deviations from licensing criteria.
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j! i ij CONTENTION 4.g. .
h 1!! " Contrary to the requirements of NUREG 0588 regarding i i! environmental qualifications, portions of the CRVPs were omitted f 2 i from PG&E's Environmental Qualification report."
i i 3h Q.1: Does the IDVP know the origin of this contention? '
4 ; A.1: The IDVP believes that the origin is E01 8056.
5[ Q.2: Please describe the issue identified in E01 8056. i 6 .A.2: The IDVP was concerned that some CRVP system Clas; IE 7 j; equiment had not been identified in the EQ Report.
8 Q.3: How was this concern resolved? -
9 ll A.3: Further verification determined that some Class IE 10 yl equipment for the CRVP system was listed not in the EQ Repret i.
11 because that list was compiled prior to preparation of the ,
12 h schematic drawings showing the modified system. However, the 13 , equipment in the CRVP system which was not listed in the EQ 14 7
Report will operate in and was designed for a mild environment.
15 The IDVP identified no CRVP Class IE equipment not listed in the '
h 16 EQ Report which is required to be qualified for a harsh environ- ,
17hment. The'IDVP review is reported in ITR-26. -
1 18 Q.4: Did the IDVP accept deviations from applicable licens-19 ing criteria relating to environmental qualification of portions 20 . of the CRVP?
21 A.4: No. The IDVP concluded 'that the CRVP equipment meets 22 > the environmental qualification requirements of NUREG-0588 and no l
23 [ deviation from licensing criteria exists.
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. .- g 0 CONTENTION 4.h.
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" Contrary to PG&E's September 14 and December 28, 1978 i licensing commitments, CRVPS equipment identified in the FSAR as 2 h necessary to maintain control room habitability during safe shut-r down has not been evaluated regarding the effects of a moderate 3yenergypipebreak."
b 4h Q.1: Does the IDVP know the origin of this contention?
5i[ A.1: The IOVP believes that the origin is E01 8050.
H 64 Q.2: Please describe the issue identified in E0I 8050.
6 7F A.2: The PGandE letters to the NRC referenced in Contention 8 j 4.h. described the moderate energy line break (MELB) evaluation 9 [; performed for DCNPP-1. In E01 8050, the IDVP recorded its con-10 cern that those letters did not identify the CRVP system among 11 I those being evaluated.
12 Q.3: How was this concern resolved?
13 F A.3: In response to E01 8050, PGandE prcvided the IDVP with 14 [ an evaluation of the effects of MELB on the CRVP system. This 15 evaluation indicated that a MELB could cause loss of one CRVP 16 system train. An assumed single failure of the redundant CRVP 17 system train could then degrade control room habitability.
18 However, if the control room should become uninhabitable, the 19 capability for plant shutdown and cooldown would be available i
20 from the hot shutdown panel.
21 ' The IDVP verification confirmed that, in the unlikely event r
l 22 that a NELB caused the control room to become uninhabitable, L
23 [ plant shutdown and cooldown capability could be maintained from l 24 " the hot shutdown panel. PGandE stated that the CRVP system was 25 p not incidded in the MELB evaluation because of its conclusion
- 26 that there was no need to evaluate the CRVP system since, even if
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i i i 1Ij the system failed, the plant could be shut down from the hot 2 shutdown panel. The IDVP judged that the PGandE conclusion was 8'l The IDVP review is reported in ITR-
- l reasonable and acceptable.
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5 j' Q.4: Did the IDVP accept a deviation from applicable .
tg 6 licensing criteria relating to evaluation of CRVP equipment 7j; regarding the effects of an MELB7-8 fl A.4: No. The IDVP concluded that there is no deviation i
9 :'i since plant safe shutdowr; capability is not impaired as a result h
10!i of ELBs which could affect the CRVP system.
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- CONTENTION 4.1.
Iq "The fire protection for the motor driven AFW pump room is
! not consistent with -the PG&E licensing commitment for fire zone 2[ separation as stated in its November 13, 1978 Supplemental infor-i mation for Fire Protection Review ("SIFPR") in that:
3i j 1. There is a large grated ventilation opening in the 4 ;, ceiling of the room; p 2. a fire damper has gaps when it is closed."
5a i,
6 P, Q.1: Does the IDVP know the origin of these allegations?
ll 7[ A.1: The IDVP believes that the origins are E01s 8038 and u
8H 8037.
h 9[ 0.2: Please describe the issue identified in E01 8038.
10' A.2: The IDVP's concern was whether a ventilation opening 11 !.! was clearly identified in the FSAR definition of the barrier 12hbetween the fire zones here involved (FSAR Amendment 51, p.
13 ' 4-18).
14 Q.3: How was the concern resolved?
A.3: E01 8038 was issued because the FSAR langi:dge was 15 l 16 subject to misinterpretation if taken literally. However, review 17 j, of postulated credible fires indicated that a fire in one zone i
! will not propagate through the opening to the other zone.
18 . - Thus, 19 y the SIFPR licensing commitment that a fire will not propagate
! h
! 20 F from one fire zone to another is satisfied and the requirement t
q 21 p (FSAR, Amendment 51, p. 5-4) that plant safe shutdown is not 22[hinderedismet. The IDVP review is reported in ITR-18.
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! 23 L Q.4: Please describe the issue identified in E01 8037.
q 24 g A.4: In E01 8037, the IDVP's initial concern was whether
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p 25 g fire damper FD-24 was UL qualified, and that it had air gaps.
26 f Q.5: How was this concern resolved?
D 27 ( A.5: Subsequent IDVP verification determined that the fire h
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l' ' damper - is. UL qualified and that the damper gaps satisfy vendor I; t
-2 The IDVP review is !
- design' and UL qualification requirements. .
8 reported in ITR-18. I 1
4 Q.6: Did the IDVP accept deviations from fire protection j 5 licensing criteria?
6I A.6: No. The IDVP concluded that no deviations from ,
1 7,!licensingcriteriaexist. <
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s lI CONTENTION 4.J.
I!
Ig "The fire protection for the AFW pump room is not consistent w ith the PG&E licensing commitment for cable separation as stated 21 in its SIFPR of November 13, 1978 in that:
- 1. the pumps for the motor driven AFW pumps and the con-8 [i trol circuitry for a flow control valve necessary for j operation of the turbine oriven AFW pump are located in 4 ll a single fire zone; p- 2. cables for some AFW circuits are not routed in accord 50" with descriptions in the SIFPR and four AFW circuits
' PG&E consnitted to identify and review in the SIFPR were 6h not included in that document."
7 l Q.1: Does the IDVP know the origin of this contention?
8 A.1: The IDVP believes that the origins are E0Is 8019 and di 9L 8021.
i; 10 ii Q.2: Please describe the issue identified in E01 8019.
11 A.2: IDVP was concerned that circuits for the motor-driven 12 O AFW pumps and the control circuitry for a flow control valve e
18 f (FCV-95) necessary for operation of the turbine-driven AFW pump p
14 [ were located in a single fire zone.
15 h Q.3: How was this concern resolved?
4 16 ;j A.3: Further verification determined that control circuitry 17 for FCV-95 was not located in this fire zone. Therefore, no vio-18 lation of separation requ % ments occurred and a single fire 19 cannot prevent proper operation of the AFW system. The IDVP 20 .I review is reported in ITR-18.
1 21 Q.4: Please describe tne issue identified in E018021.
22 /
A.4: The IDVP's concern was that as-built circuit routings h
23 I were different than indicated in the SIFPR.
24 Q.Si Was that correct?
25 A.5: Yes. Cables had been re-routed subsequent to issuance 26 i ; of the SIFPR.
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1!! Q.6: Did the re-routing violate the requirement for cable 2 separation? !
8l A.6: No. The IDVP subsequently field verified that all AFW i 4, system as-built circuit routing conformed with the licensing com- !
y l 5 mitment for separation, with the exception that the FCV-95 DC 6' circuit was improperly located at that time. However, the FCV-95 7, circuit was then in the process o'f being re-; outed for reasons ,
8l other than fire protection considerations and installation was i i i 9' not complete. That installation was subsequently congleted and i 1 10 }the IDVP verified that the completed routing of the FCV-95 l The :
11 !IJ circuit also conforms with licensing separation comitments.
12 1 IDVP review is reported in ITR-18.
13 ij Q.7: Did the IDVP accept a deviation from cable separation h
14 d requirements for the AFW system?
15 ll A.7: No. There is no deviation from separation require-p 16 , ments and a single fire could not prevent proper operation of the i
17 L AFW system.
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CONTENTION 4.k.
t 1q " Contrary to the licensing commitment set forth in its SIFPR ,
h of November 13, 1978 each of the three 4160 volt cable spreading 2 !. rooms has a ventilation opening leading up to the 4160 volt
- switchgear rooms."
8 [I 4 Does the IDVP know the origin of this contention?
Q.1:
t 5h A.1: The IDVP believes that the origin is E0I 8039.
Il 6
Q.2: Please describe the issue identified in E01 8039.
7 A.2: The IDVP's concern was whether a ventilation opening 8 was identified in the FSAR definition of the barrier between the 9 fire zones here involved (FSAR Amendment 51, p. 4-45).
i 10 i Q.3: How was this concern resolved?
11 } A.3: E01 8039 was issued because the FSAR language was 12 subject to misinterpretation if taken literally. Further review i
13 h of postulated credible fires indicated that a fire in any of li 14 q these rooms would be unlikely to propagate through the opening to 15 J the other fire zone and that, even if a fire did propagate 16 9 through the opening, it would affect only one vital bus and safe '
d 17 shutdown capability would not be affected. The IDVP review is 16 ' reported in ITR-18, 19 L Q.4: Did the IDVP accept deviations from fire protection 20 , licensing criteria?
21 A.4: No. The IDVP concluded that a single fire in any of 22 the 4160 V cable spreading or switchgear rooms would not 23 f; adversely affect plant safe shutdown capability. No deviation i
24 h from licensing criteria exists.
4 25 :
1 26 H H
27 j 28 5 u 4-20
C01tTDITISI 4.1.
1 " Contrary to FSAR Section 3.6, possible jet impingement
] loads have not been considered in the design and qualification of 2 j. safety-related piping and equipment inside containment."
s Q.1: Does the IDVP know the origin of this contention?
4 A.1: The IDVP believes that the origins are EDI 7002 and 5i SER Supplement 18, C.4-29.
6 Q.2: Please describe the issue identified in E01 7002. ,
i 7 A.2: The IDVP's concern was that no objective evidence j
s could be found in the PGandE files that the effects of jet !
