ML20213C766
| ML20213C766 | |
| Person / Time | |
|---|---|
| Site: | Black Fox |
| Issue date: | 09/25/1978 |
| From: | Gallo J, Murphy P ISHAM, LINCOLN & BEALE |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20213C771 | List: |
| References | |
| NUDOCS 7810040055 | |
| Download: ML20213C766 (189) | |
Text
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UNITED STATES OF AMERICA gp o cg%y 1S NUCLEAR REGULATORY COMMISSION -
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% O ~3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD b ' ,y e j In the Matter of the Application of ) os Public Service Company of Oklahoma, )
Associated Electric Cooperative, Inc. ) Docket Nos. STN 50-556 and ) STN 50-557 Western Farmers Electric Cooperative )
)
(Black Fox Units 1 and 2) )
NOTICE OF FILING APPLICANTS' DIRECT TESTIMONY AND IDENTIFICATION OF EXHIBITS Pursuant to the schedule, as amended, set by the Board following the 10 CFR S2.752 prehearing conference, Public Service Company of Oklahoma, Associated Electric Cooperative, Inc. and Western Farmers Electric Cooperative
(" Applicants") hereby file their direct testimony (including witness identification and qualification) and a list of exhibits which Applicants intend to offer in evidence at the construction permit radiological health and safety hearing for the Black Fox Station.
TESTIMONY Applicants' Direct Testimony on contentions remain-ing in controversy following the Board's " Order Ruling On Motions For Summary Disposition," dated September 8, 1978, and responding to questions raised therein consist of the following:
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UNITED STATFS 0 NUCLEAR REGD6ATOR)QA,MERICMY
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of the Applicati'on of )
Public Service Company of Oklahoma, ) ;
Associated Electric Cooperative, Inc. ) Docket Nos. STN 0-55 j and ) STN 50- o Western Farmers Electric Cooperative ) ,
)
(Black Fox Units 1 and 2) ) -
s, NOTICE OF FILING APPLICANTd* DIRECT TESTIMONY AND IDENTIFICATION OF EXHIBITS l l
Pursuant to the schedule, as amended, set by the Board following the 10 CFR S2.752 prehearing conference, Public Service Company of Oklahoma, Associated Electric Cooperative, Inc. and Western Farmers Electric Coopera.tive
(". Applicants"). hereby file their direct tastimony (including witness identification and qualification) and a list of exhibits which Applicants intend to offer in evidence at the construction permit radiological health and safety hearing for the Black Fox Station.
m TESTIMONY Applicants' Direct Testimony on contentions remain-ing in controversy following the Board's " Order Ruling On
+
Motions For Summary Disposition," dated September 8, 1978, and responding to questions raised therein consist of the following:
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A. Contention 1 Board Question 1 This question is directed to the NRC Staff. No response by Applicants is made at this time.
B. Contention 2 Board Question 2 This question is addressed in
" Testimony of Aaron J. Levine Concerning Ques-tion 2-1."
Board Question 2 This question is addressed in
" Testimony of Aaron J. Levine Concerning Ques-tion 2-2."
C. Contentions 3, 5 and 16, which have been consolidated for purposes of the hearing The following testimony addresses these conten-tions and Board Questions relevant thereto. )
- 1) " Testimony of Lambert J. Sobon Concerning Contention 16."
- 2) " Testimony of David F. Guyot Concerning Contentions 3 and 16 and Questions 5-1 and 12-3."
- 3) " Testimony of Lowell E. Thurman Concerning Contention 3."
- 4) " Testimony of William Gang Concerning l Contention 3."
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- 5) " Testimony of William Gang Concerning Question 5-1."
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The Board has also requested that the parties address the appropriate Method of combining loads as an additional matter under Contention 16. As the Board noted in its Order, load combination methodology has not yet been settled between Applicants and the NRC Staff. The Staff, by letter from Colleen Woodhead, Esquire to the members of the Board dated June 27, 1978, stated that "[t]he matter of load combination methods will be addressed in the final supplement to the SER."
Supplement No. 1 to the SER, dated September 1978, at Section 1.9, p. 1-2, identifies load combination methodology as an open item.
As stated in the Supplement at page 3-1, the NRC Staff is conducting a generic review of load combination methods. It also was stated that the review was expected to be completed by mid September 1978. In the meantime the l.
'NRC Staff requested the Applicants to commit -- sight unseen --
to the generic resolution. After serious consideration, the Applicants, on August 18, 1978, declined to make the commit-l ment because of its uncertain nature. We now understand that the NRC Staff has formulated its generic position on load combination methods and we are awaiting official notification of that position on the Black Fox docket. As soon as the generic position is received, Applicants will quickly analyze the information to determine its acceptability. Until that
I time, Applicants' position on load combination methodology remains as set forth in the PSAR. However, because of the l vagaries of the situation brought about by the apparent comple- J tion of the NRC Staff's generic review, we believe the prudent course would be to await official notification by the NRC Staff i !
f and review that material before pressing the issue further. We therefore reserve the right to file additional or rebuttal testimony on the question of load combination methodology once the NRC Staff's position is known.
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D. Contention 6
- 1) Board Question 6 This question is 1
directed to the NRC Staff. No response by Applicants is made at j l
this time.
- 2) Board Question 6 This question is 1
addressed in " Testimony of Robert E.
Stippich Concerning Question 6-2."
E. Contentions 7 and 8 Applicants have consolidated these contentions in their testimony. The following testimony addresses these contentions.
- 1) " Testimony of Gary R. Engmann Concerning l
l Contentions 7 and 8 (Fire Protection) . "
- 2) " Testimony of William G. Gang and Richard B.
Johnson Concerning Contentions 7 and 8 (Fire Protection)."
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i F. Contention 9
" Testimony of E. L. Cox Concerning Contention 9 (Fire Protection)."
G. Contention 10 l 1) Board Question 10-1 " Testimony of William G. Gang and Gerald M. Gordon Concerning Question 10-1" H 2) Board Questions 10-2 and 10-3 " Testimony of J. B. Perez Concerning Questions 10-2 and 10-3"
- 3) Board Question 10-4 " Testimony of Donald G.
Long Concerning Question 10-4" H. Contention 12
- 1) Board Questions 12-1, 12-4 and 12 " Testimony of C. J. Ross Concerning Questions 12-1, 12-4 and 12-5."
- 2) Board Question 12 This question is directed to the NRC Staff. No response by Applicants is made at this time.
- 3) Board Question 12 This question is answered as a part of David F. Guyot's testimony noted supra on Contentions 3, 5 and 16.
I. Contention 13 Board Question 13 This question is directed to the NRC Staff. No response is made by Appli-cants at this time.
l l J. Contention 15 !
Board Question 15-1 " Testimony of Gerald M. Gordon L Concerning Question 15-1."
K. Contention 18 Board Question 18-1 " Testimony of Dwane R. Glancy Concerning Question 18-1."
L. Contention 19 )
Board Questions, 19-1, 19-2 and 19 These ques-tions are directed to the NRC Staff. No response is made by Applicants at this time.
l M. Contention 66 i
Board Question 66 This question is directed to the NRC Staff. No response is made by Applicants at this time.
N. Contention 67 The.following testimony addresses this Contention:
- 1) " Testimony of Edward D. Fuller Concerning Contention 67 (Anticipated Transient With-out Scram)."
- 2) " Testimony of John C. Zink Concerning l Contention 67 (Anticipated Transient With- l out Scram)."
O. Additional Contention 1 " Testimony of Aaron J.
i Levine Concerning Additional Contention 1 (Preven-l tion of Off-Gas Explosions)."
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I EXHIBITS l Applicants will offer the following exhibits to be l i received in evidence. The witnesses who will sponsor the exhibits are also identified. Some of these exhibits have previously been furnished to the Board and counsel for each i of the parties. Where appropriate, copies of the exhibits l
are being furnished concurrently with this filing, but under
! separate cover, to each member of the Board and counsel for each of the parties.
A. Applicants' Safety Exhibit No. 1 to be sponsored
! by Dwane Glancy of Public Service Company of Oklahoma - Amendments 2 and 3 to the " Application For Licenses, Construction Permit State." (The Application with Amendment 1 thereto was introduced into evidence as Applicants' Exhibit No. 1 during the environmental hearings.)
B. Applicants' Safety Exhibit No. 2 to be sponsored by Vaughn L. Conrad of Public Service Company of Oklahot.2, E. L. Cox of Black & Veatch and William G.
Gang of General Electric Company - Amendments 9, 10 and 11 to the " Preliminary Safety Analysis Report." (The PSAR with Amendments 0 through 8 thereto was introduced into evidence as Applicants' Exhibit No. 2 during the environmental hearings.)
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C. Applicants' Safety Exhibit No. 3 to be sponsored by Lambert J. Sobon of General Electric Company -
l Applicants will offer the following as a group l
l exhibit:
3A) "Information Report, Mark III Containment Dynamic Loading Conditions," which is Appendix l 3B to the 238 Nuclear Island General Electric Company Standard Safety Analysis Report (GESSAR) ,
i Docket STN 50-447.
3B) Notice of Filing by GE of Amendment No. 37 To l Application For Review of 238 GESSAR, dated August 29, 1975 l 3C) Letter dated October 24, 1975 NRC Staff (Moore) to GE (Stuart) 3D) Letter dated November 7, 1975 GE (Stuart) to i
NRC Staff (Moore) 3E) Letter dated December 22, 1975 NRC Staff (Boyd) to GE (Stuart) 3F) Latter dated February 17, 1977 GE (Sherwood) to NRC Staff (Rusche) 3G) Letter dated March 25, 1977 NRC Staff (Varga) to GS (Sherwood) 3H) Letter dated April 8, 1977 GE (Gilbert) to NRC Staff (Varga) 3I) Letter dated June 13, 1977 NRC Staff (Boyd) to GE (Sherwood) l l
i D. Applicants' Safety Exhibit No. 4 to be spo7sored by Richard E. Johnson of General Electric Company -
" Power Generation Control Complex Design Criteria and Safety Evaluation," NEDO-10466, Rev. 2.
E. Applicants' Safety Exhibit No. 5 to be sponsored by E. L. Cox of Black & Veatch " Black Fox Station Fire Hazards Analysis, Construction Permit Stage, Reference Report 16."
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i In addition to the above, Applicants will offer evidence of Rural Electrification Administration load guarantee approval for !
that portion of the Black Fox Station to be financed by Asso-l ciated Electric Cooperative, Inc. and Western Farmers Electric j i
Cooperative. Evidence of final REA approval is not yet available to Applicants. It is anticipated that the authenticity of such ]
4 approval will not be challenged and that the documentation of the approval can be admitted by stipulation of the parties.
DATED: September 25, 1978.
Respectfully submitted, l l
ISHAM, LINCOLN & BEALE ( / / l A fa<Lef #
1050 17th Street, N.W.
Suite 701 Josepft Gallo Washington, D.C. 20036 (202) 833-9730 ,
Y lb Ybr4MD i ISHAM, LINCOLN & BEALE Paul M. Murphy ' l g 4' One First National Plaza Suite 4200 Chicago, Illinois 60603 (312) 786-7500
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! ,; y Index
!- ./
- 1. Additional Contention 1: Aaron Levine
- 2. -Question 2-1: Aaron Levine ,
J
- 3. Question 2-2: Aaron Levine
- 4. Contentions 3 & 16 and' Questions 5-1 & 12-3:
!' A. William G. Gang (3)
B. Lowell E. Thurman (3)
C. William G.~ Gang (5-1)
D.- David Guyot (3 & 16 and 5-1 & 12-3)
E. Lambert J. Sobon (16) j l
'5. Question 6-2: Robert E. Stippich
- 6. Contentions 7 & 8 A. William G. Gang &' Richard B. Johnson
. B. Gary R. Engmann
- 7. Contention 9: E. L. Cox
- 8. Question 10-1: William G. Gang & Gerald M. Gordon
- 9. Question 10-2 & 10-3: J. B. Perez
- 10. Question 10-4: Donald G. Long
'll . Questions 12-1, 12-4 & 12-5: C. J. Ross
- 12. Question 15-1: Gerald M. Gordon
- 13. Question 18-1: Dwane R. Glancy
- 14. Contention 67 A. Edward D. Fuller B. John C. Zink i
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I g
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g [3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD b 9 ;
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In the Matter of )
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l PUBLIC SERVICE COMPANY OF OKLAHOMA, )
l ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 l AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPE RATIVE , INC. )
)
l (Black Fox Station, Units 1 and 2) )
l Testimony of Mr. Aaron J. Levine Concerning Additional Contention 1.
l (Off-Gas System Explosions) l l September 25, 1978 l
j TESTIMONY OF MR. AARON J. LEVINE ,
CONCERNING ADDITIONAL CONTENTION 1 L (OFF-GAS SYSTEM EXPLOSIONS) r My name is Aaron J. Levine. My business address is 175 i
Curtner Avenue, San Jose, California. I am the Manager of
' Projects Licensing Unit 1 for the Safety & Licensing Operation of the General Electric Company, Nuclear Energy Business Group.
My statement of qualifications is attached to my testimony on Question 2-1.
The purpose of my testimony is to demonstrate that the Off-Gas ~ System for the Black Fox Station is designed to avoid off-gas explosions such as those which have occurred at the Millstone Station and other earlier vintage BWR plants. Addition-ally, my testimony shows that even in the event of gross failure of the Off-Gas System pressure boundary, the resulting off-site dose would be significantly less than the 10 CFR Part 100 limits.
As described in PSAR Section 11.3, the Black Fox Station Off-Gas System will be a General Electric supplied N-66 Gaseous Effluent Additional Contention 1 reads:
Intervenors contend that Applicant and Staff have not adequately analyzed the cause and means of prevention of explosions resulting from hydrogen escaping from the off-gas system. Such explosions are apparently limited to BWR reactors and have associated secondary explosions, e.g.,
ignition of hydrogen in the base of the effluent release stack.
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2-Treatment System. The system is designed to process the gaseous radioactive effluents from the plant b,y a combination of filtration and delay, and thus control off-site release.
These effluents consist of non-condensible gases which are drawn from the condenser by the steam jet air ejector (SJAE) which keeps the condenser at vacuum. Among the gases removed from the condenser are hydrogen and oxygen formed by the radiolytic l
decomposition of water. If present in appropriate quantities, the hydrogen and oxygen can form a combustible mixture.