! 9 impingement on components inside containment were considered.
l 10 Q.3: How was this concern resolved? ,
11 A.3: The DCP performed and documented a reanalysis cf the I il jet impingement effects of HELB inside containment. The IDVP 12 llf la verified the DCP efforts on a sampling basis to ensure that the l
.i ;
14 I DCP sufficiently documented its jet impingement reanalysis such ;
l 15 that- the concern of E0I 7002 could be resolved. The IDVP ;
! 16 j verification included review of the DCP reanalysis procedure; j 17 [j review of the DCP field review, including an independent walkdown i 18 . to verify DCP _ identification of jet impingement interaction with 9
~
19 safety-related targets; and review of DCP safety evaluation for 20 jet impinged targets. The IDVP verification is reported in ITR- ,
21j 48. l 1 L !
i The IDVP sample verified that assumed failure of
' 22 Cl : ,
28 j instrumentation, instrument tubing, electrical components and .
- 2.; j electrical conduits identified as jet impingement targets did not negate' the ability to mitigate the effects of the specific jet i 25 26 .impingment from the causing HELB. For safety-related structural ;
.. t p
28 4-21 l
i l _ _ _ .. . . _ .
_ _ _ _ .. , _ . _ . . , , _ . - _ _ . _ . _ . . , . , - _ _ . ~ . _ _ _ _ _ .
m . _ _ _ _ . . _ _ - . . -_ ._ _ _ _ _ _ . _ - _ _ _ _ .
. . 11 - !
ii !
i P !
1y and mechanical jet impinged targets, including pipes, piping and 2 equipment supports, structural beams and columns, concrete walls 8! and floors, the IDVP' verified that the DCP had applied the h !
4 . applicable threshold loads, if any, in accordance with the DCP l 5: jet ingingement reanalysis procedure and criteria or, as j 6 required, that the.DCP had identified these targets for further 7l analyses. These analyses would determine the structural adequacy 8p of these targets to withstand the loads from the postulated HELB l
9hjet impingement. The IDVP determined that it did not need to 10j review these further analyses to resolve E01 7002, because E01 t
11j 7002 did not identify detailed structural and pipe load analysis y
12 L as a concern.
d is , In the SER, Supplement 18, the NRC identified impingenent 14 loads on safety-related piping and equipment inside containment 11, j; as an open item. PGandE has responded to this open item in a ,
1 16 P letter to the NRC, dated September 9, 1983. This letter 1 0
17 4 summarizes the specific DCNPP-1 licensing commitments of the FSAR 18 and PGandE's compliance with those comitments. ,
t 19 1 Q.4: Did the IDVP accept any deviations from licensing 201 criteria with respect to jet impingement resulting from HELBs 21 inside containment? ;
22 A.4: No. The IDVP concluded that the DCP reanalysis pro-i 23 j! cedure- and criteria met the licensing commitments of FSAR, 24 .; Section 3.6, constituted a comprehensive technical review i 25 program, and documented the technical approach and the results. i s
26j a
27d a
28 1 3
1 4-22
-0 ;
1 l l;} CONTENTION 4.0.
" Contrary to the requirements of NUREG-0588 regarding ihf environmental qualifications, safety-related cables and cable 21' splices which could be subject to a harsh environment during a high-energy line break are not identified in the PG&E Environ- '
8 .
mental Qualification Report."
4 '
Q.1: _ Does the IDVP know the origin of this contention?
~a b' A.1: The IDVP believes that the origins are E01s 8011 and 6: 8044.
7 Q.2: Please describe the issue identified in E01 8011.
8i A.2: The IDVP's concern was that some safety-related cables 9i in the AFW and CRVP systems were not identified as environmental-10 ly qualified in Reference 5 of Appendix 3.6 of the FSAR.
11 lj Q.3: How was this concern resolved?
12 ] A.3: PGandE subsequently provided documentation showing n
13 h that all safety-related cables used in the AFW and CRVP systems 1
14 h are environmentally qualified. It should be noted that reference !
15 5 had been prepared in 1975. The IDVP verified that cable not {'
i 16 l listed in that document was purchased after 1975, qualified to I p
17 d the temperature defined in Appendix 3.6 of the FSAR, and included i 18 in the plant EQ Report. The IDVP review is reported in ITRs-21, ,
19 -25 and -26.
20j ,
g,4: Please describe the issue identified in E01 8044.
ii 21 F A.4: The IDVP's concern was that, although Reference 5 of 22 FSAR Appendix 3.6 states that "in general splices were not used,"
23 the IDVP found splices which could potentially be exposed to the I 24 l effects of HELCs. l 1
25 ;! Q.5: How was this concern resolved?
26 h A.5: In response to E01 8044, PGandE provided documentation !
r ;
27 ! showing that the splices in question are qualified to 340 F. The i h
- 28 , ,
L 4-23 il o '
. . ij l r
,k l: iIDVF verified' that no splice in the AFW or_ CRVP system is exposed ,
a 0
2 h to jet temperature in excess of 340 F and that splices are f n
8 f addressed in the EQ Report. The IDVP review is reported in ITRs-4l 25 and -26.
t
( i 5 !, Q.6: Did the IDVP accept deviations from licensing criteria :
a 6 'I with regard to environmental qualification of cables and cable 1
7;f splices?
8)} A.6: No. The IDVP concluded that all cables and splices in '
l 9
lthe AFW and CRVP systems are environmentally qualified in ,
10 l accordance with NOREG-0588.
11b-
- t 12 ((
^
13j g
14 d
15 !
l 16 fj l l
17 lj 18 19 20 q 21 .
22
$i 23 n 1
24 1 25 li 26 ,
27 1 e
h F
4-24 ii
J l CONTENTION 4.p.
f "The NSC pipe break analysis, which is Appendix A to FSAR li Section 3.6, did not include all likely sources of water in the j 2 ], calculation of flooding levels."
l 8
q Q.1: Does the IDVP know the origin of this contention? *
-4 l A.1: The IDVP believes that the origins are E01s 800s and I
5 '
8040.
i 6 f Q.2: Please describe the issue identified in those E01s. l 7h A.2: The IDVP's concern was that the water levels d
8 'j; calculated to occur in area GE/GW at elevation 11s feet, 0 inches i
9 p dua to a feedwater HELB were not conservative.
1 ,
10 '; Q.3: How was this concern resolved?
I 11 u
! A.3: Further detailed review of the PGandE calculation i 12 revealed the maximum flood heights which it calculated for the 13 area in question are conservative because PGandE used conserva- l 14 tive assumptions and methods. Although the volume of water below ;
y l
15 [I the feedwater sparger ring (inlet pipe) in the steam generator l 16 ij and from the AFW system were neglected in PGandE's calculation, 17 y the IDVP determined that other sources of water were cverpredict-18 ed by substantially greater amounts. This resulted in conserva- r 19 l tive predictions by PGandE of water release volumes and flood 20 -
heights. The IDVP review is reported in ITR-14.
21li Q.4: Did the IDVP accept deviations from the licensing il 22 1 criteria for flooding?
23 A.4: No. The licensing criteria of FSAR 3.6, Appendix A 24 ' are met.
25 26 h '
i 27 d 28 j 4-?5 i
!I i
b
. . li CONTENTION 4.q.
1Y " Contrary to PGLE's December 28, 1979 licensing commitment
- i. letter to the NRC, modifications to protect two Auxiliary 2" Feedwater valves from the effects of moderate energy line breaks l were not implemented."
3 ,
a 4 !! Q.1: Does the IDVP know the origin of this contention?
i i
5 A.1: The IDVP believes that the origin is E01 8014.
6! Q.2. Please describe the issue identif 2ed in E0I 8014.
1 7L A.2. The IDVP's concern we that two valves (FCV-436 and 8 FCV-437) identified in the December 28, 1979 letter as requiring 9 protection from the effects of MELBs did not in fact have spray 10 i shields installed.
L 11 ! Q.3: How was this concern resolved?
b A.3: PGandE demonstrated that, subsequent to the 12 d 13 6 December 28, 1979 letter, it had determined that these two valves 14 [ are not required to operate in mitigation of the effects of a F
15 ' MELB. PGandE therefore decided not to install the spray protec-16 : tion devices. PGandE will revise the December 28, 1979 letter to 17 indicate that protection for these valves is not required. The 18 : IDVP performed an evaluation and determined that the valves are 19 in fact not required to operate to mitigate tne effect of an 20 ! MELB. The IDVP review is reported in ITR-21.
21 Q.4: Did the IDVP accept deviations from applicable licens-22 ing criteria with respect to protection from MELBs?
d 23 : , A.4: No. There is no deviation from applicable licensing 24 i criteria.
25 , ,
6 27 28 4-26
o l
l CONTENTION 4.r. !
11- " Contrary to the licensing _comitment to maintain minimum
!! system ' redundancy as stated in FSA3 Section 3.6A (NSC evaluation i 2 l of pipe break outside containment), four components were identi- ;
' fied for which high energy line cracks could cause temperatures i 3 in excess of the specification temperatures of the components." ;
i 4 Q.1: Does the IDVP know the origin of this contention?
5 A.1: The IOVP believes that the origins are E01s 8028, !
8029, 8030 and 8031. !
6l 7 Q.2: Please describe the issues identified in E01s 8028, 8 8029 and 8030.
9! A.2: The IDVP identified inconsistencies in Appendix 3.6 of 10 the FSAR and Reference 5 of this Appendix. The Appendix states
~11- (pp. 3.6A-22 and -23 (Revision 3)) that HELBs (or HELCs) need not 12 be postulated in line 760 downstream of FCV-95. The Appendix (at 13 j. pp. 3.6A-68 and 3.6A-82) and Reference 5 (Table B-13) indicate '
a 14 I that breaks were postulated in line 760 downstream of FCV-95.
15 a Q.3: How was this issue resolved?~
16 l' A.3: The applicable NRC requirements are found in a letter i
17 from Mr. Giambusso of the NRC to Mr. Searls of PGandE dated 18 o December 18, 1972. The IDVP verified that the letter does not U
19 ' require postulation of HELBs or HELCs in line 760 downstream of 20 g FCV-95. PGandE has therefore, committed to correct the incon-21Nsistencies in the FSAR by eliminating any such postulation.
22 Therefore the equipment identified in these E01s will not be d
-23 exposed to harsh environments. The IDVP review is reported 'in 24 ITR-21.
p 25 g Q.4: Please describe the issue identified ir- E0I 8031.
26 h A.4: The IDVP's concern was that an HELC in line 594 could i
27 i adversely affect AFW system equipment.
28 d h 4-27 0
i l'
1 j Q.5: How has this issue been resolvedi t i 2 A.5: Further verification indicated that a crack in line i 8 594 would not cause turbine generator _ or reactor trip. Thus, in 4 accordance with the FSAR criteria one need not assume loss of
- 5) offsite power and the AFW system is not required to operate to :
6 mitigate the effects of the HELC in line SM or to achieve plant
- I 7 Il shutdown. Therefore, the AFW equipment identified in this E01 4 ji 8 j will not be required to operate to mitigate the effects of the 1
9 l HELC in line 594 or to safely shut down the plant. Accordingly, 1
10 j there is no safety consequence if the AFW equipment is exposed to i
11 ) HELC impingement temperatures. The IOVP review is reported in a
12 j ITR-21.