The Black Fox Off-Gas System is designed to minimize the 1
potential for and the effects of a hydrogen explosion both !
internally and externally to the system. To prevent internal hydrogen explosions the hydrogen concentration is controlled l below 4 per cent by volume which is below the lower limit of flammability (LLF). This is done using dilution steam and hydrogen recombiners. Hydrogen analyzers are provided in the system to measure the hydrogen concentration in the process stream. None of these were employed on the Millstone Off-Gas System at the time of the explosions.
Steam is mixed with the process stream as it leaves the air ejectors. The addition of sufficient steam raises the LLF of the gas mixture by a factor of about 2. The Black Fox system provides more steam than is necessary to result in a mixture which is significantly below the flammability limit.
The process stream then flows to one of two redundant hydrogen recombiners. The hydrogen recombiner recombines hydrogen
and oxygen to form water. The system is designed to maintain the hydrogen concentration below 4 per cent (including steam) at the inlet to the recombiner and below 1 per cent at the recombiner outlet on a volume basis. Hydrogen analyzers in the recombiner outlet line monitor the hydrogen concentration and activate a .
t control room alarm when hydrogen concentration exceeds 1 per cent. l I
The recombiners, each of which is designed to handle full system l flow, process the gas stream before it enters the rest of the Off-Gas System components, thus precluding the possibility of a combustible mixture within the remainder of the Off-Gas System.
The recombiner in the Black Fox system is one with which GE has had more than forty equipment years of domestic experience and about twenty more equipment years of international experience.
During that time there have been no reported failures of this recombiner to operate satisfactorily.
The Off-Gas System is designed to accommodate the energy release from a postulated detonation event. Piping and valving are designed to resist dynamic pressures encountered in such an l event (PSAR Paragraph 11.3).
As appropriate, drains are installed in the system to !
remove condensate. These drains are designed to allow for removal of the condensate while still sealing the process stream +
l within the off-gas system. Another difference between the Black l Fox design and the Millstone plant is the replacement of loop i seals with trap seals on the drains. A schematic diagram showing l
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the differences between the two sealing methods can be found on the attached Figure 1. With regard to the capability to seal the process stream to prevent the loss of any gas outside y the system, the trap seal is a significant improvement. A ball float in the trap seal actuates the valve which allows water to flow out the trap. As the water level decreases, the valve shuts thus sealing the discharge. The design of the trap seal is such that should any overpressure occur in the system, rather than i
the water being forced out of the trap, the pressure would cause the valve to shut, sealing the discharge. Further, the discharge l
i j from the trap, rather than being routed to a sump where it is possible for any gases to escape to other portions of the plant, is routed to the main condenser. Therefore, in the unlikely event that a trap does not seal as designed, the gas stream simply returns to the main condenser where it started.
Thus, it can be seen that the design of the system ensures i
that throughout the process the hydrogen concentration is low enough i
that explosions as a result of such natural phenomena as lightning
( will not occur at the Black Fox Station.
It should further be noted that the Black Fox Off-Gas System is not a part of the reactor coolant pressure boundary and it is not required for the safe and orderly shutdown of the reactor. Even in the unlikely event of a gross failure of the Off-Gas System pressure boundary, the resulting calculated site boundary doses are a very small fraction of the 10 CFR Part 100 limits.
.The radiological analysis was performed by GE and the resul ts are l documented in PSAR Section 15.1.
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N UNITED STATES OF AMERICA pf g NUCLEAR REGULATORY COMMISSION . :. r
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD , j s ,-
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/ 1 os In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
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Testimony of Mr. Aaron J. Levine Concerning Question 2-1 l
September 25, 1978 l
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' j Testimony of Aaron J. Levine Concerning Question 2-1 My name is Aaron J. Levine and my business ~ address is 175 Curtner Avenue, San Jose, California. I am the Manager of Projects Licensing Unit 1 for the Safety & Licensing Opera-tion of the General' Electric Company,. Nuclear Energy Business j Group. A statement of my background and qualifications is attached as Attachment I to my testimony.
My, testimony addresses the following question posed by the Licensing Board in its Order of September 8, 1978:
Question 2-1: I l
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Do doubts exist about the independence and separability of i
thermal and hydraulic effects in the specific calculations ]
used to demonstrate compliance of Black Fox Station with Appendix K? l l
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Response
! The question of independence and separability of thermal and hydraulic effects in the core spray (CS) systems, one component of the Emergency Core Cooling System (ECCS) , relates to the ability of the CS systems to perform their safety function.
The CS systems have a set of nozzles arranged to distribute water over the reactor core in the unlikely event of a loss-of-coolant-accident (LOCA). Each fuel bundle must receive a
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minimum amount of coolant from the CS systems to provide the post-LOCA spray cooling assumed in the LOCA analyses for the Black Fox Station. i General Electric had conducted full scale distribution tests in air at atmospheric pressure to determine that the necessary minimum coolant would be provided to each fuel bundle. Later data indicated that the presence of steam and/or increased pressures in and above the upper core region could affect the distribution of flow from core spray nozzles. Prior to this, tests in air were accepted as an adequate demon-stration that sufficient coolant flow would be delivered to each fuel assembly to provide adequate cooling. I i
GE has undertaken a program to verify the adequacy of the present core spray design process. The results of this program will be factored into the design of the core spray headers for the Black Fox plant to assure that adequate spray flow is present in each fuel bundle.
The distribution of spray from the nozzles on the core spray l spargers are affected by thermal and hydraulic parameters.
, i l The thermal effects are related to the capacity of the nozzle fluid to condense steam. Large droplet sprays {
representative of those that are proposed for the Black j i
Fox CS systems are much less affected by the steam environ-ment than the small droplet sprays where these effects were l
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i first observed. Tests and analyses by GE and others indicate l l
that these thermal effects are virtually completed within a very short distance from the nozzle (approximately five nozzle j i
diameters), i.e., the spray fluid is near saturation temperature. j J
There is no interaction between spray flows from adjacent nozzles prior to this point. From that point, the trajectory j is controlled by hydraulic parameters. Therefore, it is possible to divide the spray flow into two regions which can be treated independently.
On January 19, 1978, GE presented this information to the NRC i
staff at a meeting in Bethesda, Maryland. In addition, GE proposed a program for testing of the BWR/6 core spray distribu-tion based on an empirical / engineering method which recognizes the separability of the thermal and hydraulic effects. The tests consist of: Full scale single nozzle tests; full scale )
I air tests; and multinozzle (full scale 30* sector) tests to be performed in a steam environment which will confirm the separa-bility of thermal and hydraulic effects. The program combines the test results with analytical tools and should result in a representation of the full core spray distribution which would exist following a postulated LOCA. The present schedule for completing the test portion of the program is mid 1979. In a letter dated February 3, 1978, the NRC staff summarized the results of that meeting and advised GE that its proposed program was acceptable. (See letter from Ross to Sherwood attached as Attachment II to this testimony.)
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In addition, the NRC staff has issued an interim evaluation report
- on GE Licensing Topical Report NEDO-20566-3 (General Electric Company Analytical Model For Loss-Of-Coolant Analysis in Accordance With 10CFR50 Appendix K - Effect E of Steam Environment on BWR Core Spray Distribution) which
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l concludes that in view of the spray flow margin believed j
! l l- to exist and the alternatives available in the unlikely l l
event that the tests do not confirm this margin, continued operation'and 2icensing in the interim period does not present undue risk to the health and safety of the public. l In view of the foregoing, any uncertainty regarding the i
! independence and separability of thermal and hydraulic l l l effects will be resolved in ample time to be factored into l l
[ the Black Fox Station core spray design.
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- See letter dated June 13, 1978 from Mr. Olan D. Parr i of the NRC Staff to Dr. G. G. Sherwood of the General
! Electric Company.
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i Attachment I-
. Manager Project Licensing I: Aaron J. Levine Education:
B.S. Marine Engineering, 1950, New York State Maritime College M.S. Mechanical Engineering, 1956,' Columbia University Additional Background
-Professional Engineer, California (License NU1198)
Nuclear Engineering Marine Engineer License (unlimited)
Third Assistant Engineer (1949-1956)
Instructor College Freshman Physics (1956-1958)
Connecticut State Technical Institute, Hartford, CT Graduate Courses - R.P.I. Extension, Hartford, CT (1957-1956)
Union College, Schenectady, NY (1960-1966)
General Electric Courses Fundamentals of Welding Engineering - 1965 Thermal and Hydraulic Analysis of Reactors - 1967 Nuclear Boiling Water Reactor Seminar.- 1968 Professional Business Operations - 1970 Management Practices Course - 1974 Kempner - Tregoe GENCO II - 1977 l Experience As Manager Project Licensing Unit I, I am responsible for the direction of project licensing engineers in all areas of nuclear power plant licensing. This includes, but is not limited to, the preparation of Preliminary and Final Safety
Analysis Reports and the resolution of Safety and Licensing f
concerns raised by an appropriate review agency on such reports.
This effort includes participation in the technical resolution of these concerns and the application of such resolutions on those projects directly under my responsibility. Some of those concerns include Off-Gas System evaluations. I have been in that position since February, 1973 and all the plants presently in my area of responsibility are BWR/6 Mark III projects. In addition, at the direction of the Manager BWR Licensing I am continuously responsible for certain generic licensing issues such as Emergency Core Cooling calculations,
! Reactor Safety Study (WASH-1400) information, and Core Spray Distribution.
I i
From 1967 to 1973 I was a Licensing Engineer directly responsible for the preparation of Safety Analysis Reports for requisition plants. In addition, I worked with operating BWR plants assisting them in the resolution of specific con-cerns associated with operating nuclear power plants and assessed the concerns for applicability to future plants.
From 1959 to 1967 I was a Design Engineer with Knolls Atomic Power Laboratory, where I was responsible for fluid flow and heat transfer design calculations associated with l
the primary system of nuclear power plants for shipboard 1
operation. I was responsible for the preparation and production 1
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of Power Plant Manuals for two propulsion plant designs.
These manuals describe the design and operation of the nuclear and marine propulsion systems. In addition, during that time I was also a Test Engineer on the prototype nuclear power plant of a U.S. Navy surface vessel. In that position I developed, conducted, and evaluated test procedures to assure acceptable operation of reactor primary systems and marine propulsion systems.
From 1956 to 1959 I was employed by Combustion Engineer-ing Incorporated as a Design Engineer where I performed design calculations in the areas of heat transfer and fluid flow for the primary system of a nuclear powered submarine.
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Attachment II p UNITED STATES y ', NUCLEAR REGULATORY COMMISSION j E WASHINGTON, D. C. 20655
%, .... / FEB 0 31976 Dr. G. G. Sherwood, Manager Safety and Licensing General Electric Company 175 Curtner Avenue San Jose, California 95125
Dear Dr. Sherwood:
This letter is to sunnarize our understanding of the agreements reached at the January 19, 1978 meeting between the NRC Staff and the General l Electric Company Staff regarding your Core Spray Distribution program, j We agree as described herein that the overall empirical / engineering method outlined by GE at the January 19 meeting is an acceptable method for veri- i fication of the currently assumed core spray distributions which are used to justify conservatism of the spray cooling heat transfer coefficients in ECCS-LOCA licensing calculations. The outlined method will: (A) utilize full scale, single nozzle, horizontal flow tests in steam; (B) determine multinozzle effects by a calculational superposition technique used in conjunction with full scale flow tests to be conducted in air (both sector tests and full 3600 tests); and (C) confirm multinozzle effects in steam 0
by independent, multinozzle (full scale 30 sector) flow tests to be con-ducted at the new Lynn, Massachusetts facility.
' We believe that this overall empirical approach should result in a repre-sentation of the full reactor core spray distribution that would exist following a LOCA. We agree with GE in the belief that this approach will then have to be applied to each different reactor size and design for which the full-reactor-core, post-LOCA spray distribution is to be determined.
Applicability of the method to each reactor design and size will have to be justified, including empirical tests if previous tests cannot be justified to be appropriate.
The NRC Staff's agreement (as described herein) with your approach of course assumes satisfactory results will be obtained from the confirmatory tests te be conducted at Lynn (see item "C" above and the acceptance criteria requested by Question 1 of the attachment). If satisfactory results are I not obtained, thermal effects (steam condensation) and hydrodynamic effects I (nozzle-nozzle interactions) may not be separable as postulated in your approach and would indicate that the empirical / engineering method outlined above is not acceptable, because the method will be justified presuming that the separability assumption is correct. We note that GE agreed to
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Attachment FEB 0 3197E )
Dr. G. G. Sherwood {
i test the separability assumption by conducting confirmatory tests at )
Lynn using a range of steam flows covering the range of steam flows representative of LOCA conditions for which spray cooling credit is assumed. 1 Our agreement (as described herein) with your approach also assumes that GE will provide satisfactory responses to the attached Requests for Additional Information. The Staff requires answers to Questions 3, 4 and 8 as soon as possible, as they involve documentation of the tech-nical bases for continued operation of facilities and for licensing i actions, and documentation of the proposed schedule for obtaining other test data and results.
Although the NRC Staff requires additional information concerning this issue, we believe there is a sufficient technical basis to permit con-tinued plant operation and licensing in the interim period while these additional tests and information are being developed. This interim conclusion is based on:
(1) The existence of a considerable safety margin between available and required spray flow indicated by preliminary analyses and measure-ments provided for each size BWR/l through BWR/5; l
i (2) the relative ease with which ECCS re-analyses could be performed to establish an acceptable power limit in the unlikely event that test results do not support the spray flows currently assumed; j (3) the possibility that plants under construction could modify their l.
spray nozzles or aiming pattern to provide a better spray distribution.
l if future test results indicate the desirability of such changes (par-ticularly applicable to the BWR/6, where the type of preliminary measure-ments referenced in (1) above are not yet available);
! (4) the existence of counter-current-flow-limiting phenomena in many L plants would provide a steam / water layer on top of the core which should force a more even distribution of the core spray; (5) the aforementioned empirical / engineering method is expected to pro-vide timely confirmation of the spray flow margin presently believed to exist.
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i Attachment Dr. G. G. Sherwood @ I' 0 N Please contact Dr. R. Woods of the NRR' Staff at (301) 492-8050 for further information or discussion of necessary schedules.