II '
13j Q.6: Did the IDVP accept deviations from applicable licens-14 h ing criteria regarding HELCs?
U 15!! A.6: No. There is no deviation from FSAR Appendix 3.6.
16'[ Minimum system redundancy is maintained and equi; ment required to h
17 : mitigate the HELC and safely shut down the plant will not be 11 18 0 exposed to temperatures in excess of its specification tempera-19 ture.
20 b t
21 22 [
23 ;;
9 24 25 i; d
26 ;j 27d
-28 4-28
o . .. 9 jCONTENTION4.s
\
1p " Contrary to the licensing consnitment to maintain minimum U system redundancy as stated in FSAR, Section 3.6A (NSC evaluation '
2yof pipe break outside containment), a conduit was identified ;
y whose failure due to a high energy line crack could eliminate i 8 j h!redundantAuxiliaryFeedutersystemflow."
m .
4! Q.1: Does the IDVP know the origin of this contention?
- I A.1: The IDVP believes that the origin is E018049. .
5 i j!
6y Q.2: Please describe the issue identified in E0I 8049. -
i.
- 7 i'; A.2: The IDVP's concern was that the effects of an HELB jet 8 h on the AFW system conduit had not been considered.
9; Q.3: How was this issue resolved?
10 j A.3: The IDVP verified that the cable in the conduit ;
11] identified in the E01 will not be damaged due to the effects of 12 ! an HELB and that AFW system redundancy is not affected. Jet 13 !'!! pressures on the conduit are lower than allowable conduit jet 14 [ pressure. Cable in the conduit is qualified to 540 F, which is 15 above the enveloping temperature using either the FSAR 16 h methodology or ANSI-ANS 58.2 methodology. The IDVP review is n
17 reported in ITR-23.
18 h In addition, since a break in line 594 will not cause a 19 i; turbine-generator or reactor trip, the FSAR does not require an 6
20 [ assumption of loss of offsite power and AFW system operation is 21 ' not required to mitigate effects of the break or to shut the 22 plant cown. Therefore, the AFW system is not required to 23 mitigate the HELB in question or to safely shut down the plant.
24 n Q.4: Did the IDVP accept deviations from licensing criteria 0
25 % in evaluating HELB effects on AFW system conduits?
26 e A.4: No. There is no deviation from licensing criteria.
27 ?
Y 28 H 4-29 L
a
1 1 o . .
, CONTENTION 4.t..
1 ! " Contrary to the FSAR Section 8.3 comitment to provide !
, switchgear buses with adequate short circuit interrupting 2 capability, the calculated duties for circuit breakers on 4160 V !
buses F, G, and H were above the nameplate ratings for those !
8 buses."
4
.Q.1: Does the IDVP know the origin of this contention?
5 I A.1: The IDVP believes that the origin is E018022.
e h
6 Q.2: Please describe the issue identified in E01 8022. ;
7 A.2: The IDVP's concern 'was that the circuit breaker's i
8 current-interrupting rating less than the l nameplate was ,
8 ; ; calculated current-interrupting duty required, n
10 Q.3: How was this concern resolved?
11i A.3: In response to E01 8022, PGandE provided the 12 ) manufacturer's verification that the 4160 V circuit breakers are 4
13 capable of interrupting the maximum available short circuit 14
! current. This conclusion is based on tests performed by the man-it
.15! ufacturer in 1976. The IDVP review is reported in ITR-24.
I 16 Q.4: Did the IDVP accept a deviation from licensing 17 criteria in connection with short circuit current-interrupting e '
18ll capability?
t 19 A.5: No. The IDVP concluded that, since the breakers will 20 $ interruot the calculated short circuit current, no deviation from ,
21 j! licensing criteria exists.
22 il 23 i!
d 24 b 1
25lF 26 j!
27 f c
4-30 i
l5
+ . ,
l l
l CONTENTION 4.u. i i
L" Contrary -to single failure criteria stated in FSAR Section ' j
-1[l3.1.1,reviewsoftheAuxiliaryFeedwater'andControlRoomVenti-i
- 2 l i lation and Pressurization systems identified circuit separation j,and single failure ' deficiencies. .Similar deficiencies were
.8 ' identified. in additional verification reviews,- which included
.other safety-related systems."
~
4 1 i i 5} -Q.1: Does the IDVP know the origin of this contention?
y 1 .
6 A.1: The IDVP believes that the origins are E0Is. 8017 and l I
-7i E8057.
i
+
8[ Q.2: Please describe the issues identified in E0Is 8017 and r
9 L 8057.
10 A.2:'- The IDVP. review of the initial sample systems (AFW and 11 CRVP systems) identified Class IE electrical control circuits in 12 l enclosures (i.e., panels and- termination boxes) that were not
~ 13 separated by the methods listed in the FSAR, Section 8.3.3. This ;
I 14 ' issue is the ' subject of E0I 8017. The IDVP also identified an j 15 electrical control transfer switch in the CRVP system to which !
16 [ mutually -redundant . Class IE power sources were connected such 3
17 1 that the DCNPP-1 single failure criterion was not satisfied.
i 18 [ This issue is the subject of E018057.
U Q.3: Were these issues generic?
20 h A.3: The IDVP believed the issues were generic.
21 Q.4: How were these concerns resolved?
i 22 ;j A.4: Since these. concerns were considered by the IDVP to be 0
23!! generic, the DCP. reviewed all PGandE-designed safety-related 24 systems for. similar _ concerns. The DCP review identified all i
25 h mutually redundant Class IE circuits and devices within the same 26 j enclosure that -required separation. The DCP .then conducted a 27 " field review of all those identified circuits to ensure that they n
28 !,
L
" 4-31
. :f i y
. - - _-.. . . - . . . . _ . . . _ - _ , _ . _ _ . . _ . - _ . . _ . _ . . _ _ - _ - . _ . - . ._ - , . . _ - ~
i o 4 ..
g l
1 are separated by the methods stated in the FSAR, Section 8.3.3.
2 lAlso, as part of .the review, the DCP identified electrical si 3 h devices which have mutually redundant circuits connected to them.
4 If such an electrical ' device was identifed, a single failure 5 !i analysis was performed by the DCP to establish the ability of the 6 system to perform its' design basis function.
7 Upon completion of the DCP' analysis, the IDVP selected four 8 ij of the PGandE-designed safety-related systems as samples to fverify the DCP review. The IDVP reviewed each sample system's 9
10 circuit drawings to determine whether the DCP had correctly 1
11 d identified the mutually redundant circuits. The IDVP then per-a 12 / formed a field inspection of enclosures included in tne sample 13 ! systems to determine whether the installation of mutually 14 i redundant circuits within the same enclosure met the separation 4
15 E criteria committed to in FSAR Section 8.3.3. The IDVP verifica- .
16 tion did not identify any cases where the DCP had failed to 17 identify a system's mutually redundant circuits or where separa-18 l tion in the field did not meet FSAR requirements. The DCP had 19 identified modifications that were required to be made in the 20 : sample systems. The IDVP field verified that those modifications 21 had been implemented such that compliance with the FSAR separa-22 Ltion criteria exists.
23 , In addition to verifying circuit separation, the IDVP also 24 0 reviewed the DCP's single failure analyses in the four sample 25 qsystems. The DCP provided drawings marked to show where mutually In those 26 .! redundant circt.ns were connected to the same device.
27 cases, the DCP performed a single f ailure analysis to establish 2e 1
4-32 i
o*. !! {
4 l f
1 .
the ability of the system to perform its design basis function. j 2 The IDVP reviewed these drawings &nd analyses to verify that all l
'i I 3 ! mutually redundant circuits connected to the same device had been t
4 ll identified. The IDVP verification did not identify any case 5 where a single failure could adversely affect the operation of a i l
6 ijsamplesystemandnomodificatonswererequired. The IDVP review l i
t 7 j is reported in ITRs-27, -28 and'-49.
~
.i 8
ll Q.5: Did the IDVP accept deviations from licensing criteria il 9 llwith respect to circuit separation and single failure criterion i!
10
]asappliedtosafety-relatedsystems?
'I A.5: No. The concerns identified in E01s 8017 and 8057 11 f
12 ; have been eliminated in all PGandE-designed safety-related b
13 ! systems. - Circuit separation and single failure requirements of 14 tlFSAR, Sections 8.3.3 and 3.1.1, are met. .'
15 -lu! ;
16 h 17 !
g 18 0 19
. 20 - !
n o
21 'l 22 d
23 !!
1 24 d
25 .
d 27 28 4-33 4
og AFFIDAVIT OF WILLIAM E. COOPER The undersi;..cs, William E. Cooper, this 12th day of October, 1983, upon his oath states that the attached Resume is a true and correct statement of his education and professional experience, w
0 William E. Cooper October 12, 1983 jhY- /&. ..
Notary Public lliLLMM G, pgcoggg fCTai;Y .eUCLic 7 CCTJ?"?ECil EXi3fRES AL'3UCT G,1007 -
U 1%T M ENGNEERING SERVCES ATTACHMENT 1 DR. WILLIAM E. COOPER Consulting Engineer Resume Education Stevens Institute of Technology (1941-1943)
Oregon State College, U.S. Army, Spec. Training, M.E. 4-7(1943-1944)
Oregon State College (1946-1948):
B.S. in Mechanical Engineering (1947)
M.S. in Mechanical Engineering (1948)
Purdue University (1948-1951):
Ph.D. in Engineering Mechanics (1951) - "
Honors Fellow, American Society of Mechanical Engineers (1972)
Purdue Disti>guished Engineering Alumnus (1973)
Certificate of Appreciation, Pressure Vessel Research Connittee (1977)
The William M. Murray Lectureship, -
Society for Experimental Stress Analysis (1977)
B. F. Langer fleclear Codes and Standards Award, American Soc:ety of Mechanical Engineers (1978)
Centennial Award, American Society of Mechanical Engineers (1980)
Pressure Vessel and Piping Medal, American Society of Mechanical Engineers (1983)
Sigma Xi (Research), Pi Tau Sigma (Mechanical Engineering),
Sigma Pi Sigma (Physics)
Who's Who in: America; Ingineering; Atoms l American Men and Women of Science Registered Professional Encineer l Indiana (1952), New York (1958), Massachusetts (1963)
Membership American Society of Mechanical Engineers Society for Experimental Stress Analysis Atomic Industrial Forum Addresses ,
l Business: Teledyne Engineering Services 130 Second Avenue Waltham, Massachusetts 02254 (617) 890-3350 Home: 83 Fifer Lane Lexington, Massachusetts 02173 (617) 861-7007 l
l
TN
.- ENGINEERNG SERVICES DR. WILLIAM E. COOPER Consulting Engineer Employment Teledyne Engineering Services (formerly Teledyne Materials Research or Lessells and Associates, Inc.):
Consulting Engineer (1976-)
Senior Vice President & Technical Director (1974-1976)
Vice President & Manager, Engineering (1MS-1973)
Engineering Manager (1963-1968)
Consulting engineering services in the design and analysis of mechanical systems and structures, primarily for energy conversion.
Massachusetts Institute of Technology, Lecturer, Reactor Safet*y (1975-)
Electric Power Research Institute, Consultant (1974-1978)
Oak Ridge National Laboratory Advisory Comittee, Engineering Technology Division (1975-1978)
Advisory Committee on Reactor Safeguards, Consultant (1967-1974)
Knolls Atomic Power Laboratory, General Electric Company:
Consulting Engineer, S5G Structural Mechanics (1963)
Manager, S5G Structural Evaluation (1959-1963)
Consulting Engineer, SAR Structural Evaluation (1957-1959) .
Specialist, SAR Mechanical Analysis (1955-1977)
Engineer, SIR Mechanical Analysie (1952-1954)
Technically responsible for the structural integrity of components and systems of sodium- and water-cooled naval reactor power plants.
Union College Instructed graduate courses in Theory of Plasticity (1962 and 1963)
Purdue University:
Instructor in Engineering Mechanics (1949-1952)
Instructor in Engineering Drawing (1948-1949)
Instructed courses in drafting, statics and dynamics, experimental stress analysis, plasticity, dynamics of materials, physical metallurgy, and applied metallography Oregon State College:
Graduate Teaching Assistant in Engineering Drawing (1947-1948)
Student Teaching Assistant in Physics (1947) -
U.S. Army, Sergeant, Construction Foreman (1943-1946)
General Electric Company, Mechanical Draf tsrnan (1942-1943)
W TELEDYNE N MES
- DR. WILLIAM E. COOPER Consulting Engineer Comittee Participation American National Standards Institute:
Board of Directors (1981-)
Nuclear Standards Management Board (1975-1977)
Technical Advisory Group to TC/85, Nuclear (1974-1979)
International Standards Organization U.S. Representative to TC/85, SC/3, Nuclear Power (1976-1979)
Expert to TC/85, SC/3, WG/6 Primary Boundary (1974-)
American Society of Mechanical Engineers:
Senior Vice President and Chairman, Council on Codes and Standards (1981-)~
Vice President for Codes and Standards and Member Executive Comittee of Council (1980-1981)
Comittee on Budget (1978-1980) (Chairman 1979-1980)
Council (formerly Policy Bor.rd) on Codes and Standards (1972-)
(Chairman 3980-) ,
Nuclear Codes and StancV.s Comittee (1974-1980) (Chairman 1975-1977)
Boiler e.nd Pressure 6: ael Comittee:
Honorary Member (1W -)
Main Committee (196: 980)
Executive Comittee (1971-1976)
SC on Design (1967-1975) (Chairman 1967-1972)
SC on Nuclear Certification (1973-1977) (Chairman 1973-1975)
SC on Nuclear Power (1964-1980) (Vice Chairman 1966-1969)
Special Comittee to Review Code Stress Basis (1955-1967) 831 Code for Pressure Piping:
Mechanical Desic i Comittee (1957-1959,1963-1967)-
831.7 Nuclear Piping (1965-1971)
Code for Nuclear Pumps and Valves (1965-1969)
Metals Engineering Division, Chairman (1960)
Hudson-Mohawk Section, Chairman (1958-1959)
Atomic Industrial Forum:
Subcomittee on Materials Requirements (1981-)
Welding Research Council:
~
Pressure Vessel Research Comittee:
Main Comittee (1954-1974)
Desigr. Of vision (1954-1974) (Chairman 1969-1973)
v ENGINEERING SERVICES ;
1 1
y DR. WILLIAM E. COOPER Consulting Engineer Publications and Major Presentations j
" Determination of Principal Plastic Strains," Transactions, ASME, July 1952.
" Structural Problems of a Sodium-Cooled Nuclear Reactor," ASME, Paper No.
54-SA-25, with D. R. Miller.
" Proposed Structural Design Basis for Nuclear Reactor Pressure Vessels, Problems in Nuclear Engineering," edited by D. J. Hughes, S. McLain, C.
Williams, Pergamon Press,1954. * *
"The Significance of the Tensile Test to Pressure Vessel Design," Welding Journal Research Supplement, January 1957.
"Experiaental Determination of Stresses in the Vicinity of Pipe Appendages to a Cylindrical Shell," Proceedings SESA, XIV, 2, with F. J. Mehringer, 1957. ,
" Safeguards Aspects of Reactor Vessel Design," TheIW elding Journal Re-search Supplement, January 1958; Journal American Society of Naval Engi-neers, with D. R. Miller, May 1958.
"The Scope of Pressure Yessel Codes and Activities Towards Improved Con-tent," Preprint 78, Nuclear Engineering and Science Congress,1958.
" Design Basis for Thermal Stress," Proc. SESA, XV, 2, 1958.
" Structural Design Basis for Reactor Pressure Vessels and Associated Com-ponents," U.S. Office of Technical Services PB151987, with B. F. Langer (Westinghouse) and J. L. Mershon (BuShips), December 1958.
" Implications of Radiation Effects to Reactor Pressure Vessel Design, AEC l
Conference on the Status of Radiation Effects Research on Structural Materials and the Implications to Reactor Design," October 1959.
" Stresses in a Pipe Bent into a Circular Arc," Transactions, ASME, Journal of Engineering for Industry, 83, B, 4, with N. A. Weil and J. E. Brock, November 1961, pp. 449-459.
" Specification Guidelines for Nuclear Pressure Vessels," USAEC NY0-3416-1, with D. F. Landers, October 1964.
i
" Design Criteria for High-Pressure, High-Temperature Bolting," Nuclear Engineering a'nd Design, 8, with R. Widmer, J. A. Signorelli, R. F.
Brodrick, 1968, p. 125.
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.' ENGdNEERING SEECES t DR. WILLIAM E. COOPER Consulting Engineer Publications and Major Presentations (Cont'd)
" Interaction of Material and Design Problems in Critical Vessels," Invited Keynote Lecture, First International Conference on Pressure Vessel Tech-nology, Delft, Proceedings, Part III, 1969, p. 29.
" Construction, Rating, arJ Inservice Inspection of Test Tanks," Proceed-ings, 7th U.S. Navy Sympusium of Military Oceanography, Vol. 1, with B. H.
Schofield, 1970, p. 104.
" Experimental Efforts on Bursting of Constrained Disks as Related to the Effective Utilization of Yield Strength," ASME Paper 71-PVP-79, with E. H.
Kottcamp and G. A. Spiering. '
" Codes: Asset or Liability," Fatigue at Elevated Temperature, ASTM STP 520, 1972.
" Development and Operation of the ASME Boiler and Pressure Vessel Code,"
and "An Introduction to the Design Procedures of the ASME Boiler and Pressure Vessel Code," U.S.-Japan Joint Symposium, Pressure Vessel Tech-nology and Pressure Component Codes, Tokyo, 1973. .-
" Nuclear-Pressure Vessels and Piping-Materials: Where to Next?" ASME Joint Conference, Miami, 1974.
"A Personal Viewpoint on the Development of ASME Code Rules for Nuclear
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Components," ASME Winter Annual Meeting, 1974.
" Nuclear Vessels are Safe," Mechanical Engineering, with B. F. Langer, April 1975.
" Improving Reactor Pressure Vessel Availability by Design," Nuclear Safety, 17(1), January-February 1976.
"ASME Section XI Flaw Evaluation Procedures and Application to Nozzles," !
Nondestructive Examination Conference, Washington, 1976, also UKAEA, j Risley, 1976.
" Experimental Mechanics and Nuclear Power," The William R. Murray Lecture, 1977 Experimental Mechanics, 17, 10, October 1977. <
" Safety Evaluation of Reactor Vessel Nozzle Cracks," ASME Paper No.
78-PVP-90, with P. C. Riccardella, 1978. -
" Minimization of Safety and Reliability Concerns by Consideration of Operating Experience," Conference on the Quality of Nuclear Power Stations from American and German Viewpoints, K61n, 1978.
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TE M ENGINEER!NG SERVICES i DR. WILLIAM E. COOPER Consulting Engineer Publications and Major Presentations (Cont'd)
" Analysis of Inservice Inspection Flaw Indications," Maintenar.ce Welding in Nuclear Power Plants, American Welding Society, 1979.
"The Future - an ASME Viewpoint," ANS Executive Conference on Nuclear Power Plant Owner Certification, Washington, 1980.
" Concepts in the Design and Analysis of Welded Joints," AWS, Indianapolis, 1980.
"What Happened to Comon Sense," ASME Emerging Technologies ' Conference, San Francisco,1900.
"The Development of Ccdes and Standards for Superconducting Magnet Struc-tures," DOE-NBS Workshop on Materials at Low Temperatures, Vail, 1980.
" International Involvement of U.S. Standards," U.S. Department of Comerce Conference, 1980.
"0wner Certification," Atomic Industrial Forum Workshop on Reactor Con-struction and Operation in the New Environment, Atlanta, 1980.
"Requalification of Nuclear Class 1 Pressure Boundary Components, SMIRT Paris, 1981 (also EPRI Report NP-1921).
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.l AFFIDAVIT OF ROBERT L. CLOUD The undersigned, Robert L. Cloud, this 12th day of October, 1983, upon his oath states that the attached Professional Resume is a true and correct statement of his education an pr essional experience.
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Robert L. Cloud g October 12, 1983 naamremm OFFICIAL SEAL PATRICIA L HOLMES count o ku e Mr Commissies Empires sept.5.1988 i wmwn '
Notary Public 1
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ATTACHMENT 2 Robert L. Cloud Associates, Inc.
ROBERT L. CLOUD PRINCIPAL Professional Resume Education Texas A & M College BSME 1956 Texas A & M College MSME 1957 Univ. of Pittsburgh Phd ME 1964 Experience 1979 to Present: Robert L. Cloud Associates, Inc.,
Berkeley, Ca. Design Criteria, Seismic design and analysis, Piping design criteria, Piping analysis, Project management, Failure analysis .
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1978-1979: Engineering Decision An'alysis Co., Palo Alto, Ca., Exec. Vice President. -Project management, Design criteria Failure analysis, Piping and Mech.
Equipment design and analysis.