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Sincerely, ,
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(YL kll1.ll Darrell G. Eisen u't, Assistant Director for Operational Technology Division of Operating Reactors Office of Nuclear Reactor Regulation
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Den' wood F. Ross, Assistant Director for Reactor Safety Division of Systems Safety Office of Nuclear Reactor Regulation
Enclosure:
As stated
l Attachment l 1
REQUEST FOR ADDITIONAL INFORMATION GE CORE SPRAY DISTRIBUTION PROGRAM
'The items marked 1/19/78 below include all requirements identified at the January 19, 1978 GE-NRC Core Spray Distribution meeting in Bethesda.
The list below also includes a compilation of outstanding questions from all other question lists on this subject. The list below therefore re-places those previous lists. Those marked 12/15/77 below were previously asked at the NRC-GE meeting in San Jose on that date (see 12/29/77 minutes of that meeting); those marked 9/2/77 below were previously contained in a letter of that date from 0. Parr, NRC, to G. Sherwood, GE, concerning our review of Amendment 3 to NED0-20566 which addresses this subject.
You will note that questions 1-a, 1-b, 2, 3-a, 3-b, 3-d, 3-e, 3-f, 4, 5, 9, 10 and 12 of the 9/2/77 list are not included below. Although we still require the information requested by those questions, we believe more comprehensive information will be available in those areas when results are available from the new test facility at Lynn, Massachusetts; we there-fore defer our requirements for this information until that time.
- 1) (1/19/78) Provide a list of the General Electric Company's criteria for acceptance of the experimental results from the full scale, 300-sector-in-steam tests. The criteria should state qualitatively and quantitatively: a) what parameters will be measured and exactly how GE will determine whether the results verify or contradict the hypoth-esis that thermal and hydrodynamic effects are separable; and b) how the spray distribution under accident conditions will be conservatively represented in licensing analyses.
- 2) (1/19/78) Provide copies of the references cited by Dr. Sandoz at the 1/19/78 meeting regarding size of the steam condensing region sur-rounding a nozzle. Describe why GE believes that this data is appro-priate for application to a BWR spray system (e.g., that the geometry, spray flow rates, subcooling, and steam pressures are similar in the referenced tests and in BWR's following a postulated LOCA). Please include pictures of typical BWR single nozzle spray patterns in steam.
- 3) (1/19/78) Present a clear schedule of the overall program, including all experimental and analytical steps presently planned, to determine the predicted core spray distribution in a steam environment for the BWR/6 design and any other designs for which tests are currently planned. Include tests to be run at the Lynn facility, at the San Jose single nozzle steam facility, and at the Vallecitos full scale air facil i ty.
Attachment
- 4) (1/19/78) Discuss how and when GE will administrative 1y inform BWR licensees and applicants that GE has the capability of determining steam environment core spray distributions for various plant sizes and designs. For example, will GE volunteer to perform this deter-mination for older plant designs, or will GE issue a letter to older plants that the methods are available upon request, or will GE expect the licensees and applicants to make the initial inquiries regarding availability of the service, etc?
- 5) (1/19/78) We have heard several presentations regarding test programs to be accomplished at the Lynn, Massachusetts full scale 300 -sector steam test facility. Each presentation has emphasized investigation of either core spray (CS) distribution or counter-current-flow-limiting (CCFL) phenomena. in reality, the two are closely coupled. Please provide a written description regarding how the facility will be utilized to investigate the closely coupled relationship of CS and CCFL phenomena.
- 6) (1/19/78) Quantify the expected effects of the smaller amount of steam condensation that is expected to occur in the " hydrodynamic" region. Why does GE expect that this condensation will not invalidate the " separability" hypothesis? (The January 19 meeting disclosed that approximately 25% of the total condensation is expected in this region.)
- 7) (1/19/78) What air updraf t velocities will be utilized in future Vallecitos air-water full scale tests to simulate steam velocities in the post-LOCA environment? Justify the conservatism of the simulation, in::1uding magnitude and direction of the air flow with respect to pre-dicted steam magnitude and direction following a LOCA.
- 8) (1/19/78) Describe and document the results given at the 12/15/77 meeting regarding: a) minimum flow currently predicted per channel, without consideration of steam effects, for BWR/2 through BWR/6 plants, and b) provide a comparison of that minimum flow to " minimum required flow" (three different definitions should be used for this quantity as discussed at the 12/15/77 meeting). The material was
-presentedonslidesTWC-10(12/12/77) and TWC-11 (12/12/77) at the 12/15/77 meeting.
- 9) (12/15/77) Provide documentation regarding why GE believes steam and water flow patterns in the Lynn 30 0 test facility will adeouately represent the flow patterns that might be present in a full 360 0 reactor upper plenum following a LOCA. Include discussion of tests both with and without the " pie-shaped baffle" in place.
~.
Attachment I
- 10) (12/15/77) Quantify the conservatisms resulting from certain features of the present GE-ECCS-LOCA model, which were qualitatively discussed l
at the 12/15/77 meeting., , , ^
t l 11) (12/15/77) Provide the "CCFL delay vs. zero spray coefficient" tradeoff results (discussed in slides SCR-5 through SCR-8 (12/15/77)) for the sizes and types of jet-pump BWR plants whose results were not presented '
at the 12/15/77 meeting, and for the second most limiting break location for "LPCI-Modified" BWR's.
- 12) (Previous number 1-C, 9/2/77) The proposed tests do not include possible effects due to the different steam qualities that might be present under 1 l
various conditions. Water droplets entrained in the steam may change the interaction of the s'eam and the spray cone. Describe how GE plans to quantify such possible effects experimentally and/or analytically.
- 13) (Previous number 3-C, 9/2/77) For " Air Mockup of Steam Environment" tests that will employ Vee Jet Nozzles, will those nozzles be modified ,
l to simulate steam effects, and if so, how? '
l
- 14) (Previous number 6, 9/2/77) Provide the data for the lower sparger test discussed in the first paragraph of page 4-6.
- 15) (Previous number 7, 9/2/77) What updraft was present in the tests reported by Figures 4-5 and 4-6?
- 16) (Previous number 8, 9/2/77) There appears to be a discrepancy between Figure 4-6 and Table 4-1 on the minimum measured channel flow. For example, no channel in the periphery had a 3.4 gpm flow for VNC nozzles with deflectors in and no intermediate channel had a minimum flow of 6.8 gpm with VNC nozzles, deflectors out. Please explain the apparent discrepancy.
- 17) (Previous number 11,9/2/77) Justify your assumption that one-half of the " Appendix K" quoted core spray heat transfer coefficients can be used when spray flow to a bundle is below minimum design flow. You should provide results of experimental spray heat transfer coefficient measurements taken at lower spray flows. Also, you should quantitatively demonstrate that actual penetration of the assumed (lower) flow into the bundle is consistent with your CCFL data and correlations, under all conditions predicted by your ECCS calculations where this assumption I of lower heat transfer coefficients is made.
- 18) (Previous number 13,9/2/77) GE has changed the type of nozzles used in various BWR designs, for example between the BWR/3 and BWR/4. Please provide the rationale for such changes, including a description of any tests which indicated the desirability for the above mentioned change, and the results of those tests.
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i
l l
l In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 l
COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) ) I l
l l
1 Testimony of Mr. Aaron J. Levine Concerning Question 2-2 4
September 25, 197:
1
TESTIMONY OF MR. AARON J. LEVINE CONCERNING QUESTION 2-2 My name is Aaron J. Levine. My business address is 175 Curtner Avenue, San Jose, California. I am the Manager of Projects Licensing Unit 1 for the Safety & Licensing Operation of the General Electric Company, Nuclear Energy Business Group.
My statement of qualifications is attached to my testimony on I
Question 2-1.
My testimony addresses the following question posed by the Licensing Board in the Order of September 8, 1978: 1 2-2. What "recently discovered" errors may exist l
l in GE ECCS evaluation codes? Are there any errors other than those set forth in the SER i
at p. 6-10 of Appendix A?
The errors identified on page 6-10 of Appendix A to the Black Fox SER, as noted by the Board, have already been corrected and their effects documented for the Black Fox Station. GE has not identified any other errors in their ECCS codes.
There was identified a need for ensuring a consistency in the input to the' codes. Therefore, as part of a GI: continuing effort to improve and upgrade the ECCS calculations a GE initiated code input reverification program was put in place to review the inputs to all the ECCS codes and improve the accuracy of the calculations. This program is aimed at reviewing and refining all the inputs which are used in the ECCS codes by putting into ef fect procedures which standardize methods for determining those l
inputs and verifying that they are correct and applicable for the particular plant being analyzed. The reverification program has identified some changes, including some errors, which are applicable to BWR/6. It is important to note that the analyses which are conducted for the Preliminary Safety Analysis Report (PSAR) use preliminary data available at that time to characterize the reactor systems and the ECCS. An updated calculation using the inputs more characteristic of the final design is submitted to the NRC Staf f during their review of the Final Safety Analysis Report (FSAR). It is expected that there will be differences between the PSAR analyses and the FSAR analyses as a result of the evolution of the final design parameters and code changes which j could occur in the interim, as well as any changes identified as appropriate as a result of the reverification program.
The outcome of any particular change can result in either an increase or decrease in calculated peak cladding temperature (PCT) for any particular design. The calculated PCT is the controlling parameter for the BWR/6 ECCS evaluation. The reverifica-tion program is at this time approximately 80 per cent complete.
Some of the changes which have been identified are; improving the accuracy of the table entries in the low pressure range from which break flowrate is determined; modification of the cladding
. specific heat to account for the latent heat of the alpha to beta phase transformation of Zircalloy; correcting the upper tie plate flow area; improving the accuracy of the mass used in the shroud i
't J
node; improving the accuracy of the volume of the fluid used in the downcomer node. Other changes are made to reflect the evolving design such as use of natural uranium in the fuel, thickness of. channels, number of water rods, size of flow areas, etc.
GE has carried out sample calculations to assess the integrated effect of these changes. It is currently estimated that the' result of all the changes for a typical BWR/6 results in a PCT decrease of 20*F over the last calculated value. The program has progressed sufficiently to have a high degree of confidence that any other changes as a result of completing the !
remaining portion of the program will not result in an increase in the calculated PCT associated with the design basis accident.
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UNITED STATES OF AMERICA 'I #
NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC )
STN 50-557 COOPERATIVE, INC. ) l
) l (Black Fox Station, Units 1 and 2) )
1 l
1 i
l Testimony of Mr. William G. Gang j Concerning Contention 3 September 25, 1978
( TESTIMONY OF WILLIAM G. GANG CONCERNING CONTENTION 3 l My name is William G. Gang and I reside at 6428 Paso Los
( Cerritos, San Jose, California. I am the Project Manager for '
l
- the supply of the nuclear steam supply system components for the Black Fox Station working within the Nuclear Energy Projects Division of the General Electric Company. A statement of my qualifications is attached as Attachment I to my testimony.
The purpose of my testimony is to address Contention 3 which reads as follows:
Intervenors contend that the Applicant has not adequately demonstrated that the structures and components within the suppression pool have been designed to withstand the hydrodynamic forces of a high vertical water swell which results from the postulated Design Basis Accident for Black Fox 1. and 2.
The only component supplied for Black Fox by General Electric which would be affected by the hydrodynamic forces of a vertical pool swell are the hydraulic control units (HCU). The effects of such forces on structures and other components within the suppression pool are discussed in the testimony of Messrs. Guyot ,
1 and Thurman.
The HCU sit on a concrete floor 22 feet and 2 inches above the suppression pool surface. This floor is approximately 1 foot thick and is supported by wide-flanged steel bases approximately i 2 feet deep. The bottom of the beams are therefore, approximately 19 feet and 2 inches above the surface of the suppression pool.
The location and design of this floor are in the scope of the plant f
I designer, and this information has been obtained from discussions with~its structural engineering personnel.
GE's Confirmatory Test Program indicates that pure bulk pool swell terminates at levels lower than 18 feet above the suppression pool. Consequently, we are conservatively using 18 feet as the elevation-of bulk pool swell with a linear transition from water to froth in.the space of 18 feet to 19 feet above the normal pool surface. Therefore for design application, we have conservatively stated that the impact of water .from bulk' po'ol swell l would be applied at or below elevations 18 feet above the surface of the suppression pool. The hydrodynamic force felt by the beams and floor beginning at elevation 19 feet and 2 inches, as described above, would be a froth impingement load. This loading is discussed l
I i l in Section 4.1.6 of Mr. L. J. Sobon's testimony. The response of the l
floor would subsequently transmit a load to the bottom of the HCU.
The. magnitude of this load for Black Fox will be computed by the plant designer in his plant unique dynamic analysis ~and that analysis will be provided to GE for assessment of impact on the HCU.
GE and the plant designer will assure that the capability of the HCU will be adequate to withstand the transmitted load.
GE believes it is unlikely that the HCU will require modifica-tion to accommodate these forces. The HCU designed for earlier model reactors have been seismically tested up to 25g. The HCU used on EWR/6 is of the same configaration, but has a slightly larger accumulator and gas bottle. This HCU has been analyzed to the seismic capability of 18g at its natural frequency of 24 hertz.
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J It-is expected that the transmitted load will not exceed a couple of g's.
The HCU is therefore designed to withstand the hydrodynamic forces of a high vertical _ water swell which results from a .
postulated Design Basis Accident for Black Fox 1. and 2.
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ATTACHMENT I WILLIAM G. GANG EDUCATION:
BS, Engineering Science, U. S. Military Academy, 1966 MS, Nuclear Engineering, Massachusetts Institute of Technology, 1970 MBA, Finance, Fairleigh Dickinson University, 1974 EXPERIENCE:
I am a Nuclear Power Plant Project Manager, employed by'the Nuclear Energy Projects Division of the General Electric Company in San Jose, California. I am currently the General Electric Project Manager for the Black Fox Project. In this capacity I have overall responsibility for the supply of the Nuclear Steam Systems on each of the Black Fox Nuclear Power Plants planned for construction by the Public Service Company of Oklahoma near Inola, Oklahoma. My duties consist of directing and coordinating the design, fabrication and supply of the plant nuclear systems and components. I also am responsible for providing technical direction for construction, testing and startup of the reactor and asso-ciated systems.