1971-1978: Westinghouse Electric Corp., PWR Systems Divisibn; Manager of Mechanicsand Materials Technology.
Responsible for design criteria, stress and dynamic -
analysis and materials engineering for the primary
- system of Westinghouse Pressurized Water Reactor Systems.
1969-1971: Teledyne Materials Research, Waltham, Mass.
Manager Analytical Engineering, Design criteria, Analysis and research on equipment and piping, Failure analysis.
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1962-1969: Westinghouse Electric Corp., Bettis Atomic Power Lab. Stress Analysis Engineer to Manager, Mechanics and Materials Engineering, Design criteria, Fracture Mechanics Studies, Analysis and research on pressure i vessels and piping.
l 1957-1962: Westinghouse Electric Corp.. Large Rotating -
l Apparatus Division. Stress analysis and development work on large central station turbo-generators.
I 1956-1957: Texas A & M University Instructor, Mechanical Engineering.
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. l Membershio
- 1. American Society of Mechanicti Engineers a) Past Chairman, Design and Analysis Committee, PVP Division b) Past Chairman, Pressure Vessels and Piping Division c) Past Member, ASMF, Boiler and Pressure Vessel- Code, Subgroup on Openings and Attachments .
. 12 . Past member, Pressure Vessel Research Committee, WRC Lectures a
- 1. Eisenment Lectures, Fracture Control, 1970, American Society for Metals, Philadelphia, Pa.
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- 2. Teledyne Materials Research ASME Boiler and Pressure Vessel Code Seminar a) Brittle Fracture b) Nozzles, Tubesheets, & Special Problems c) Plastic Limit Analysis
- 3. Principal Division F.-Lecture,," Structural Mechanics Applied to Pressurized Water Reactor Systens",
4th International Conference on Structural Mechanics in Reactor Technology, San Francisco, California, 1977.
Publications
" Minimum Weight Design'of a Radial Nozzle in a Spherical Shell,: Transactions of the ASME, Journal.of Applied Mechanics, Vol. 32, Series E. No. 2, June, 1965.
"The Limit Pressure of Radial Nozzles in Spherical Shells" Nuclear Structural Engineering, Vol. 1, No. 4, April 1965.
l "Interpr'etive Report on Pressare Vessel Heads:, Welding Research Council, Bulletin No. 119, January 1967.
" Approximate Analysis of the Plastic Limit Pressure of Nozzles in Cylindrical Shells" with E.C. Rodabaugh, Transactions of the ASME, Journal of Engineering for Power, Vol. 90, Series A, No. 2, April 1968.
l Thermal Buckling and Frictional Effects on Postbuckling -
Behavior of Sealed Electric Liners" with J.H. Dittmar, Transactions of the ASME, Journal of Engineering for Industry, Vol. 90, Series B, No. 3, August, 1968.
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" " Assessment of the Plastic Strength of Pressure Vessel Nozzles" with E. C. Rodabaugh, Transactions of the ASME Journal of Engineering for Industry,.Vol. 90, Series B, N. 4, November, 1968. -
" Evaluation of Experimental and Theoretical Data en Radial Nozzles in Pressure Vessels" with E. C. Rodabgugh.
R. J. Atterbury, and F. J. Witt, U.S. Atomic Energy Commission, TID - 24342, 1968. ~
" Proposed Reinforcement Design Procedure for Radial Nozzles in Cylindrical Shells with Internal Pressure" with E. C. Rodabaugh, Welding Research Council Bulletin
! No. 133, September 1968.
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" Fracture Mechanics Criteria for the Prevention of Erittle Fracture in Nuclear Reactor Vessels," 1967,(Classified) with others, Bettis Atomic Power Lab., Westinghouse Electric Corporation.
" Pressure Vessel Head Design" chapter in "The Stress Analysis of Pressure Vessels and Pressure Vessel Com-Ponents" Editor, S. S. Gill, Pergamon Press, 1970. .
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" Fracture Prevention in Nuclear Plants" ASM Conference on Fracture Control, Philadelphia, Pennsylvania, 1970.
Editor, " Pressure Vessels and Piping: Design and
. Analysis", 2 Vol., American Society of Mechanical Engineers, 1972. ,
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" Dynamic Analysis of Nonlinear Pipe Whip Restraints" -
with S. Palusamy, and W. L. Patrick, Pressure Vessels and Piping Conference, Miami Beach, Florida, June 1974.
" Nonlinear Seismic Analysis of the Ice Condenser System" with W. S. LaPay, A. J. Soroka, and G. J. Bohm, Structural Design of Nuclear Plant, ASCE 1975 New Orleans, Louisiana.
" Dynamic Analysis of Structures with Solid-Fluid Inter-action" with R. R. Pedrido, A. N. Nahavandi, Transactions of the 4th International Conference on Structural Mech-anics in Reactor Technology (Smirt-4), San Francisco, California, August, 1977. ,
" Structural Mechanics Applied to Pressurized Water Reactor Systems", Vol. 46, No. 2, Nuclear Engrn. &
Design, April, 1978.
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. " Dynamic Events in Nuclear Reactors", Survival of Mechanical -Systems in Transient Environments, T. L. Geers et al, Editors, ASME AMD-Vol. 36, 1979.
" Creep Instability in Flexible Piping Joints" with R. D. Campbell and D. Bushnell, 1980. To be published. .
" Seismic Performance of Piping in Past Earthquakes:, .
Specialty Conference on Civil Engineering and Nuclear Power, September 1980, Knoxville, Tenn.
"A Sum: nary and Critical Evaluation of Stress Intensity Factor Solutions of Corner Cracks at the Edge of a Hole" with S. S. Palusamy, Welding Research Council Bulletin No. 276, April 1982.
" Interpretive Report on Dynamic Analysis of Pressure Components - Second Edition", Chapter 3, Welding Research Council Bulletin No. 269, August In81.
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AFFIDAVIT OF J0l#1 E. KRECHTIllG The undersigned, John E. Krechting, this 12th day of October, 1983, upon his oath states that the attached Resume is a true and correct statement of his education and professional experience.
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Z hn E. Krechting a,
October 12, 1983 i
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Notary Public
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ATTACHMENT 3 August 1983
- KRECHTING, JOHN E.
PROJECT ENGINEER POWER DIVISION Ei;UCATION U.S. Naval Academy - Bachelor of Science, Naval Science 1965 LICENSES AND REGISTRATIONS Professional Engineer - Rhode Island EXPERIENCE
SUMMARY
Mr. Krechting has over 18 years of experience in the engineering field.
Currently as Proj ect Engineer for the Diablo Canyon Nuclear Power Plant Independent Design Verification Program, he is responsible for the NRC required design verification to establish that installed safety-related systems meet their licensing and operational commitments.
Since joining Stone & Webster Engineering Corporation (SWEC) in July 1974 as an Engineer in the Power Division, Mr. Krechting has been assigned to positions of increasing responsibility. He has been assigned to the Charlestown Nuclear Power Plant project, which was in the design development
' and PSAR production stage; to the high temperature gas-cooled reactor (HTGR) 3,000 MWt Reference II Design Study for General Atomic Company which developed a conceptual reference plant design; and to the Sundesert Nuclear Plant project which was in the design development and PSAR production stage.
Mr. Krechting was assigned as the Principal Nuclear Engineer on the North .
Anna Power Station - Units 1 and 2 project, and subsequently as the Lead Power Engineer on the North Anna Power Station " project. He was assigned as Supervisor of the Systems Engineering Group responsible for'the development and maintenance of fluid system descriptions for the SWEC reference / standard nuclear, fossil, and industrial plants; development and maintenance of fluid system related Power Division Technical Procedures and Guidelines; and resolution of generic fluid system design problems.
Prict to joining SWEC, he was employed by Westinghouse Nuclear Energy Systems as a Senior Systems Engineer on the project to determine the feasi-bility of floating nuclear power plants. He developed the design of many of the nuclear and reactor auxiliary systems for the Offshore Power Systems' floating nuclear power plants.
His experience includes 6 years in the operation and maintenance of U.S.
Navy submarine nuclear power plants, including two years as the Chief Engineering Officer of a nuclear submarine power plant. -
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O DETAII.ED EXPERIENCE RECORD KRECHTING, JOHN E. 50109 STONE & WEBSTER ENGINEERING CORPORATION, BOSTON, MA (July 1974 to Present)
Appointaents: . t Supervisor, Systems Engineering Group - July 1980 Senior Power Engineer - March 1979 -
Power Engineer - December 1977 Engineer, Power Division - July 1974, Diablo Canyon Nuclear Power Plant, Pacific Gas and Electric Company (Nov 1982 to Present) e As PROJECT ENGINEEF (Nov 1982 to Present), directly responsible for the t.
l' safety-related system design portion of the NRC mandated Independent Design Verification Program (IDVP) for the Diablo Canyon Nuclear Power Plant (DCNPP). The project is unique because it 'is the first and most comprehensive IDVP required by the NRC. Responsibilities include the technical supervision of the mechanical, electrical and instrumentation and control verification of selected safety-related systems. Responsible for the analysis to develop environmental temperatures and pressures due to high energy line break outside the containment. Also responsible for staffing; establishing and meeting schedules, estimating and controlling costs; and maintaining client and NRC liaison.
As I.EAD POWER ENGINEER (. Tune 1982-Nov 1982), directly responsible for the Independent Design Verification of the mechanical and nuclear design of selected safety-related fluid and HVAC systems. Responsibilities included technical and administrative supervision of Power Division Engineers assi3ned to the project.
Systems Engineerina Group, Power Division (July I980-Nov 1982) i As SUPERVISOR of the Systems Engineering Group, directly responsible for development of Reference Fossil Power Plant (RFPP) fluid systems design, including preparation and maintenance of system descriptions and P&ID's; development of Reference Nuclear Power Plant (RNPP) fluid systems design, including preparation and maintenance of system descriptions and P&ID's;
' development of the Industrial Reference Power Plant (IRPP) fluid systems design, including preparation and maintenance of system descriptions and P&ID's; development and maintenance of system-related Power Division Tech-nical Procedures and Guidelines; and resolution of nuclear and fossil plant fluid system related generic engineering and design problem reports and development of preferred solutions.
North Anna Power Station - Unit 2, Virginia Electric and Power Company (Aug 1977-July 1980) i As I.EAD POWER ENGINEER (June 1978-July 1980), directly responsible for-the supervision and administrative control of all Power Division personnel assigned to the 900 MWe project, including nuclear, mechanical, facilities 7SW46-1815 1
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+ JEK and piping Engineers and Designers; technical responsibility for the power plant's nuclear systems, steam plant systems, and HVAC systems, including equipment calculations, and and piping arrangements, conformance to design codes, performance
' drawings; preparation and technical adequacy of nuclear,
-steam plant and HVAC equipment and process specifications; coordination and approval of proj ect work performed by the Power Divi:: ion staff groups; development of engineering man-hour estimates and schedules to ensure timely
-completion of work; and coordination of interface between the Power Division and other engineering disciplines, such as Structural, Electrical, Engi-
.neering Mechanics, and Control Divisions.