Formerly, I was a senior project engineer and project engi-neer on the Black Fox Project. In these capacities, I provided technical support and guidance to the utility customer and l l
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! my agents on the twin reactor project. My further respon-sibilities were to review and approve engineering documents, to define project requirements to the Engineering Department, to i
ensure co'.1 tract commitments were met and to coordinate any changes I i
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(using the Licensing Section). I was also responsible for planning and scheduling these activities to support the over-l all project schedule. '
Before taking a position at General Electric, I worked from 1971 to 1974 as an Assistant Professor in the Department
.of Physics at the United States Military Academy, West Point, New York. I taught modern physics, lectured on nuclear power, and restructured and taught an advanced senior elective 1
course in nuclear physics. From 1966 to 1971, I served as an officer in the U. S. Army Corps of Engineers. During these years, I worked as an engineer command and staff officer respon-sible for many horizontal and vertical construction projects both in combat and in peacetime atmospheres. Today, I remain active in the U. S. Army Reserve.
In 1966, I was awarded a Nuclear Engineering Fellowship by the Atomic Energy Commission after national scholastic competition.
As an officer in the U. S. Army, I am the recipient of three Bronze Star Medals, two Air Medals, the Vietnamese Gallantry Cross, and the Army Commendation Medal.
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NUCLEAR REGULATORY COMMISSION i {l ,' ;
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD Li 3 '
In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE, INC.,) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPERATIVE, INC. )
)
(Black' Fox Station, Units 1 and 2) )
i 1
Testimony of Mr. Lowell E. Thurman
, Concerning Contention 3 l
l September 25, 1978 1
TESTIMONY OF LOWELL E. THURMAN CONCERNING CONTENTION 3*
DESIGN OF MECHANICAL COMPONENTS SUBJECTED TO HIGH VERTICAL WATER SWELL LOADS My name is Lowell E. Thurman. I reside at 10400 Walmer, Overland Park, Kansas 66212. I am employed by Black & Veatch Consulting Engineers as the Supervising Engineer of the Pipe Stress Analysis Group. I received my formal engineering education i 1
at the Missouri School of Mines and Metallurgy and received a BS degree in Mechanical Engineering. I am a Registered Professional Engineer in the S' tate of Virginia. A statement of my background and qualifications is attached as Attachment I to my testimony.
My testimony will deal with the design and analysis of mechanical components (piping, valves, supports, etc.) within I
the scope supply of Black & Veatch and Public Service Company l of Oklahoma located in the suppression pool area. This testimony identifies how our design will interpret the loads presented in Appendix 3C of the Black Fox Preliminary Safety Analysis Report l
Contention 3 reads: l Intervenors contend that the Applicant has not adequately demonstrated that the structures and components within the suppression pool have been designed to withstand the hydrodynamic forces of a high vertical water swell which result from the postulated Design Basis Accident for Black Fox, 1 and 2.
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l 2-(PSAR) and demonstrates that mechanical components will be adequately designed. These loads were established by the General Electric Company (GE) and the NRC Staff after review of data prepared by GE. My testimony does not address design of the Hydraulic Control Units which are in the GE scope of supply and are discussed in the Testimony of William Gang.
Mechanical coiaponents are located in the following suppression pool areas which are shown on figure 1:
(1) Between the basemat and the suppression pool surface.
Components in this are completely submerged in the suppression pool.
(2) Transition area which includes parts of the area between the basemat and the Hydraulic Control Units' (HCU) floor. Components in this area are partially submerged in the suppression pool.
(3) Between the suppression pool surface and the bottom of the HCU floor.
(4) Between the bottom of the HCU floor and approximately 10 feet above the HCU floor at elevation 600'-7 3/4".
All mechanical components in-the suppression pool area will be designed for the following list of loads (hereinafter referred to as " generic loads"):
(1) ' SRV loads including structural feedback and building motions (hereinafter referred to respectively as " inertial" and " anchor motions").
1 (2) Dead load (3) Operating Basis Earthquake (OBE) inertial and anchor motions !
l (4) Safe Shutdown Earthquake (SSE) inertial and anchor motions (5) Internal pressure (6) Thermal expansion and anchor motions Components in Area (1) include suction strainers from three Residual Heat Removal (RHR) pumps, one High Pressure Core Spray (HPCS) pump, one Low Pressure Core Spray (LPCS) pump, and one Reactor Core Isolation Cooling (RCIC) pump. In addition, main steam Safet Relief Valve (SRV) discharge piping and quenchers are located in Area (1). These components will be designed using the loads defined in Section 8 of Appendix 3C of the Black Fox PSAR.
The loads considered in Section 8 include the generic loads discussed above plus vent clearing, vent / coolant interaction and pool swell loads due to a postulated loss-of-coolant accident (LOCA), and safety relief valve loads discussed in Attachment A I to Appendix 3C. Hydrodynamic mass effects will be considered in the natural frequency and force calculations for these components.
Pool swell impact and froth loads need not be considered since components are completely submerged.
Components in Area (2) include Emergency Core Cooling System I (ECCS) SRV and test return piping. The portion of piping which is submerged will be designed as indicated in Area (1) above and the piping above the suppression pool surface will be designed for the loads specified in Section 10 of Appendix 3C of the Black
i 1
Fox PSAR as indicated in Area (3) below.
- Components in Area (3) include portions of the ECCS piping from the pump rooms to the reactor vessel. These components will be designed using the loads defined in Section 10 of Appendix 3C of the Black Fox PSAR. The loads will include the generic loads discussed above plus pool swell impact, drag and
. fallback loads. A dynamic time history analysis will be performed I to account for the dynamic effects using histograms specified in l
Appendix 3C of the Black Fox PSAR. {
t Components in Area (4) include portions of the ECCS piping from the pump rooms to the reactor vessel. These components l will be designed using the loads defined in Section 12 of Appendix l l
3C of the Black Fox PSAR. The loads will include the generic I
. loads discussed above plus froth impact, drag and fallback loads.
A dynamic time history analysis will be performed to account for dynamic effects using the histogram specified in Appendix 3C of the Black Fox PSAR.
All safety class components will be designed to meet the requirements of the applicable section of ASME III considering l l
all potential event combinations. The initial analyses will include design margins and appropriate load combinations and service level limit designations to insure a satisfactory final design.
The following steps will be taken to insure the suppression pool loads are properly analyzed:
(1) Pipe routing will be performed to minimize the amount of piping in the suppression pool swell and froth areas.
(2) Loading criteria established and documented in Appendix 3C of the Black Fox PSAR, and approved by thelNRC Staff will be conservatively applied in the i
s analysis of mechanical components.
stre's (3) NRC accepted hydrodynamic and mechanical design stress
- analysis procedures
- and techniques as outlined in
-Appendix 3C of the Black Fox PSAR will be used to evaluate the design adequacy of mechanical components. {
l The foregoing discussion demonstrates that the mechanical I l
components located in the suppression pool area-can be adequately j
' designed for all loads including pool swell.
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n ATTACHMENT I POSITION AT BLACK AND VEATCH:
Supervising Engineer, Pipe Stress Analysis Group l EDUCATION:
l BS Mechanical Engineering, 1963
(. Missouri School of Mines and' Metallurgy i
ADDITIONAL EDUCATION AND TRAINING:
i Westinghouse PWR Seminar, 1974 B&V Seismic Analysis Seminar, 1974 l B&V Nuclear Equipment Design Seminar, 1974 B&V Seismic Specification Seminar, 1975 Newport-News Shipbuilding and Dry Dock Company Management Training, 1970 General Electric Seminar on Structures Submerged in the Suppression Pool, 1977 PROFESSIONAL REGISTRATION:
Professional Engineer, Virginia, PE 4455, 1968 EXPERIENCE:
As Pipe Stress Analysis Group Supervising Engineer, I am responsible for stress analysis and support design for all piping systems engineered by the Black and Veatch Power Division.
I am also responsible for preparation of mechanical technical and procurement specifications for support components.
Since my assignment to BFS in 1974, I have supervised and contributcd to the preparation of the following items:
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(1) Chapter 3.2, 3.6, and 3.9 of the PSAR l (2) Various Component Design Specifications (3) Initial stress analysis and support design for Nuclear Island and Balance of Plant Piping Systems (4) Initial pipe break postulations, fluid dynamic blowdown analyses and pipe whip restraint designs.
In addition to my BFS assignment, I am also responsible for the superviolon of all other Power Division Stress Analysis and Pipe Support Designs. This work includes numerous large and small fossil fueled generating stations. I assumed my present position shortly af ter joining Black and Veatch in 1973.
Prior to joining Black and Veatch I was employed by Newport News Shipbuilding and Dry Dock Company and spent ten years in i l
various assignments in the Navy Nuclear Program. These assign-ments included one and one half years as a Systems Engineer, one l year as a Mechanical Test Engineer, two years as a Design Engineer, l one and one half years as a Senior Design Engineer and four years as a Piping Stress Analysis Design Supervisor.
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. in o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 l
AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
)
(Black Fox Station,' Units 1 and 2) )
- p. 52
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- @ S -
Testimony of Mr. William G. Gang '3. 0
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Concerning Question 5-1 ed' #
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I September 25, 1978 l
TESTIMONY OF WILLIAM G. GANG CONCERNING QUESTION 5-1 My name is William G. Gang and I reside at 6428 Paso Los Cerritos, San Jose, California. I am the Project Manager for the supply of the nuclear steam supply system components for j the Black Fox Station working within the Nuclear Energy Projects Division of the General Electric Company. A statement of my qualifications is attached to my testimony on Contention 3.
The purpose of my testimony is to address the following question posed in the Licensing Board's Order of September 8, 1978:
5-1 Is the treatment of vertical motion in an earthquake of importance t se design of pressure vessel supports and pedestals and, if so, has it been accommodated?
I The General Electric Company is responsible for furnishing the pressure vessel support skirt for the Black Fox Station, and I will address the question in that context. Mr. David Guyot of Black & Veatch, the plant designer, will address the question as it applies to the reactor pedestal. The treatment of vertical motion generated during the course of an earthquake i
is one of several important considerations in the design of a pressure vessel support skirt, and the load is accommodated in the design. It is noted that the consideration of seismic
loading in skirt design is a routine engineering task and is not a new inclusion in the design process.
The design process for the Reactor Pressure Vessel (RPV) skirt begins when GE selects a design basis envelop of loads.
These loads and their magnitudes are based on a conservative generic analysis which envelopes all expected soil and seismic conditions to be experienced at any site location. This envelope is conservatively established and includes thermal, deadweight, vertical and horizontal seismic, normal operation l piping reaction, pipe rupture, and live (reactor scram) loads. ;
l GE then provides this set of design loads to the RPV skirt l l
vendor. The vendor, using the GE design basis envelope, designs the skirt and performs a stress analysis using the GE envelope loads. The vendor then provides GE with a stress report certified to meet the stress limits and analytical methods of Section III of the ASME Code. This step in the design process demonstrates compliance with the acceptance criteria of Section III of the ASME Code.
Simultaneously with the GE-RPV skirt vendor interaction, GE provides the plant designer, on the reactor interface control document, with the maximum allowable loads at the RPV skirt-to-pedestal interface, l
The plant designer' performs a dynamic analysis for Black i
Fox Station which includes horizontal and. vertical seismic, LOCA and safety relief valve actuation loads, and verifies that the maximum allowable loads specified by GE at the skirt-I pedestal interface have not been exceeded. GE is then provided i
a set of dynamic data (response spectra and/or time history) at'the base of pedestal. That data is used by GE to perform an analysis of the NSSS equipment including the skirt to ensure that the selec.ed design basis envelope has not been exceeded.
1 GE and the plant designer work together to ensure that the design basis envelope is satisfied. Certain recently identified feedback response loads which result from the sup-pression pool related hydrodynamic events (e.g., LOCA, safety l
relief valve actuation) are now also being included in the design evaluation of the vessel support skirt. The evaluation of these loads allows GE and the plant designer to confirm the adequacy of the preliminary skirt design, part of the normal sequence of events between preliminary and final design as described above.
l The conservatisms in the preliminary skirt design are such that changes to this design from the confirmatory analysis are not expected. For example, the peak seismic ground accel-eration for the design for the skirt was a .3g SSE, whereas at the Black Fox site this same parameter is only .12g. Further, 1
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the normalized site design response spectra are less than those given in Regulatory Guide 1.60, which were used for the '
preliminary design. GE has done some preliminary calculations l involving the previously mentioned hydrodynamic events. These calculations were done for a typical BWR 6 Mark III design using the above conservative site parameters. These calculations demonstrated adequacy of the skirt design.
From the above discussion of the RPV skirt design process, it is demonstrated that the vertical component of an earthquake is important and has been accommodated in the design.
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD s ,- ,
- f In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
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(Black Fox Station, Units 1 and 2) )
Testimony of Mr. David Guyot' Concerning Contentions 3 and 16 and Questions 5-1 and 12-3 September 25, 1978 1
TESTIMONY OF DAVID GUYOT CONCERNING CONTENTIONS 3 AND 16 AND QUESTIONS 5-1 AND 12-3 My name is David F. Guyot. I reside at 10315 Long Street, Overland Park, Kansas. I am Project Engineer, Structural Systems, for the Black Fox Station design project within the Civil-Structural Engineering Department at Black & Veatch Con-sulting Engineers in Kansas City, Missouri, Architect / Engineer-ing firm employed by Public Service Company of Oklahoma. A statement of my background and qualifications is attachad as Attachment I to my testimony.
Part I of my testimony deals with the following loads identified in Contention 16 which relate to the design of a Mark III pressure suppression containment:
(1) vent clearing (2) vent / coolant interaction (3) pool swell (4) pool stratification (5) pressure loads and flow bypass Part II of my testimony addresses Contention 3 regarding the design of structures located within and above the suppres-sion pool and their ability to withstand the pool swell loads identified in the first portion of this testimony.
Part III of my testimony addresses Licensing Board Ques-tion 5-1 regarding the design of the Reactor Pressure Vessel pedestal and its ability to withstand the loads resulting from
q l
l f the Design Basis requirement of 10 CFR Part 50, Appendix A, Criterion 2 relating to earthquakes.
i Part IV of my testimony addresses Licensing Board Question l 12-3 regarding the spent fuel design and its ability to with-stand a vibratory ground motion event, including the vertical j i"
effects of such an event.