As PRINCIPAL NUCLEAR ENGINEER (Aug 1977-June 1978), directly responsible for the technical design of the plant's nuclear and nuclear auxiliary systems, including piping arrangements, conformance to design codes, and preparation of design calculations.
Also responsible for the supervision and coordina-tion of the Engineers in the Nuclear Engineering Group, including scheduling of work and preparatica of nuclear equipment specifications and purchase 1
l orders.
Sundesert Nuclear Power Plant, San Diego Gas & Electric Company (Jan 1976-Aug 1977) l; As ENGINEER on the 900 MWe project, directly responsible for coordination of the layout of the annulus building to ensure compliance with system design criteria, conformance with NRC high energy line criteria, optimization of space utilization, and development of layout requirements. Developed PSAR write-ups for the NSSS systems, including reactor coolant system, chemical and volume control, residual heat removal, and safety injection. Respon-sible for liaison with the NSSS vendor to resolve interface requirements.
I 3,000 MWt. Reference II Design Study, General Atomic' Company (July 1975-Jan 1976)
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As ENGINEER, coordinated the design of the piping and eqiiipment arrangement inside the containment with the goal of reducing HTGR plant costs. The j
various disciplines coordinated.to accomplish this cost reduction included structural, pipe stress, engineered safeguards, and engineering mechanics.
The work included development of containment structures; analysis of high energy line break (both for pipe restraint and containment design pressure determination); application of high temperature pipe stress criteria to piping arrangement; arrangement and location of pipe whip restraints.
Responsible for developing pipe sizes for the major steam (main, hot, and cold reheat) and the feedwater systems within the constraints of minimum costs, pipe stress criteria, allowable pressure drops, and maximum fluid
! velocities.
1200 MWe Nuclear Power Plant, New England Power Company and Central Maine Power Company, Power Plant (July 1974-July 1975)
As ENGINEER, responsible for the development of design criteria and !
implementation of those criteria for the layout and arrangement of the plant's annulus building. Responsible for the development of design bases, l
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.. JEK system description, equipment specifications, and PSAR write-ups for several NSSS and reactor auxiliary systems, including chemical and volume control, residual heat removal, boron recovery, liquid waste, gaseous waste, and
' solid waste. In addition, coordinated the development of the Source Term section of the PSAR and Environmental Report.
PWR SYSTEMS DIVISION AND OFFSHORE POWER SYSTEMS, WESTINGHOUSE EI.ECTRIC CORPORATION (Aug 1971 - Jun 1974)
As SENIOR SYSTEMS ENGINEER, responsible for design of the reactor plant auxiliary systems (e.g., component cooling water, service water, spent fuel pool cooling and purification, containment leak detection, combustible gas control). Responsibilities included development of design criteria, con-formance to design codes, PSAR write-ups, system descriptions, heat balance and fluid flow calculations, and equipment specifications. Supervised the layout and arrangement of assigned systems.
U.S. NAVY - NUCI. EAR SUBMARINE FORCE (June 1965-July 1971)
As CHIEF ENGINEER, responsible for the operation and maintenance of the nuclear submarine's propulsion plant. Directed ship's force and coordinated shipyard work during an extensive submarine overhaul. Supervised 4 officers and 35 enlisted men.
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ATTACHMENT 4 RO6ER F. REEDY, P.E. .
Mr. Reedy has worked in the pressure vessel and . nuclear power industries since 1956. His experience includes the design, analysis, fabrication, and erection of nuclear power plant components and implementation of the applicable quality systems. His background encompasses boiling water, pressurized water, and HTGR nuclear power plants, as well as pressure vessels and storage tanks for petroleum, chemical, and other energy industries. Mr. Reedy is an acknowledged expert in the design of pressure vessels and nuclear components meeting the requirements of the ASME Boiler and Pressure Vessel Code.
He has been involved in licensing, engineering review, project coordination, and . training of personnel. He has testified as an expert witness in litigations and before regulatory groups, ircluding USNRC, ASLB, and ACRS on topics such as design criteria, applications, fabrication techniques, and material applications.
Mr. Reedy has been an active participant for the past 15 years as a member and as chairman of major nuclear Codes and Standards Committees 7
in the development of design, construction and quality criteria for nuclear power plant components. He has served utilities, i architect / engineers, and manufacturers as a consultant on all aspects of nuclear power plant licensing, design, quality considerations, and 4
construction.
Roger F. Reedy is currently chairman of the ASME Section III Code for Nuclear Power Plant Components. He is also a member of the N626.3 Committee which developed the rules concerning duties and responsibilities of engineers designing ASME Code components for nuclear plants. This standard specifies minimum qualifications and 1
details the engineer's responsibilities with regard to coordinating
! material application, fabrication details, quality assurance and non-destructive examinations of the component.
He has worked with the Republic of China Atomic Energy Council to set up an independent quality assurance and inspection program for all nuclear components installed in Taiwan. In addition, for about the past ten years, Mr. Reedy has g(ven lectures on the ASME Code and ~
quality assurance to NRC I & E inspectors in each of the Regions.
Mr. Reedy was one of the initial members of the Pressure Vessel and Piping Division of ASME and helped start the ASME Training Programs for engineers. The program was so successful that other engineering groups have developed similar programs.
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Professional Bacirground American Society of Mechanical Engineers
, . Boiler and Pressure Vessel Comittee Chairman, Subcommittee on Nuclear Power (Section III)
. Executive Comittee, member In 1980, he was awarded the 1980 ASME Centennial Medal by the Policy Board for Coaes and St.andards in recognition of his decades-long contribution to the development of the Boiler and Pressure Vessel Code.
. Subgroup on Containment, past chairman
. Subgroup on Fabrication and Examination, former member ASME Pressure Vessel and Piping Division
. Past Chairman Nuclear Codes and Standards Comittee, member
. ANSI /ASME N626.3 Specialized Professional Engineers Comittee, member Professional Registration Professional Structural Engineer .... Illinois Professional Civil Engineer .... California Illinois Indiana Michigan Wisconsin
3 i $ Professional Experience 1981 - Present R.F. REEDY, INCORPORATED Los Gatos, California President .
Currently consulting with utilities, manufacturers and
- architect / engineers. .
1976 - 1981 NUCLEAR TECHNOLOGY, INCORPORATED l San Jose, California l Successively Manager, Special Projects and C41ef Consultant l
As Manager, Special Projects, he was responsible for coordinating NUTECH'S quality assurance program and their role as Monitor of the Mark I Containment Modification Project.
His CBI experience and ASME Code (Section III) expertise was a key element in working with the utilities and General Electric to define and execute a modification program i acceptable to the U.S. Nuclear Regulatory Commission.
Was then advanced to Chief Consultant, serving as ex-officio advisor to all in-house projects and all clients on design, quality and construction questions concerning application of .
the ASME Code.
f During his term at NUTECH, Mr. Reedy developed and wrote Code Ca)sule, a biennial consnentary on the changes to the ASME Bo'ler and Pressure Vessel Code.
1956 - 1976 -CHICAGO BRIDGE AND IRON COMPANY Oak Brook, Illinois Successively Designer, Staff Engineer, Project Engineer, Design Manager and Senior Engineer.
Duties included design of pressure vessels and storage tanks, including cryogenic vessels, vacuum chambers, multi-layer vessels, environmental chambers, and high-pressure chambers. His duties required close liaison with shop and field personnel, providing Mr. Reedy with an intimate knowledge of practical shop and field construction techniques, including the applicable quality requirements.
! He has designed more than 50 containment vessels and was the responsible Design Manager for most of the nuclear containment vessels fabricated by CBI. He also designed the ~
l first field-erected nuclear reactor.
As Senior Engineer, he consulted with the design staff and i
other departments concerning ASME Code requirements and special projects.
l Education I
B.S., Civil Engineering, Illinois Institute of Technology, 1956 Qualified Lead Auditor, ANSI N 45 J.23
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w AFFIDAVIT OF JOHN M. BIGGS The undersigned, John M. Biggs, this 12th day of October,1983, upon his oath states that the attached Curriculum Vitae is a true and correct statement of his education and professional experience.
,' . hW vv John M. Biggs October 12, 1983 Dal~f5L Notary Public DEBRA A. BURTON, Notary Public My Commissiga Expires December 30,1988 r-
' ATTACHMENT 1 CURRICUEUM VITAE JOHN M. BIGGS Education Massachusetts Institute of Technology ,
Bachelor of Science in Civil Engineering 1941 Master of Science in Civil Engineering 1947 Professional Employment Stress Analyst - Curtiss-Wright Corporation 1941-42 Instructor of Civil Engineering - Robert College, Istanbul, Turkey 1942-45 Structural Designer - Fay, Spofford & Thorndike, Consulting Engineces, Boston 1944-49 Massachusetts Institute of Technology 1947-Date Instructor of Civil Engineering 1947-49 A'_s't. Prof. of Civil Engineering 1949-55 Assoc. Prof. of Civil Engineering 1955-63 Professor of Civil Engineering 1963-Date Emeritus Professor of Civil Engineering 1982-Date Director, Civil Engineering Systems Laboratory 1964-67 Acting Head, Structures Division 1967-68 Head, Structures Group 1976-82 Partner, Hansen, Holley and Biggs, Consulting Engineers, Cambridge, MA 1955-80 Director, Hansen, Holley and Biggs, Inc. Cambridge, MA 1975-Date Professional Societies etc. Registered Professional Engineer, Commonwealth of MA Member, American Society of Civil Engineers Chairman, Structural Division 1970-71 Executive Committee, Structural Division 1968-72 Chairman, Committee on Wind Forces 1955-60 Member, Committee on Electronic Computaticn 1957-60 Member, Committee on Plasticity Related to Design 1957-59 Member, Committee on the Limitations of Bridge Deflection 1956-61 Member, Administrative Committee on Loads and Stresses 1955-60 Member, Committee on Lifeline Earthquake Engineering 1973-74 Member, Boston Society of Civil Engineers Chairman, Structural Section 1957-58 Director 1959-61 Vice-President 1964-67 President 1966-67 1
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< Member, Column Research Council 1958-72 l Member, Mayor's Committee on the Revision of the i Boston Building Code l Subcomittee on Steel Design 1956-63 l Subcomittee on Loads 1958-60 Member Advisory Committee, Massachusetts State Building Code 1974-Date -
Member, Advisory Panel on Bridges, AASH0 Road Test Member, Comittee on Bridges, Highway Research Board, i - National Academy of Sciences 1961-65 Member, Comittee on Design Loads for Buildings, American Standards Association 1962-66 Recent Publications
" Introduction to Structural Dynamics," McGrap Hill Book Co., NY, 1964
" Structural Response to Seismic Input," page 306, Seismic Desian for Nuclear Power Pla.nts, MIT Pren , Cambridge, MA, 1970
" Seismic Analysis of Equipmerit Mounted on a Massive Structure," page 319, Seismic Desitn for Nuclear Power Plants, MIT Press, Cambridge, MA, 1970 (with J. Loessst).
i " Computer System for the Analysis and Design of Reinforced Concrete Structures," ACI Journal, April, 1970 (with P.J. Pahl and H.N. Wenke).
" Soil-Structure Interaction in Nuclear Power Plants," 3rd Japanese Symposium on Earthauake Engineerina, Tokyo, November,1975' Twith R.sJ Whitman).
" Integrated System for RC Building Design," Journal of the Structural Division. ASCE. Vol. 07, January, 1971 (with P.J. Pahl and H.N.
Wenke).
" Earthquake Code Evolution and the Effect of Seismic Design on the
! Cost of Buildings " NIT Department of Civil Engineering, Report No. R72-20, May, 1972 (with S.J. Leslie). J
" Seismic Response Spectra for Equipment in Nuclear Power Plants,"
Proceedings. First International Conference on Structural Mechanics in Reactor Technology, Berlin, Ju'ly, 1972.
" Parametric Analysis of Soil-Structure Interaction for a Reactor .~
Building," Proceedinas, First International Conference on Structural Mechanics in Reactor Technoloqy, Berlin, July, 1972 (with J.T. Christian and R.V. Whitman).
" Seismic Response of Buildings Designed by Code for Different Earthquake Intensities," MIT Department of Civil Engineering, Report No. R73-7, January, 1973 (with P.H. Grace).
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Y " Seismic Design Decision Analysis," Journal of the Structural Division. ASCE, Vol. 101, STS, May, 1975 (with R.V. Whitman, et al).
" Comparison of Seismic Analysis Procedures for Elastic Multi-Degree Systems," MIT Department of Civil Engineering, Report No. R76-5, January, 1976 (with E.H. Vanmarcke et al.). " Variability of Inelastic Structural Response Due to Real and Artificial Ground Motions," MIT Department of Civil Engineering, -
Report R76-6, January, 1976 (with E.H. Vanmarcke, Robert A. Frank, et al.).
" Studies on the Inelastic Dynamic Analysis and Design of Multi-Story Frames," MIT Department of Civil Engineering, Report R76-29, July, 1976 (with W.H. Luyties and S.A. Anagnostopoulos). " Inelastic Response Spectrum Design Procedures for Steel Frames," MIT Department of Civil Engineering, Report No. R76-40, September, 1976 (with Richard W. Haviland). "On the Safety Provided by Alternate Seismic Design Methods," MIT Civil Engineering Department, Report No. R77-22, July,1977 (vith D.A.
Gasparini).
" Inelastic Dynamic Design of Steel Frames to Resist Seismic Loads,"
MIT Civil Engineering Department, Report No. R77-23, July. 1977 (with J.H. Robinson,Jr.). .
"Use of Inelastic Spectra in Aseismic Design," Journal of the Structural Division. ASCE. Vol. 104, No. ST1, January 1978 (with S.A.
Anagnostopoulos and R.W. Haviland).
" Inelastic Response Spectra for Aseismic Building Design," Journal of
{ the Structural Division. ASCE, Vol. 106, No. ST6, June, 1980 (with ! S.P. Lai).
" Seismic Effectiveness of . Tuned Mass Dampers," Journal of the l
Structural Division, ASCE Vol.107, No. ST8, August,1981 (with A.M. Kaynia and D. Veneziano).
" Seismic Damage in Reinforced Concrete Frames," Journal of the l
Structuraf Division. ASCI, Vol. 107, No. ST9, September, 1981 (with H. I Banon and H M. Irvine). l l " Flexible Sleeved-Pile Foundations for Ascismic Design," MIT Civil l Engineering Department, Report No. R82-04, March, 1952. 3 i I
AFFIDAVIT OF MYLE J. HOLLEY, JR. The undersigned, Myle J. Holley, Jr., this 12th day of October, 1983, upon his oath states that the attached statement of Professional Experience is a true and correct statement of his education and professional experience, h_ , _
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Myle J. Ho) y, Jr. October 12, 1983 l l I di Notary Public WILLIAM S. MOONAN NOTARY PUBLIC MY COMMISSION EXPIRES - AUGUST 6,1987 - 1
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, s. ATTACHMENT 2 STA1EENT OF PROFESSIONAL EXPERIENCE MYLE J. HOLLEY, JR. Mr. Holley received the S.B. and S.M degrees in Civil Engineering from MIT in 1939 and 1947, respectively. From 1939 to 1946 he was employed by the S. Morgan - Smith Co. (now the York, PA Division of Allis-Chalmers Manufacturing Co.) as a stress analyst and designer of heavy machinery. In 1946 he joined the Faculty of MIT in the Department of Civil Engineering. While on that faculty he taught subjects in structural analysis and design, and supervised structural research projects. The latter included work in the fields of massive reinforced concrete structures, prestressed concrete, structural applications of granite, high-strength reinforced concrete beams, and the performance of thin arch concrete dams. For several of his years on the MIT Faculty, Mr. Holley was in charge of the Structural Division of the Civil Engineering Department. In 1955 Professor Hollcy and his colleagues, Professors John M. Biggs and Robert J. Hansen, formed the consulting partnership Hansen, Holley and Biggs. Since 1975 the group has functioned as Hansen, Holley and Biggs, Inc. Mr. Holley's participation in the professional efforts of the group has continued undiminished since his retirement from teaching in 1974. The professional assignments of Hansen, Holley and Biggs have been related, almost exclusively, to complex problems of structural design and structural behavior. Their clients include both engineering firms and owners of major constructed facilities. A substantial fraction of their practice has involved advice and assistance in the resolution of problems arising in the design and construction of nuclear power plants. In this area of their practice, clients have included: Stone and Webster Engineering Corporation United Engineers and Constructors Gibbs and Hill American Electric Power Corporation Rochester Gas and Electric Company Portland Gas and Electric Company Mr. Holley has been extensively involved in structural aspects of nuclear power plant projects for all of the above companies. In addition, he has been a consulting member of several internal design review boards conducted by Stone and Webster Engineering Corporation. Mr. Holley is a registered Professional Engineer in the Comonwealth of Massachusetts. He is a member of the American Society of Civil Engineers, the - American Concrete Institute, and the American Society for Engineering Education. He has served on numerous prof essional committees. This has included several years on ACI 349 Concrete Nuclear Structures and ACI 359 Nuclear Reactor Components, and he currently is a consulting member of these (ommittees.
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AFFIDAVIT OF RONALD WRAY The undersigned, Ronald Wray, this 12th day of October, 1983, upor his oath states that the attached Professional Resume is a true and correct statement of his education and professional experience. fW Ronald Wray October 12, 1983 dd/$w lh ~--- Notary Public WILLIAM S. MOONAN NOTARY PUCLIC MY COM.".11SS!ON EXPIRES AUGUST. Gt .1987. 9- - w a n e - w, , --m - r,,-- +- e- e -n,--,-----m e-- - - - - . - - - - -
ATTACHMENT 3
'RTELEDYNE ENGINEERING SERVICES RONALD WRAY Manager, Engineering Analysis Professional Resume '
Education Northeastern University, B.S. in Civil Engineering, 1956 Rensselaer Polytechnic Institute, M.S. in Engineering Science, 1962 Experience Teledyne Engineering Services, and Teledyne Materials Research, since 1971: theoretical stress analysis of pressure vessels, piping systems and frame structures utilizing computer program solutions and finite element methods; performed and directed static and dynamic analyses of Nuclear and LNG Piping Systems; conducted design reviews of Nuclear Piping Systems. Instructor at Franklin Institute of Boston, Evening Division A'!C0 Systems Division, 1962-1971: performed detailed stress and buckling analysis of various reentry vehicle shell structures under combined reentry pressure and inertia loads and heating. Designed and analyzed large vacuun and pressurized chambers for a portable sterilization / clean room facility built for NASA / Langley; responsible for the structural design and evaluation of space power systems and planetary probe systems. l Pratt & Whitney Aircraft, Canal Division, 1958-1962: performed and directed detailed analyses and design evaluation of nuclear reactor core components and pressure vessels; conducted thermo-structural analysis of system piping and heat exchangers involving liquid metal coolants under conditions of high temperature operation an severe transients; established design criteria for components exposed to long-life, high-temperature conditions, U.S. Army Corps of Engineers, 1st Lieutenant, 1956-1958: seried as project officer on military construction sites; field experience in reinforced concrete an steel erection. Membership ~ ASME, Boiler and Pressure Vessel Code, Chi.irman, Special Working Group on Dyna.nic Analysis. 1/'d
o LownwTatx, NzwwAw, Rana O Axzt.cAn, P. C. 10/14/83 s IDVP Exhibit List (1) Independent Design Verification Program Final Report, Diablo - Canyon Nuclear Power Plant - Unit 1 (as revised through 10/10/83).
Purpose:
Summary of the IDVP efforts and statement of conclusions and evaluations of the IDVP. Sponscring witness: Dr. William E. Cooper (2) (a) Diablo Canyon Nuclear Power Plant Design Verification Program Management Plan, Phase I (March 29, 1982) (b) Diablo Canyon Nuclear Power Plant Design Verification Program Management Plan, Phase II (June 18, 1982)
Purpose:
Description of program plans of the IDVP. Sponsoring witness: Dr. William E. Cooper (3) Interim Technical Reports (ITR's) of IDVP, as listed in Attachment A hereto.
Purpose:
Depending upon the ITR, documentation of programmatic aspects of the IDVP or report of detailed technical results. Sponsoring witness: Dr. William E. Cooper for ITR's issued by TES Dr. Robert L. Cloud for ITR's issued by RLCA Mr. John E. Krechting for ITR's issued by SWEC Mr. Roger F. Reedy for ITR's issued by RFR
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. ATTACHMENT A TO EXHIBIT LIST I.