The design af structures, systems, and components for the Black Fox Station in and near containment are based on response to interaction loads developed by various phenomena, including l
seismic events, operational events and postulated accidents. J These interaction loads, such as seismic and loads due to actuation of safety relief valves have been included in the design either explicitly as they have been identified for l l
the design of structures or implicitly as design margin in the design of some components such as the Pressure Vessel skirt. In all cases, the final design will be demonstrated to adequately meet the loading requirements and ensure the safety and welfare of the general public.
l l
3-Part I Contention 16*
Containment Dynamic Loads The purpose of this part of my testimony is to (i) document the establishment of load criteria for Black Fox Station regard-i ing the phenomena which specifically relate to the Mark III j pressure suppression containment, namely, vent clearing, vent /
coolant interaction, pool swell, pool stratification and pres-1 sure loads and flow bypass; and (ii) address the application of the generic Mark III containment load definitions established by the General Electric Company to the Black Fox Station con-tainment system. The development of technical information and documentation thereof for the generic (reference) Mark III con-tainment system is discussed by Mr. Sobon in his testimony.
On August 1, 1975, General Electric Company transmitted advance copies of General Electric Company Information Report NEDO-ll314-08 (Preliminary), Reference 16-1, to the NRC staff.
On August 29, 1975, General Electric Company transmitted
- Intervenors contend that the Applicant has not established the integrity of the Mark III containment in that the following items have not yet been resolved:
(1) vent clearing; (2) vent / coolant interaction; (3) pool swell; (4) pool stratification; and (5) pressure loads and flow bypass
L Amendment 37 to the Standard Safety Analysis Report (GESSAR) 238 Nuclear Island, Docket STN-50-447, which presented General Electric Company Information Report NEDO-ll314-08 (Final),
Reference 16-2, to the NRC staff ~as Appendix 3B to GESSAR.
t Both documents address the load definition for the postulated 3 l loss-of-coolant accident and safety relief valve events. The final version also addresses the application of containment i I
dynamic loads to affected structures of the General Electric ]
Company reference Mark III containment. Subsequently, General Electric Company filed Amendments 40 and 43 to the GESSAR docket which updated the original Appendix 3B submittal.
, Although the final version of NEDO-11314-08 was placed f
on the GESSAR docket by the General Electric Company as Appendix 3B, the Nuclear Regulatory Commission staff in a letter dated November 23, 1976 from O. D. Parr to B. H. Morphis,
. i l Reference 16-3, required the utilization of NEDO-ll314-08 I (Preliminary) as the design bases for containment dynamic loading specification for the Black Fox Station. Since the !
NEDO-11314-08 (Preliminary) document did not contain complete and. current applicable data and information, the Applicant prepared a load definition report unique to the Black Fox Station for containment dynamic loads. This document was designated as Appendix 3C, and it was submitted to the NRC staff as a portion of Amendment No. 8 to the Black Fox Station Preliminary Safety Analysis Report. Appendix 3C is a compilation l l
of References 1 and 2, and the NRC staff Safety Evaluation Report, including Supplement 1 to the Safety Evaluation Report, for the GESSAR docket, References 16-4 and 16-5, respectively.
Appendix 3C incorporates the portions of each referenced document which are acceptable to the NRC staff and are appli-cable for the Black Fox Station. Appendix 3C is identical to Appendix 3B of the GESSAR docket except for the following items:
(1) Appendix 3C includes changes which have occurred subsequent to submittal of Amendment 43 on the GESSAR docket, particularly the incorporation of the resolution of pool swell load definition docu-mented in GESSAR Safety Evaluation Report, Supple-ment B, Reference 16-6.
(2) Appendix 3C includes only data presented in Appendix 3B regarding the 238 standard containment configuration which is being used for the Black Fox Station. Information and data concerning other standard containment configurations was deleted.
(3) Appendix 3C addresses the unique design features of the Black Fox Station including the addition of an elevator inside the containment and utiliza-of a lower design temperature for the service water system.
(4) Appendix 3C incorporates a more conservative design procedure for. evaluating the loads on structures l and components submerged in the suppression pool. ,
(5) Appendix 3C corrects typographical errors which exist in Appendix 3B.
1 The differences identified in Item (3) above are within the enveloping design parameters established for the reference Mark III containment and therefore do not invalidate the appli-cability of this data and information for the design of Black Fox Station. These items will be further addressed herein-after in greater detail.
In General Electric Appendix 3B and in other General Electric Company information documents, GE identifies the critical geometry and establishes other parameter limits for the standard Mark III containment using the 238 inch diameter i Reactor Pressure Vessel. As evidenced by a comparison of information provided in Appendix 3B and Appendix 3C; particu-larly Figures 2.2-2, 2.2-4, 2.2-5, A4.1, A4.2, and A4.3; Sections A2.0, A7.2, and A10.1; and Tables A4.4 and A10.4, the Black Fox Station containment design is identical to the General Electric Company reference containment except as identified under Item (3) above.
The addition of the elevator at and above elevation 592 feet 10 inches inside the containment only affects the available vent area at the HCU floor. This change in vent area influences the differential pressure that can occur below this elevation. Figure 6.16 of Appendix 3C is provided to account for differences in floor designs. Allowing approx-imately 50 square feet for the loss of vent area due to the addition of the elevator and using Figure 6-6, the differential pressure due to loss of the vent area does not exceed the 11 psid specified for the reference containment design.
Therefore, the addition of the elevation to Black Fox Station does not affect the specified load criteria provided in Appendix 3B.
The use of 95*F design water temperature to service the Emergency Core Cooling System will not significantly effect the peak calculated, long-term pressures and temperatures following postulated loss-of-coolant accidents. For these long-term design conditions, we utilize design temperatures and pressures which exceed the peak calculated pressures and-temperatures. Specifically the peak design pressure used in the containment design is 15.0 psid compared with the peak l
calculated pressure, based upon the design service water temperature, of 9.8 psid. Therefore, the use of lower service water temperature does not influence containment design loads.
_g-Mr. Sobon's testimony, including Appendix 3B, establishes that load definitions for the generic Mark III containment design have been appropriately and adequately established considering all phenomena associated with the loss-of-coolant accident events and anticipated safety relief valve transients.
These load definitions have been approved by the NRC staff as a part of its Preliminary Design Approval for the GESSAR docket. My testimony demonstrates that the Mark III containment proposed for the Black Fox Station is essentially identical to f the Containment 238 inch diameter reactor pressure vessel j described by Mr. Sobon and, therefore, the generic load definitions are appropriate for use and applicable to the Black Fox Station containment design. Thus, it should be concluded that the Black Fox Station containment design ade-quately accounts for the phenomena identified by the Inter-venors in Contention 16.
l i
The applicant is employing these aforementioned design bases in the design of Black Fox Station. Based upon pre-liminary calculations performed by the General Electric Company l
and the Applicant, it has been determined that these phenomena will affect structures, systems, and components which are in direct interaction with the loads at the point of load appli-cation. Additionally, the Reactor Building structure also responds to the effects of some of these loads. This structure response causes feedback loads to be transferred to other
structures, systems and components which are not directly affected by these phenomena. The actuation of the safety relief valves is the principal loading phenomena which causes the feedback effects. The design of Black Fox Station will consider both the direct and the indirect effects of these loads.
PART II ,
1 Contention 3*
Design of Structures for i Affects of Pool Swell The purpose of this part of my testimony is to address the design of structures located within and above the suppres-1 sion pool,** particularly with regard to their ability to withstand the hydrodynamic forces of vertical swell of water within the suppression pool which result from a postulated loss-of-coolant accident. My testimony concerning Contention 16 documents the established load criteria for Black Fox Station regarding the pool swell phenomenon.
- Contention 3 reads:
l Intervenors contend that the Applicant has not adequately demonstrated that the structures and components within the suppression pool have been designed to withstand the hydrodynamic forces of a high vertical water swell which result from the postulated Design Basis Accident for Black Fox, 1 and 2.
- The design adequacy of components within the context of Contention 3 is discussed by Messrs Gang and Thurman in their testimony.
J J
10 -
With the exception of the attachments and platforms identified below, the drywell and the containment vessel form i
the vertical sides-of the suppression pool within and above J
the, pool. As such, loadings which are imparted to these boun- .l l
dary structures due to the vertical motion of the pool swell l l
are limited to those loads which are transferred to the struc- 1
}
1 tures by the attachments and platforms attached thereto. (The attachments and platforms are discussed in the next paragraph.)
. The drywell and .he containment vessel are designed for these i
)
effects. The loading combinations and acceptance criteria for the drywell and the containment vessel are identified in Sub- ,
sections 3.8.3.1 and 3.8.2 respectively, of the Black Fox Station Preliminary Safety Analysis Report.
Platforms, stairways, and attachments to the drywell and containment are generally indicated in Figures 14.9, 14.13, 14.15, and 14.16a of Appendix 3C. These figures indicate the preliminary. arrangement of the structural members and sizes of structural sections. The structural concrete attachments to the drywell are designed in accordance with the loading combinations and acceptance criteria specified in Subsections 3.8.3.6 and 3.8.3.4.3.2 of the Black Fox Station Preliminary Safety Analysis Report. The structural steel attachments, stairs, and platforms are designed in accordance with the load-ing combinations and acceptance criteria specified in Subsec- ,
I' tions 3.8.3.6 and 3.8.3.4.3.'3 of the Black Fox Station Pre-liminary Safety Analysis Report.
1 l
i 1
Although Contention No. 3 only addresses the ability l 4
of the structures within and above the suppression pool to I withstand the pool swell loads, the design of these structures must also consider other loadings from phenomena in addition to pool swell that can occur concurrently with the postulated I
loss-of-coolant accident. Therefore, the subject structures will be discussed hereinafter considering all of the loading combinations which are applicable to their design.
The individual loadings defined in the loading combina-tions for the design of the structures and attachments identi- 1 f
fled above were expanded with regard to the actuation of safety relief valves and the effects of the postulated loss-of-coolant accidents using the data presented in Figures 6.1, 6.2, and 6.3 and more specifically Figures 8.1, 9.1, and 10.1 i
of Appendix 3C. These figures present the temporal distribu-tion of the phenomena associated with the postulated loss-of-coolant accident. In addition, the bar charts identify other loading conditions such as seismic accelerations and safety relief valve actuation events which occur during the particular postulated loss-of-coolant accident. For the identified phenomena, i the effects of safety relief valve loads were considered as live loads (L) and the effects of pipe break accidents were considered as accident pressure (Pa and Ta).
1
)
Considering the above loading combinations which include pool swell affects and acceptance criteria, the design margins, i.e., the amount by which the allowable design stress exceeds the calculated stress, in two typical steal beams, one located at elevation 576 feet 7 inches and the second at elevation 592 feet 10 inches, for the governing loading combinations which include Pa and Ta (Equations (2)a4, (2)a5, and (2)a6 in Subsection 3.8.3.4.3.3) is greater than 70%. This design margin is typical for all the structural elements located within the area affected by the pool swell. -These preliminarily calculated design. margins were established considering the dynamic response of the structures by combin'ing the peak stresses from dissimilar events, considering the full effects of pool swell, and the structural feedback effects of the safety-relief valve actuation.
With respect to other issues associated with the suppression pool response due to the postualted loss-of-coolant accident event and previously addressed by the Intervenors, the following additional issues are addressed.
Regarding the elevator within the containment, the design of the elevator has been modified to preclude the referenced
. control interlock system which directed the elevator to an l l
upper area of the containment. By utilizing a partially open pit area beneath the elevator and above the suppression pool at the platform at elevation 592 feet 10 inches, it was necessary 1
for.the elevator to remain at an elevation higher than the area affected by pool swell. The current elevator design has raised the elevator bottom such that a froth impingement shield now protects the elevator and its associated appur-tenances from the direct effects of the vertical pool swell.
This enables the elevator to operate at any elevation.
Regarding the available vent area at the platform at elevation 592 feet 10 inches, we have reviewed the differential pressure which results at this platform due to pressurization of the volume above the suppression pool surface and below 1
the platofrm at elevation 592 feet 10 inches. This data is )
discussed in Part I of this testimony.
Regarding other structures and attachments within the affected pool swell areas, there are three additional major structural appurtenances, the drywell personnel air lock, the drywell transverse in-core probe (TIP) station, and the containment equipment hatch at elevation 576 feet 7 inches.
These appurtenances are designed for the applicable load effects due to pool swell in accordance with the data presented in Appendix 3C. The drywell personnel air lock and the TIP sta-tion are protected by the inipactive ef fects of the pool swell by deflector structures which extend beneath the surface of the suppression pool. Other major structures which are attached to the drywell or the containment such as the steam line piping l
r ---
. process tunnel, the containment personnel air locks, and the fuel transfer tube are located at or above the bottom of the platform at elevation 592-feet 10 inches. If required by their location, these appurtenances are designed for the effects of l
the froth impingement as discussed in Appendix 3C.
In conclusion, Part II of my testimony demonstrates that the Applicant has appropriately considered the vertical pool i
swell loads which result from the postulated LOCA and other concurrent loadings in the design of the structures within i
and above the suppression pool.
i Part III Question 5-l*
Design of Reactor Pressure Vessel Pedestal The purpose of this part of my testimony is to address i
the design of the reactor pressure vessel pedestal, particularly with regard to its ability to withstand the loads resulting from ;
the Design Basis requirements of 10 CFR Part 50, Appendix A, Criterion 2 relating to earthquakes.
L I
- Question 5-1 reads:
Is the treatment of vertical motion in an earthquake of importance to the design of the pressure vessel supports and pedestals, and if so, has it been accommodated?
r l The reactor pressure vessel pedestal provides support l for the reactor pressure vessel and the biological shield wall. The pedestal consists of two concentric steel cylinders l
joined by radially placed stiffeners. The annulus formed by the two steel concentric cylinders will be filled with con-crete. At the top of the pedestal a bearing plat is attached j
)
to receive the reactor pressure vessel. The reactor pressure !
vessel is bolted to the pedestal.
I The reactor pressure vessel pedestal is subject to the interface loading between the pedestal and its attachments such as the reactor pressure vessel, the biological shield wall, and other attachments. The pedestal is also subject {
to external loadings such as vibratory ground motion event and feedback effects due to safety relief valve actuation. (
The pedestal is also subject to the effects of the postulated l loss-of-coolant accident. Because these phenomena have been identified and the acceptable design bases established in a time frame such that their effects may be considered in che design of the Black Fox Station, the BFS pedestal is being designed to include all of the effects.
The reactor pressure vessel pedestal is designed in accord-ance with the loading combinations and acceptance criteria specified in Subsections 3.8.3.4.3 and 3.8.3.4.5 of the Black Fox Station Preliminary Safety Analysis Report.
Although the contention only addresses the ability of the pedestal to withstand the seismic event, the design of this structure must consider all other loadings which can occur concurrently with the design basis seismic events. Therefore, the pedestal will~be 5'iscussed hereinafter considering all l 1
of the loading conditions which are applicable to its design.
l t
Considering the above loading combinations and acceptance criteria, the design margins, i.e., the amount by which the i
allowable design stress exceeds the calculated stress, at the l i
bottom of the inner steel cylinder shell plate for the two 1
governing loading combinations (Equations (1)a4 and (2)b5 in Subsection 3.8.3.4.3.3) is greater than 15%. The design margin 1
for other portions of the pedestal will generally exceed this value since this is the critical area of the pedestal design, i
l These preliminarily calculated design margines were I established considering the structural response of the pedestal, reactor pressure vessel, and biological shield wall. The design margins were calculated by combining the peak stresses from dissimilar events and considering the full effects of the seismic event, including vertical motion, the structural feed-back effects of the safety-relief valve actuation, and the full effect of the asymmetric loading due to pressurization of the annulus between the biological shield wall and the reactor pressure vessel during the postulated loss-of-coolant accident.
4
Therefore, I conclude that the reactor pressure vessel pedestal can be adequately designed to accommodate the loadings which can or may result.
l l
The Applicant also will provide interface loading data to the General Electric Company to support the design verification )
J of the reactor pressure vessel skirt. This interface loading )
J data will include the effects of all phenomena which may be transmitted to the reactor pressure vessel through the uupport-1 ing pedestal, including the seismic event and the feedback I effects of the safety-relief valve event and the postulated loss-of-coolant accident. For the seismic event, the Applicant will employ the methods outlined in Section 3.7 of the Black Fox Station Preliminary Safety Analysis Report to develop 7 input data at the skirt-to-pedestal interface data. For the feedback loads, interface data will be calculated employing finite element approach to evaluate foundation structure interaction.
For all loads the Applicant will provide General Electric I
Company the necessary horizontal and vertical motion input data in the form of time history data or response spectra. Floor l response spectra inputs will be generated from the time history method, taking into account variations in parameters by peak broadening.
l In conclusion, Part III of my testimony demonstrates that the Applicant has appropriately considered the vertical input motion, which result both from a seismic event and the feedback effect of a safety relief valve and postulated loss- j of-coolant accident event, in the design of reactor pressure vessel skirt and support pedestal.
l Part IV Question 12-3*
Seismic Affects on Spent Fuel Pool Design Part IV of my testimony addresses ASLB Question 12-3 regard-I ing the spent fuel pool design and its ability to withstand a vibratory ground motion event, including the vertical effects of such an event.
The design of Black Fox Station, particularly for the purpose of this discussion -- the Fuel Building fuel pools, ,
include all of the effects of a vibratory ground motion event.
The procedures for performing these analyses due to this event is documented in Section 3.7 of the Black Fox Station
- Question 12.3 reads:
l Is treatment of vertical motion in an earthquake a matter
! of importance to the spent fuel pool design, and if so, has it been accommodated?
l l
l Preliminary Safety Analysis Report. In this section, particu-larly subsection 3.7.2.1, the use of seismic models which consider loadings due to two orthogonal horizontal directions ;
and the vertical direction are discussed in some detail.
l Therefore, the Applicant has demonstrated in the Black ]
Fox Station Preliminary Safety Analysis Report the importance of the vertical input motion due to the seismic event and l 1
I documented the procedure being utilized to consider its effect.
References Reference Description 16-1 General Electric Company Information Report NEDO-11314-08 (Preliminary), Mark III Containment Dynamic Loading Conditions, July, 1975.
16-2 General Electric Company Information Report NEDO-ll314-08 (Final), Mark III Containment Dynamic Loading Conditions, July, 1975.
16-3 Letter dated November 23, 1976 from Olan D. Parr, Nuclear Regu latory Commission, to B. H. Morphis, Public Service Company of Oklahoma.
Subject:
Black Fox Station Containment Dynamic Loading Criteria.
16-4 Nuclear Regulatory Commission, " Safety Evaluation l
Report Related to the Preliminary Design of the GESSAR - 238 Nuclear Island Standard Design,"
l l NUREG - 75/110, Washington, D.C., December, 1975.
f 16-5 Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Preliminary Design of the GESSAR - 238 Nuclear Island Standard Design--
Supplement No. 1," NUREG - 0124, Washington, D.C.,
September, 1976.
16-6 Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Preliminary Design of the GESSAR - 238 Nuclear Island Standard Design--
Supplement No. 3, " NUREG - Washing ton, D.C.
'l 1
' Attachment I 4
PROJECT ENGINEER-STRUCTURAL SYSTEMS: D. F. Guyot
!. EDUCATION:
l B.S., Architectural Engineering, University of Kansas, 1970 l
. Additional Education and Training Nuclear, Structures Seminar, Black & Veatch, 1973
- j. Earthquake and Extreme Loading for Nuclear Power Facilities,
. Case-Western Reserve University, 1973
[ Specialty Conference on Structural Design of Nuclear Plant Facilities, American Society of Civil Engineers, 1973 Nuclear. Equipment Design Seminar, Black & Veatch, 1974 Advanced STRUDL Seminar, Black & Veatch, 1974 Westinghouse PWR. Technology, 1974 I Vibration and Seismic Analysis Procedu.tes for Nuclear Power Plants, Black & Veatch, 1974 i Specialty Conference on Code Requirements for Nuclear Containments, American Society of Civil Engineers, 1974 QA Training, Black & Veatch, 1975 Specialty Conference on Structural Design of Nuclear Plant Facilities, American Society of Civil Engineers, 1975
- EXPERIENCE:
My present position is Project Engineer-Structural Systems, i on the BFS Project. I function as responsible Project Engineer f for civil, structural, and architectural activities associated with the central complex structures for BFS and, as such, I am responsible for the coordination, accuracy, and completeness of the detailed design, engineering inputs to specifications, and other work produced within the above disciplines. In addi-l tion, I participate in engineering and construction procedures J
l
preparation and scheduling activities, provide input to licens-ing support, and coordinate proje,ct activities for physical security provisions at the station.
The design activities relating to the topics set forth in my testimony are principally performed within the Civil-Structural Engineering Department. I or individuals under my direct super-vision have and will participate in the design activities related to each of these items including design calculations, construction drawings and construction specifications. I have provided input to and reviewed chapters in the Black Fox Station Preliminary Safety Analysis Report dealing with these l I
subjects and the material contained therein is true and corect !
to the best of my knowledge and belief. I have also prepared responses to or reviewed responses to Nuclear Regulatory 1 Commission Staff questions relating to these areas.
On a previous project, I coordinated the structural layout l and construction cost estimate work for a twin unit nuclear reference plant. I also served as structural design consultant and directed arrangement work and prepared building system estimates for a Nuclear Steam Supply System Bid Evaluation.
l
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD b #
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In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. , STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
Testimony of Lambert J. Sobon Concerning Contention 16 September 25, 1978 l
l
TESTIMONY 0F. LAMBERT J. SOBON CONCERNING CONTENTION 16 i
'My:name is Lambert J. Sobon. .I reside at 992 Redmond Avenue, San Jose, California. I am the Manager of BWR Containment Licensing, Containment Improvement Programs in the Nuclear Energy Business Group. A Statement of'my background and
' qualifications is attached as Attachment I to my testimony.
- My testimony deals with Contention 16 regarding the resolution of the following. phenomena and. associated loads which relate to the design of a Mark III Pressure Suppression Containment
- 1) Vent Clearing l 2) Vent / Coolant Interaction 3)- Pool Swell
- 4) Pool Stratification
- 5) Pressure Loads and Flow Bypass
- 1. Introduction f
My testimony presents technical information regarding the Mark III Containment System that may be used by a utility applicant in a Preliminary Safety Analysis Report (PSAR) design review. This information is set forth in
l loss-of-coolant accident (LOCA) .
- The containment system
)
employs the pressure-suppression concept, in which a i large pool of water (the suppression pool) is used to condense reactor steam which issues from a postulated ,
I reactor system pipe rupture. The suppression pool also acts as a reservoir for reactor energy under certain normal or anticipated operational conditions, such as i i
safety / relief valve operation (as would occur during cer-tain transients) and shutdown, i
The important pressure suppression features of the Mark III containment design are the drywell, suppression pool and containment upper pool. A schematic drawing of the Mark III reactor building which shows the location and orientation of the drywell, containment, suppression pool l
1
- LOCA is the sudden break of a high energy pipe in the reactor coolant pressure boundary of the nuclear steam supply system. The largest postulated break could be either the break of a main steam or a recirculation line.
This LOCA is the design basis accident (DBA). Other small line breaks result in LOCAs, and although their energy release do not result in large dynamic loadings, their thermal effects may control the design of structures.
The intermediate break accident (IBA) and small break accident (SBA) fall into this category. The size of the SBA is defined as that which will not cause automatic l
l depressurization of the reactor. The SBA is of concern l because it imposes the most severe temperature condition inside the drywell. The IBA is of concern because it is postulated to include the automatic actuation of the safety relief valves associated with the automatic depressurization system.
l The full-scale PSTF testing performed between November )
I 1973 and February 1974 obtained data for the confirmation of the analytical model. In March 1974 pool swell tests were performed in the PSTF. These full-scale tests involved air blowdown into the drywell and suppression i
pool to identify bounding' pool swell impact loads and breakthrough elevation, i.e., that elevation at which L the water ligament begins to break up and impact loads are significantly reduced. Impact load data was obtained on selected targets located above the pool.
In June of 1974, after the PSTF vent and pool system was i converted to 1/3-scale, four series of tests were performed to provide transient data on the interaction of pool swell q l
with flow restrictions above the suppression pool surface.
Other areas where data was obtained included vent clearing, ;
drywell pressurization, and jet forces on pool walls.
The next series of 1/3-scale testing began in January 1975 and was directed at obtaining local impact pressures and total loads for typical small structures located over the pressure suppression pool including I-beams, pipes, and grating. Data from this test series expanded the data base from the full-scale air tests. A further series of I 1/3-scale tests was added in June 1975 to obtain comparable data on pool swell velocity and breakthrough elevation to the full-scale air tests.
8-It should be noted that although most of the emphasis in i
the testing described above was directed at the evaluation of the pool swell phenomena, each test run consisted of a simulation of the postulated blowdown transient for various {
postulated break sizes up to two times the Design Basis Accident for the containment. Data was recorded at selected locations around the test facility suppression pool through-out the blowdown so that the hydrodynamic conditions asso-ciated with each phase of the blowdown is available for l i
selecting appropriate design loading conditions. General Electric has utilized this data to develop Appendix 3B. It provides numerical information for thermal and hydrodynamic loading conditions in the GE Mark III reference plant pres-sure suppression containment system during the postulated LOCA. Appendix 3B also presents information on thermal and hydrodynamic loading conditions during the anticipated safety relief valve (SRV) discharge and related dynamic events. This information is appropriate for PSAR evalu-ations. Separate test data has been utilized to estab- )
lish the SRV air clearing load prediction model presented in Appendix 3B to GESSAR as well as the SRV thermal per-formance. The GE reference plant report contains infor-mation and guidance to assist the containment designer in evaluating the design conditions for the various structures which form the containment system.
whose magnitude is dependent upon both location and the geometry of the structure. The pool ~ swell phenomenon can be considered as occurring in two
' phases, i.e., bulk pool swell followed by froth pool swell. The pool swell dynamic loading conditions on a particular structure in the containment annulus are dependent upon the type of pool swell that the I
In addition to locatio.n, the L structure experiences.
size of the structure is also important. Small pieces of equipment and structural items will only influence i
the flow of a limited amount of water in the immediate vicinity of the structure. Large platforms or floors, 1
l
! on the other hand, will completely stop the rising pool, l
l and thus incur larger loadings. For this reason such platforms and floors are located above the bulk pool swell zone, (e . g . , the-HCU floors).. This subject is discussed in Section 4.1.6.
t l
4.1.5.1 Impact L ads The PSTF air test data shows that after the pool has risen approximately 1.6 times vent submergence i.e.,
12 ft, the ligament thickness has decreased to 2 ft or less and the impact loads are then significantly reduced. Conservative bulk pool swell impact loading of 115 psi on beams and 60 psi for pipes are applied uniformly to any structures within 18 ft of the pool surface. For evaluating the time at which a
impact occurs at various elevations in the contain-ment annulus, the maximum observed water surface velocity of 40 ft/sec is assumed. Bulk pool swell would start 1 sec after the LOCA.
The basis for the loading specification is the PSTF air test impact data. These tests involved charging the reactor simulator with 1000 psia-air and blowing down through a 4.25 inch orifice. Fully instrumented targets located over the pool provided the impact data.
1 Additional tests have been conducted which provide impact data for typical structures that experience bulk pool swell. Data from these tests indicates that the specified design load is conservative.
It should be noted that impact loads are not specified for gratings. The width of the grating surfaces (typica.11.y 1/4 inch) do not sustain an impact load.
This has been verified in the 1/3-scale PSTF test.
Grating drag loads are calculated.
For structures above the 18 ft elevation, the conserv-ativa froth impingement load of 15 psi should be used.
Again, this impingement load is applied uniformly to all structures with the time. This is also based on data generated during the PSTF air test series.
For structures between 18 and 19 feet above the suppression pool design loads and duration are linearly extrapolated from the values of 115 or 60 psi to 15 psi.
The influence of seismic induced submergence variations on the pool swell transient and resulting impact loads has been considered. It has been concluded that the effect on the magnitude of the pool swell impact load is not significant. This conclusion is based on a consid ration of the influence of submergence on swell velocity and the significant load attenuation which will result from the pool surface distortions. The very significant margins between the specified loads and the expected loads provides confidence that any local increase in swell velocities will not result in loads in excess of design values.
4.1.5.2 Drag Loads In addition to the impact loads, structures that experi-ence bulk pool swell are also subject to drag loads as the pool water flows past them. Drag loads are calcu-lated assuming a velocity of 40 ft/sec. between the pool surface and HCU floors.