REV ISSUE ISSUED ITR NO. DATE BY TITLE 1 1 821022 RLCA Additional Verification and Additional Sampling (Phase 1) 2 0 820623 TES Comments on the R.F. Reedy, Inc. , Qual-ity Assurance Audit Report on Safety-Related Activities Performed by PGandE Prior to June 1978 3 0 820716 RLCA Tanks 4 0 820723 RLCA Shake Table Testing 5 0 820819 RLCA Design Chain 6 0 820910 RLCA Auxiliary Building 7 0 820917 RLCA Electrical R6ceway Supports 8 0 821005 RLCA Independent Design Verification Program for Verification of PGandE Corrective Action 9 0 821015 RFR Development of the Service-Related Con-tractor List for Non-Seismic Design Work Performed for DCNPP-1 Prior to June 1, 1978 10 0 821029 RLCA Verification of Design Analysis Scsgri Spectra 11 0 821102 TES PGandE-Westinghouse Seismic Interface Review 12 0 821105 RLCA Piping 13 0 821105 RLCA Soils - Intake Structure 14 2 830725 SWEC Verification of the Pressure, Tempera-ture, Humidity, and Submergence Envi-ronments used for Safety-Related Equip-
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ment Specifications Outside Containment for Auxiliary Feedwater System and CRVP System A-1
ATTACHMENT A TO EXHIBIT LIST REV ISSUE ISSUED ITR NO. DATE BY _ TITLE , 15 0 821210 RLCA HVAC Duct and Supports Report 16 0 821208 RLCA Soils - Outdoor Water Storage Tanks 17 0 821214 RLCA Piping - Additional Samples 18 1 830524 SWEC Verification of the Fire Protection Provided for Auxiliary Feedwater System Control Room Ventilation and Pressuri-zation System Safety-Related Portion of the 4160V Electric System 19 0 821216 SWEC Verification of the Post-LOCA Portion of the Radiation Environments used for Safety-Related Equipment Specification Outside Containment for Auxiliary Feed-water System and Control Rocm Ventila-tion and Pressurization System 20 2 830725 SWEC Verification of the Mechanical /Naclear Design of the Control Room Ventilation and Pressurization System 21 1 830503 SWEC Verification of the Effects of Hign Energy Line Cracks and Moderate Energy Line Breaks for Auxiliary Feedwater System and Control Room Ventilation and Pressurization System 22 2 830725 SWEC verification of the Mechanical / Nuclear Portion of the Auxiliary Feedwater System
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23 1 830527 SWEC Verification of High Energy Line Break and Internally Generated Missile Review Outside Containment for Auxiliary Feed-water System and Control Room Ventila-tion and Pressurization System A-2'
, ATTACHMENT A TO EXHIBIT LIST t REV ISSUE ISSUED ITR NO. DATE BY TITLE 24 1 830504 SWEC Verification of the 4160V Safety - Related Electrical Distribution System 25 1 830429 SWEC Verification of the Auxiliary Feedwater System Electrical Design 26 1 830502 SWEC Verification of the Control Room Venti-lation and Pressurization System Elec-trical Design 27 2 830725 SWEC Verification of the Instrument and Con-trol Design of the Auxiliary Feedwater System 28 2 830725 SWEC Verification of the Instrument and Con-trol Design of the Control Room Ventil-ation and Pressurization System 29 0 820117 SWEC Design Chain - Initial Samples 30 0 830112 RLCA Small Bore Piping Report 31 1 830804 RLCA HVAC Components 32 1 830401 RLCA Pumps 33 1 830428 RLCA Electrical Equipment Analysis 34 1 83' '24 SWEC Independent Design Verification of DCP Efforts by SWEC 35 0 830401 RLCA Independent Design Verification Program Verification Plan for DCP Activities 36 1 830620 SWEC Final Report on Construction Quality Assurance Evaluation of G.F. Atkinson 37 0 830223 RLCA Valves 38 2 830620 SWEC Final Report on Construction Quality
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Assurance Evaluation of Wismer and Becker 39 0 830225 RLCA Soils - Intake Structure Bearing Capacity and Lateral Earth Pressure A-3
. ATTACHMENT A TO EXHIBIT LIST REV ISSUE ISSUED ITR NO. DATE BY TITLE 40 0 830309 RLCA Soils Report - Intake Sliding Resistance - 41 0 830419 RFR Corrective Action Program and Design Office Verification 42 0 830415 RFR R.F. Reedy, Inc., Independent Design Verification Program Phase II Review and Audit of PGandE and Design Consul-tants for DCNPP-1 43 0 830414 RLCA Heat Exchangers 44 0 830415 RLCA Shake Table Test Mounting Class 1E Electrical Equipment 45 0 830517 SWEC Additional Verification of Redundancy of Equipment and Power Supplies in Shared Safety-Related Systems 46 0 830627 SilEC Additional Verification of Selection of System Design Pressure and Temperature and Differential Pressure Across Power-Operated Valves 47 0 830627 SWEC Additional Verification of Environ-mental Consequences of Postulated Pipe Ruptures Outside of Containment 48 0 830727 SWEC Additional Verification of Jet Impinge-ment Effects of Postulated Pipe Ruptures Inside Containment 49 0 830623 SWEC Additional Verification of Circuit Sep-aration and Single Failure Review of Safety-Related Electrical Equipment
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50 0 830722 TES Containment Annulus Structure Vertical Seismic Evaluation 51 1 830915 TES Containment Annulus Structure - Verification of DCP Corrective Action A-4
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ATTACHMENT A TO EXHIBIT LIST REV ISSUE ISSUED ITR NO. DATE BY TITLE 54 1 831003 RLCA Corrective Action Containment Building 55 1 831001 RLCA Corrective Action Auxiliary Building 56 1 830924 RLCA Corrective Action Turbine Building 57 1 830908 RLCA Review of DCP Activities Fuel Handling Building 58 1 831001 RLCA Verification of DCP Activities Intake Structure 59 1 830924 RLCA Corrective Action large Bore Piping 60 1 831003 RLCA Corrective Action large and Small Bore Pipe Supports 61 1 831002 RLCA Corrective Action Small Bore Piping 63 -1 831002 RLCA Corrective Action HVAC Ducts, Raceways, Instrument Tubing and Associated Sup-ports 65 1 831010 RLCA Corrective Action Rupture Restraints ,, 67 1 830909 RLCA Corrective Action, Equipment 68 1 831004 RLCA Verification of HLA Scils Work A-5
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9 fg UNITED STATES OF AMERICA NUCLEAR-REGULATORY COMMISSION BEFORE THE ATOMIC S,AFETY AND LICENSING APPEAL BOARD
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In the Matter of ) PACIFIC GAS AND ELECTRIC ) Docket Nos. 50-275 0.L. COMPANY ) 50-323 0.L.
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~(Diablo Canyon Nuclear Power )
Plant, Units 1 and 2) )
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CERTIFICATE OF SERVICE I hereby certify that copies of the letter from Maurice Axelrad to the Appeal Board dated October 14, 1983, and its enclosures (pre-filed direct testimony of three panels of IDVP witnesses, qualifica-tions of IDVP witnesses, and IDVP exhibit list) have been served on the following by' deposit in the United States mail, first class, postage pre-paid, this 14th day of October, 1983, except that, in the case of individuals designated by an asterisk, arrangements have been made for delivery by courier or personal delivery no later than 10:00 a.m. on October 17th: f
*Dr. John H. Buck John F. Wolf, Esq.
Atomic Safety and Licensing Administrative Judge i Appeal Board Atomic Safety and Licensing - U.S. Nuclear Regulatory Board i Commission U.S. Nuclear Regulatory Washington, D. C. 20555 Commission Washington, D. C. 20555
*Dr. W. Reed Johnson
. Atomic Safety and Licensing Mr. Glenn O. Bright Appeal Board Administrative Judge 4
U.S. Nuclear Regulatory Atomic Safety and Licensing Commission Board Washington, D. C. 20555 U.S. Nuclear Regulatory Commission
- Thomas S. Moore, Esq., Chairman Washington, D. C. 20555 Atomic Safety and Licensing Appeal Board
- Lawrence J. Chandler U.S. Nuclear hagulatory Office of Executive Legal Commission Director Washington,.D. C. 20555 BETH 042
, U.S. Nuclear Regulatory Dr. Jerry Kline Commission Administrative Judge Washington, D. C. 20555
- Atomic Safety and Licensing Board
- Philip A. Crane, Jr., Esq.
U.S.' Nuclear Regulatory Pacific Gas and Electric J Commission Company Washington, D. C. 20555 P.O. Box 7442 San Francisco, CA 94120 i
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j . o Elizabeth Apfelberg Richard E. Blankenburg, 1415 Cozadero Co-publisher San Luis Obispo, CA 93401 Wayne A. Soroyan, News Reporter South County Publishing Mr. Gordon Silver Company Mrs.-Sandra A. Silver P.O. Box 460 1760 Alisal Street Arroyo Grande, CA 93420 San Luis Obispo, CA 93401 Harry _M. Willis
*Joel R. Reynolds, Esq. Seymour & Willis John R. Phillips, Esq. 601 California Street Center for Law in the Public Suite 2100 Interest San Francisco, CA 94108 10951 West Pico Boulevard Third Floor 'Janice E. Kerr, Esq.
Los Angeles, CA 90064 Lawrence Q. Garcia, Esq. 350 McAllister Street Arthur C. Gehr, Esq. San Francisco, CA 94102 Snell & Wilmer 3100 Valley Center
- Mr . Jame s O . Schuyler Phoenix, AZ 85073 Nuclear Projects Engineer Pacific Gas and Electric Mark Gottlieb Company California Energy Commission 77 Beale Street MS-18 San Francisco, CA 94106 1111 Howe Avenue Sacramento, CA 95825 Paul C. Valentine, Esq.
321 Lytton Avenue
- Bruce Norton, Esq. Palo Alta, CA 94302
.Norto.', Burke, Berry Frent, P.C. David S. Fleischaker, Esq.
2002 Eass Osborn Street P.O. Box 1178 Phoenix, AZ 85064 Oklahoma City, OK 73101
- Michael J. Strumwasser, Esq. Richard B. Hubbard Susan L. Durbin, Esq. MHB Technical Associates Peter H. Kaufman, Esq. 1723 Hamilton Avenue 3580 Wilshire Boulevard Suite K Suite 600 San Jose, CA 95125 Los Angeles, CA 90010 John Marrs, Managing Editor Mr. Frederick Eissler San Luis Obispo County Scenic Shore Preservation Telegram-Tribune Conference, Inc. 1321 Johnson Avenue 4623 More-Mesa Drive P.O. Box 112 Santa Barbara, CA 93105 San Luis Obispo, CA 93406 Mrs. Raye Fleming
- Docketing and Service Section 1920 Mattie Road U.S. Nuclear Regulatory Shell Beach, CA 93449 Commission Washington, D. C. 20555
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IB Mr. Thomas H. Harris Atomic Safety and Licensing Energy Writer Board San Jose Mercury News U.S.-Nuclear Regulatory 750 Ridder Park Drive Commission SaniJosa, CA 95190 Washington, D. C. 20555 -
- Atomic Safety and Licensing Appeal Board
-U.S. Nuclear Regulatory Commission Washington, D. C. 20555 i'
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