4.1.6 Loads on Expansive Structures at the HCU Floor Elevation At the HCU floor elevation there are portions of the floor which are comprised of beams and grating and other portions that are solid expansive structures.
approximately 10 lbm /ft3 The analytical model used I to simulate the HCU floor flow pressure differential has been compared with test data. These tests indicate .
the HCU floor pressure differential is more realistically l in the 3 to 5 psi range.
The potential for circumferential variations in the pressure transient in the wetwell region beneath the HCU floor have been examined and on the basis of bound-ing calculations it is concluded that the pressure variation will be less than 0.5 psid.
l 4.1.7 Loads on Small Structures at and Above the HCU Floor Elevation l
Small structures at the HCU floor elevation experience i
" froth" pool swell which involves both impingement and drag .ype forces. PSTF air tests show that the struc-tures experience a froth impingement load of 15 psi lasting for 100 milliseconds. Structures must be designed for this short term dynamic impingement load.
Grating structures are not subjected to this impinge-ment load as discussed in Section 4.1.5.1. Following the initial froth impingement there is a period of froth flow through the annulus restriction at this elevation with a pressure differential as discussed l
in the previous section.
i Those small structures above the HCU floor.that could be exposed =to pool swell froth are exposed to a drag load. The drag load is determined for the geometric shape of the structure using a froth density of 18.8 lbm/ft 'as in the HCU floor' differential pressure cal-f culation and the velocity of the froth at the elevation l
l of the structure. The velocity used is'50 ft/sec at )
4 19-1/2 f t above the suppression pool and is decelerated by the effects of gravity. The velocity of 50 ft/sec is a bound of the available data. Pool swell is not assumed for structures located more than 30 ft above the suppression pool.
t 4.2 Vent Clearing As the drywell pressure increases following a postulated LOCA, I'
the water initially standing in the vent system accelerates into the suppression pool and the vents are cleared of water.
The process of vent clearing affects the maximum pressure i
that will be reached within the drywell.
GE has examined vent clearing performance as a part of its confirmation of the analytical model for computing drywell pressure response for postulated LOCA events. This was done in one-third and full-scale tests. Predicted drywell pressure responses from these tests agreed well with observed data thus confirming the adequacy of vent clearing predictive methods. In addition vent clearing loads were obtained from the one-third and full-scale tests. These loads are specified l
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for GE's standard plant in a form directly applicable to l plant design and are identified in Appendix 3B.
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The Mark III test program with respect to the vent clearing i phenomenon is complete, and the program provides adequate l
data to assure that the Mark III containment vent clearing loads are properly defined.
l The following sections discuss the ve.c clearing loads identified in Appendix 3B.
4.2.1 Loads on Drywell (Drywell Pressure)
During the vent clearing process, the drywell reaches a peak calculated differential pressure of 21.8 psid.
During the sub' sequent vent flow phase of the blowdown, the peak pressure differential does not exceed 21.8 psid value even when it is assumed that pool swell results in some two-phase flow reaching the contain-mont annulus restriction at the HCU floor. Interaction between pool swell and the limited number of structures at or near the pool surface does not adversely affect the drywell pressure. The calculated drywell pressure during the Design Basis Accident includes the HCU floor pool swell interference effects. The containment response analytical model was used to calculate these values.
1 1
- 20.- i During the blowdown process, the drywell is subjected to differential pressures between levels because of 1
l flow restrictions. This value varies with the size ,
1 of the restriction, but a bounding value for a.25 per-cent restriction is 0.5 psi. On the basis of this calculation, it has been concluded that differential pressures within the drywell during the Design Basis Accident will be small and as such, need not be specifically. included in the drywell loading specifica- j tions.
4.2.2 Loads on Weir Wall l' The pressure drop at any point on the weir wall due to l
l the acceleration of water during vent clearing is less than the local hydrostatic pressure. Therefore, there is no net outward load on the weir wall due to vent l
clearing. This conclusion is based on the predictions l-of the containment response analytical model.
1 I
Once flow of air, steam and water droplets has been established in the vent system, there will be a static pressure reduction in the weir annulus that leads to approximately a 10 psi uniform outward pressure on the l
weir wall. This loading was calculated with the vent flow model and for design purposes is assumed to exist during the first 30 seconds of blowdown. i l '.
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i 4.2.3 Loads On Containment (Water Jet)
Examination of applicable PSTF data indicates some evidence of a loading of the containment wall due to the water jet associated with the vent clearing process (e.g., less than 1 psi), as indicated by a small spike at 0.8 sec. These water jet loads are negligible when compared to the subsequent air bubble pressure dis-cussed in Section 4.1.2 and are not specifically included as a containment design load.
4.3 Vent / Coolant Interaction (Vibratory Steam condensation) i Chugging i l
Following the vent clearing and pool swell transient asso-ciated with drywell air venting to the suppression pool there is a period of high steam flow through the vent system followed by reduced steam flow as the primary system high i
During this phase the I energy fluid inventory is depleted.
top row of vents are able to sustain the steam flow and the lower two rows are completely covered with water- As the steam flow through the vents decreases to very low values, the water in the top row of vents begins to oscillate back and forth. This action results in dynamic loads within the top vents and on the weir wall opposite the top vents.
Oscillatory pressure loadings can also occur on the drywell, suppression pool basemat, and containment. This low-steam-flow oscillatory process, named " vent / coolant interaction" by the Intervenors, is referred to as " chugging."
1 1
)
The chugging loads described above have been evaluated in 1/3-scale and full-scale experiments as part of the Mark III test program. For GE's standard plant the loads are specified in a form directly applicable to plant design in Appendix 3B.
Additional testing is ongoing which will provide more 1
data for evaluating steam condensation / chugging loads -- '
a very localized loading condition; however, the experiments referred to above provide a sufficient basis to select design loads for preliminary design purposes. l l
The following sections discuss the vent / interaction loads identified in Appendix 3B.
4.3.1 Loads on the Drywell 4.3.1.1 Condensation Loads Following Design Basis Accident (DBA)
Following the initial pool swell transient caused by the venting of drywell air to the containment free space, there is a period of 1 to 5 minutes (depending upon break size ard location) when the vents can experience high steam mass flow rates. Vent steady state steam 2
mass fluxes of up to 25 lbs/sec/ft occur as a result of either a main steam or recirculation line break. The PSTF facility has undergone single vent' steam blowdown
simulate both small and large breaks, and the blowdown of both saturated liquid and saturated vapor was tested. In parallel with the above, test data was obtained for use in understanding the loading conditions on submerged structures located within the suppression pool. Although the emphasis of this data acquisition was on the LOCA water jet and air bubble formation submerged structure loads, data on conden-sation/ chugging was also obtained. {
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The second phase of the remaining Mark III confirmatory testing program consists of a full-scale test series with the same basic objectives as the above described 1/3-scale ter,t series. This testing has also been completed and documented. The final test phase consists of a 1/9-scale test series in which a nine-vent array will be utilized to evaluate multivent effects. Installation of this vent con-figuration has been completed and testing is scheduled to be completed in 1979. Final documentation of the Mark III confirmatory test program results is scheduled to be completed in the first quarter of 1980. The results and interpretation of these tests have been and will continue to be transmitted to the NRC Staff and Applicants on a timely ba' sis.
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UNITED STATES OF AMERICA r, ,
8 NUCLEAR REGULATORY COMMISSION 'v ,
C BEFORE THE ATOMIC SAFETY AND LICENSING BOARD s
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l In ' the Matter of )
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FUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE,-INC., ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 l
COOPERATIVE, INC. )
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l (Black Fox Station, Units 1 and 2) )
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l Testimony of Mr. Robert E. Stippich Concerning Question 6-2 l
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l September 25, 1978 l
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TESTIMONY OF ROBERT E. STIPPICH CONCERNING QUESTION 6-2 l
My name is Robert E. Stippich. I reside at 4233 Mercier f l
Street, Kansas City, Missouri. I am a Consultant in the Systems Department at Black & Veatch Consulting Engineers in Kansas City, ,
1 Missouri; Architect / Engineering firm employed by Public Service Company of Oklahoma. The sections of the Black Fox Station PSAR on tornado protection were prepared under my supervision. A statement of my background and qualifications is attached as Attachment I to my testimony.
My testimony addresses the following question posed by the Licensing Board in its Order of September 8, 1978:
6-2. What connection, if any, is implied between the UHS cooling tower discharge nozzles and the off-gas systems' potential for radioactive release by the statement at pp. 1.9-22 and 1.9-23 of the PSAR?
There is no connection whatever between the off-gas systems' potential for radioactive release and the UHS cooling tower discharge nozzles. They are completely independent and unrelated systems.
In committing to Regulatory Guide 1.117 " Tornado Design Classification," Revision 0, June 1976, the applicant took exception to the need to protect the off-gas system from tornadoes on the ground that failure of the unprotected system could not 1
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result in offsite exposures that are a significant fraction of the guideline exposures of 10 CFR Part 100. This exception was based on a judgment that although some components of the off-gas ;
system are designated in the Appendix of the guide to require protection from tornadoes, the offsite radiological consequences of a gross failure of the system would be a small fraction of the exposure guidelines of 10 CFR Part 100 and would therefore make protection of the system unnecessary under Paragraph B of the guide. Paragraph B defines 3 classes of systems which should be protected from tornadoes, none of which include the off-gas system.
This judgment was subsequently vindicated when the NRC Staff issued Revision 1 to Regulatory Guide 1.117 in April 1978. The revised guide requires protection only for "(t] hose portions of the gaseous radwaste treatment system whose failure due to tornado effects could result in potential offsite exposures in excess of the criterion given in subitem (3) of the regulatory position." Subitem (3) states: "Those [ systems] whose failure could lead to radioactive releases resulting in calculated offsite exposures greater than 25% of the guideline exposures of 10 CFR Part 100 using appropriately conservative analytic methods and assumptions." Because the offsite consequences of a gross j failure of the off-gas system for the Black Fox Station would be a small fraction of 25% of 10 CFR Part 100 guidelines, the system need not be protected from the effects of tornadoes. A failure analysis of the off-gas system is presented in BFS-PSAR Section 15.1.36 " Main Condenser Gas Treatment System Failure."
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ATTACHMENT I SYSTEMS ENGINEER: Robert E. Stippich EDUCATION: BS, Civil Engineering, 1949 Washington University EXPERIENCE:
I joined Black & Veatch, Consulting Engineers in 1953, and I am presently a Consultant in the Systems Engineering Department. I was_Section Leader of the Systems Engineering Department Speciality Services Section with responsibilities for licensing and design tasks in the areas of seismic and pipe
' rupture design, turbine missile analysis, and coordination review of all regulations, reculatory guides, and industry codes and standards. I was formerly a systems engineer with responsibility for conducting systems analyses and for performing engineering design of the civil and structural engineering. features of large nuclear and fossil fuel electric power. plants.
I have participated in several nuclear power plant design and feasibility study projects. On these projects, I was responsible for structural layouts, design, functional analyses and cost estimates for containment structures including pressure suppression containments for boiling water reactors and dry containments for pressurized water reactors. In addition, I was responsible for the evaluation of the geologic, seismologic, hydrologic, and meteorologic site factors related to the contain-ment design. I also participated in studies of the technical and economic feasibility.of the Molten Salt Reactor, developing l
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l plant arrangements and conceptual designs for containment and structures. I supervised preparation of detailed design, plans and specifications for structures, foundations and other civil engineering features of large steam-electric power plants.
In all, I have participated in the design of more than 50 power plants.
I am presently serving on the Task Committee on Turbine Foundation of the ASCE Power Division.
Prior to joining Black & Veatch, I was employed by Boeing Airplane Company, Seattle, Washington where I served as a senior facilities engineer working in design and in equipment vibration control, structural dynamics and noise control. I 1 I
served as an officer in the United States Navy. I l
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1' ECL\TED connespayug.g f #/ ..o "d' %Y ,
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UNITED STATES OF AMERICA . 3 , .>.~
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l NUCLEAR REGULATORY COMMISSION k ,f/
l 9 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
l In the Matter of )
l )
l PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC. ) STN 50-557 l AND WESTERN FARMERS ELECTRIC )
COOPERATIVE,.INC. )
)
(Black Fox Station, Units 1 and 2) )
Testimony of Messrs William G. Gang and Richard B. Johnson Concerning Contentions 7 and 8 (Fire Protection)
September 25, 1978 l
TESTIMONY OF WILLIAM G. GANG AND RICHARD B. JOHNSON Concerning Contentions 7 and 8 (Fire Protection)
My name is William G. Gang and I reside at 6428 Paso Los Ceritos, San Jose, California. I am the Project Manager for supply of nuclear steam supply components for the Black Fox Station working within the Nuclear Energy Projects Division i
of General Electric Company. A statement of my background j l
and qualifications is attached to my testimony on Contention 3. j j
My name is Richard B. Johnson and my business address is General Electric Company, 175 Curtner Avenue, San Jose, California.
I am employed by the General Electric Company as a Senior Licensing Engineer in the Nuclear Energy Projects Division. A Statement of my background and qualifications is attached as Attachment I to this testimony.
This testimony concerns Intervenors' Contentions No. 7 and 8b/ concerning and the compliance of 'ne Black Fox Station l
cabling with the flame retardancy and separation requirements of 10 CFR Part 50, Appendix A, Criterion 3.
1/ 7. .Intervenors content that in order for the Applicant to meet 10 CFR Part 50, Appendix A, Criterion 3, Black Fox, 1 and 2 must utilize cables with fire retardant insulation.
- 8. Intervenors contend that in order to meet 10 CFR Part 50, Appendix A, Criterion 3, the Applicant must separate the cable trays, including those in the cable spreading rooms so as to prevent a recurrence at Black Fox, 1 and 2 of the type of fire which took place in the cable spreading room at Browns Ferry.
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Black Fox Station cabling in the GE scope of supply will meet the requirements of 10 CFR Part 50, Appendix A, Criterion
- 3. Information supporting this position as it relates to the reactor manufacturer, General Electric Company, is supplied herein. GE's scope of supply, for purposes of these conten-tions, includes the Power Generation Control Complex (PGCC),
which consists of three major sets of assemb?.ies: 1) termina-l tion cabinets, 2) steel floor sections, and 3) interpanel cables. GE also supplies the control panels in the control room and certain local panels throughout the plant. Testimony concerning the cabling within the scope of supply furnished by the architect engineer, Black & Veatch Consulting Engineers, is set forth in Mr. Engmann's testimony. Black & Veatch's scope of supply, for purposes of these contentions and in relation to that of GE, originates at the termination points within the termination cabinets of the PGCC design and at other termination points throughout the plant, extending to the remainder of the plant. This also includes the interface with the fire protection features (detection and suppression) pro-vided in the GE scope of supply.
Criterion 3 to Appendix A of 10 CFR Part 50 requires that structures, systems and components important to safety shall l
( be designed and located to minimize, consistent with other safety requirements, the probability and effects of fire and
explosions. With respect to the PGCC for the. Black Fox Station, Tefzel. cable insulation is used on substantially all cables.2/
Tefzel, a Dupont Trade name, is a high temperature, flame retardant fluoropolymer material which has passed the IEEE 383 l (1974) flame test and is required by GE specification.3,/ This j test is appropriate for electrical cable construction included I in the PGCC.A!
Where it ray be desirable to use cabling material other than Tefzel, other materials which meet IEEE 383 (1974) will be used. In addition to GE's use of flame retardant insula-l tion, any cables added to the PGCC by the applicant /AE must be Tefzel or its equivalent from the standpoint of flame retardant i
properties. Tests conducted by the Structural Research 1 Laboratorv of the University of California at Berkeley, and witnessed and approved by NRC' representatives, showed that the I PGCC floor sections provided adequate fire protection features to meet the requirements of General Design Criterion 3. In addition to the use of flame retardant insulation, the PGCC design contains other features to minimize the probability of
~2/
See NEDO-10466, Rev. 2, " Power Generation Control Complex-Design Criteria and Safety Evaluation," page 4-11.
3,/ Ibid., Section 5.1.8, page'5-2.
4/ Regulatory Guide 1.120, Revision 1, " Fire Protection Guidelines for Nuclear Power Plants," dated November 1977,
- p. 120-13.
I
fires and protect plant personnel. Industry codes and standards provide more than adequate safety margins on the selection of wire for electrical loads, precluding the possibility of ignition from an electrical overload. Additionally, PGCC floor sections are designed to limit the flow of air and exhaust gases that may be generated from a fire by the enclosed nature of the floor sections and by sealing of all penetrations with semi-permanent fire stops. Thus, should a fire start it would have only a limited amount of oxygen available and hence would be suffocated. This was demonstrated by the tests conducted at Berkeley which are referenced above. Also, the PGCC for Black Fox Station, will have fire suppression and detection systems installed. These systems of the PGCC are covered in detail in NEDO 10466, Rev. 2. The use of fully-enclosed electrical raceways in underfloor spaces which characterizes the PGCC floor sections is consistent with the requirements of Regulatory Guide 1.120, Rev. 1.
With respect to the requirement for cable separation, the tests referenced above also showed that the cable separa-tion within the PGCC floor sections satisfies IEEE 384-1974 relating to separation of Class lE equipment. IEEE 384 specifies that under certain circumstances, which are appli-cable to the PGCC floor sections, separation criteria (i.e.,
minimum separation distance and/or adequate separation barriers) can be established by analysis and testing of the proposed cable i
installation. As Regulatory Guide 1.75 indicates, the IEEE 384 !
cable separation test has been accepted as satisfying 10 CFR Part-50, Appendix A, Criterion 13. The referenced tests and analysis established that-the cable installation in the PGCC floor section' provide adequate separation. This installation will be used at Black Fox Station. The Black Fox PGCC cable design complies with the requirements of Regulatory Guide 1.75.
In its " Order Ruling on Motions for Summary Disposition,"
the Licensing Board appears to have concluded that IEEE 383 (1974) and IEEE 384 (1974) standards have been invalidated by the occurrence of the Browns Ferry fire. This is indeed not the case. IEEE 383 (1974) and IEEE 384 (1974) remain the applicable standards required by the NRC. Browns Ferry was designed prior to the adoption of these 1974 standards, and was not necessarily designed to meet the equivalent require-ments. The entire. area of fire protection for nuclear power reactors has been reviewed as a result of the Browns Ferry l fire experience (see the report by Hanauer entitled NUREG-0050,
" Recommendations Related to the Browns Ferry Fire," published in February, 1976), and these 1974 standards remain the'appli- j 1
cable standards for present plant design.
l
[
The guidelines of the Hanauer report have been included in Regulatory Guide 1.120, Rev. 1, and PGCC for Black Fox Station meets this Regulatory Guide. Gage-Babcock, mentioned by the Licensing Board in its Order at p. 26 as a new source
-1
l of information, participated in the review of NEDO 10466, Rev.' 2, as NRC consultants. -They concurred with the NRC evaluation and conclusions as stated in the NRC approval letter for this report (Olan D. Parr, NRC, to Dr. G. G.
i.
I Sherwood, GE, dated July 13, 1978) -(copy attached as Attach-ment II). ;
I I
! l l It has also been suggested that the use of raceways under i
j the false floor for the TGCC, as contemplated by GE in its i
L design, conflicts with that portion of item J (p. 29) of l
l BTP 9.5-1 (dated May 1, 1976) which provides:
" Cables should not be installed in floor.
trenches or culverts in the control room."
l This assertion is incorrect. Cables for the PGCC for the
. Black Fox Station will be installed in cable trays, racewaye
. _ and conduit as distinguished from trenches or culverts. Race-ways, cable trays and conduits are not considered to be a synonomous term with trenches or culverts. Trenches or culverts are, in GE's opinion, recesses in a floor which~could collect I
water.during the course of a fire suppression activity using water. This water could cause electrical short circuits in cables.in the trenches or culverts. This result should be avoided and raceways, cable trays, and conduit, proposed by i
GE to be installed under the false floor of the PGCC will not collect water. In any event, the May 1976, issue of Branch I '-
Technical Position (BTP) 9.5-1, which is the source of inter-venors' question has been superseded by BTP 9.5-1, Revision 1, issued in Spring of 1978. This revision does not contain the ]
provision of item (j) mentioned above.5/
A question has also been raised as to whether the PGCC .
for the Black Fox Station meets a " requirement" in Reg. Guide 1.120, Rev. 1 for an automatically initiated area fire suppres- l l
i i
sion system in the event underfloor cables are used in fully enclosed raceways which are not protected by an automatic suppression system. At the time Reg. Guide 1.120 was proposed, state-of-the-art technology for cable insulation used in nuclear power plants did not account for the excellent flame retardant l
characteristics of Tefzel insulated cabling. Accordingly, a requirement for an automatic suppression system as stated above was included in Reg. Guide 1.120 as initially proposed. How-ever, as a result of favorable test experience on Tefzel insulated cable, GE r.evised its firo protection guidelines 1
as provided to the NRC Staff in March 1978 to state that:
"For Tefzel insulated cable-cased plants, the fire suppression system shall be manually initiated. For non-Tefzel based plants, the fire system shall be automatically initiated."
(NED0-10466 Rev. 2, Erratta and Addenda Sheet, dated June, 1978, Section 5.1.9, P. 5-2).
t
-5/ It should be noted that Revision 1 to BTP 9.5-1 is the same in content as Regulatory Guide 1.120, Rev. 1.
I
In apparent recognition of the fire retardant qualities of Tefzel, the'NRC Staff approved this design requirement as a part of.its approval. .In Attachment II the NRC states that:
" Based on our review of the design basis and criteria for. fire protection, interface infor-mation, and the fire tests, we conclude that the proposed design is in accordance with Regulatory Guide 1.120 and General Design Criterion 3, and is, therefore, acceptable."
For the above stated reasons, the Black Fox cabling supplied in( GE meets Criterion 3 of Appendix A to 10 CFR, Part 50 and the NRC Staff concurrence to that effect is set forth in the attached letter.
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Attachment I i
RICHARD B. JOHNSON l EDUCATION:
BS, Engineering, University of South Florida, 1970 EXPERIENCE:
I am currently the licensing engineering responsible for the generic licensability of the PGCC design (including the fire protection features) and the fire protection features of the GE standard plant design. I have held the position of project licensing engineer since 1973 and in this capacity I have the responsibility for all GE licensing activities associated with the Hartsville Nuclear Plant, the Phipps Bend Nuclear Plant, the Zimmer Power Station and various generic licensing issues such as PGCC, GE Standard Reactor Island Design (STRIDE), High Pressure Core Spray (HPCS) System Diesel Generator Power Supply and BWR/5 Recirculation Flow Control.
My general duties include the coordination and prepara-tion of technical information needed to demonstrate the design adequacy, or justification of the generic GE position for the issues under my responsibility, and the necessary licensing support to reach resolution and NRC approval. My project duties include the coordination and preparation of answers to questions regarding the normal Nuclear Steam Supply System (NSSS) and the extended scope of the Standard Reactor Island
(
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! Design (STRIDE).
l l
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Following graduation from the University of South Florida, l I was employed by Florida Power Corporation where I served 1
successively as a Design Engineer in the Generation Engineer-ing Department and as a Quality Surveillance Engineer in the Generation Quality and Standards Department. In these posi-tions, I was involved with the design and construction of the Crystal River Unit No. 3 Nuclear Plant. As a Quality Surveil-l lance Engineer, I was a site corporate representative for assuring adequate QA and QC operations during construction.
I joined General Electric Company in 1973.
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Attachment II
\ ... [. /
j Central ifles - Topical Reports A L I 3 i373 General Electric Ccmpany '
ATif4: Dr. G. G. Sherwcod, Manager
! Safety and Licensing -
175 Curtner Avenue
{
San Jose, California 95114 Gentlemen: ,
SUBJECT:
REVIEW 0F GE';Er<AL ELECTRIC TOPICAL REPORT NE00-10466, i ~l '
" POWER Gl?1EPAT IO' C0'liPOL COPPt.EX DESIG'4 CRITERI A A*10 SAFE T Y E'.' ALUAT 10'1" We have completed our review of Revision 2 of the subject topical
! report. Based on this review, we conclude that '.he report is accnptable for reference in license applications as specified in
$ the enclosed evaluatica. ;
1 The staff does not intend to repeat its review of this topical report when it appears as a reference in specific license appli".ations.
l ercept to assure that the topical report is applicable to the spacific
- piants involved.
Shoisld r equlatory criteria or regulations chanqe such that our conclu*. ions concerning this topical report are invalidated., you '. vill
, t:e notified and will be given the opportunity to revise and resubmit ,
yr tr topical report for review, should yois so desiro. f I
l
' In accordince with established procedura, we requ"st that Gnneral l
Electric issue a revised version of the subject tepical report to {
include any supplementary information provided for our review of the j report, this acceptance letter, and the enclosed staff evaluation.
l Sincerely.
Ul'o(n Y ar'r.Ff ef Light Water Peactors Brae- No. 3 Division of Project Manage ent
Enclosure:
Topical report Evaluation cr w/ercim.ure:
"r. L. "'*ferd O, a1 Ioctric Company
'7?^ " m ery Lane ct- .i, "aryland 20014
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.EtlCt.,0SURE TOPICAL. REPORT EVALUATION I
Report No.: NE00-10466, Revision 2 Report
Title:
Power Generation Control Complex Design !
Criteria and Safety Evaluation I
. Report Date: March in/8 .
Originating Organization: General Electric Co'Tany Review By: Auxiliary Systems Branch Instrumentation and Control Systems Branch SUtMARY OF TOPICAL. REPORT The Power Generation Control Cortplex (PGCC) is a prefabrication concept I for routing and installing cabling for the control room of boiling (
witer reactors. fGCC consists of three major asseMilins: (1) a set j of steel floor sections. (?) a set of terrtination cabinnts, one for .
j each floor section, and (3) a set of interpiriel cables. The <,tcel 4 floor sect ions contain raceviays to provide routing and separation of all I intcrpanol cabting. The operator panels anj the sigual Conditioning panels will be riounted on the floor sections. These Pinels are not j considered a part of PGCC and will be revicaed on a plant-by-plant J t'a s i s . The termination cabinets provide the interf ace between the field installed cabling and the factory installed cables in the floor sections. ;
3 T'e fire protection system for the PGCC will consist of fire detection arl surpression for the floor sections and fire detection for the t eninatton cabinets.
S':""ARY Of RE Gift ATORY EVAL.UATION The staf f's review of the original revision of the topical report was not co pleted pending the develop"ent of regulatory guidance with res;:ect to scraration of electrical systems. In February 1974, the staff issued j Requiat9ey ;9ide 1. 75, " Physical Independence of Electric Systo-s."
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Subsequently GE revised the topical report to address this renulatory guido.
- Ona area of concern was the physical separation of the cabling of the wire ways in the floor sections. In sorre cases, inderendent Class IE , j circuits or Class IE and ::en Class IE circuits, are sorarated by only a j l
metal bartier within the flonr section. Ihm ofore it wa ; necessary for j j Gi to dennnstrate that there was sufficient ind-pendent" betwenn thesc . j circuits. This was done by a series of fire tests and analyses. (
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i 75a flRC conwi tant s . 'iage-Baboc k and Assoc ia tes . Inc. o f Chicago !
f ra ticipatrJ to the review of the PGCC Topical Report and of the fire I barrier test. <e cribed below). This evaluation includes the recomen-l dations of the ecnsultants. Gage-Babock and Associates, Inc. concur E with our evaluation and conclusions.
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q General Electric has perfor-ed fire barrier tests for the floor sections of PGCC at the University of California laboratories in order to demon-strate that a fire in one cable safet'/. division v:ould not propagate or cause unacceptable damage to a redundant cable safety division in an l adiacent racaway. The test fire and acceptance criteria were aqrced '
The tests were perforled to 1 to Dy GC and the staf f before the tests. I deionstrate that PGCC had acceptabla fire barriers between cable divisions.
A descript ion of the fire tests an<t their resnits are contained in the PGrC f oric al 1:rport. Conservative fire tests weie conducted by placing co- bust ible n'aterial within one cat ie division e ,iceway and removing ,
finor plates from each end of a floor section to provido ventilation.
These te'.ts were perfonred with no fire sa;iprossion throuqhout the i duration of the tests. The fire test resnits shosed that (a) non-Tefzel l cable resulteJ in a fire which propagated partially in one raceway and l produced rnnsiderable smoke; and (b) Tefiel insulated cable resulted in little or rio fire propagation in the raceway and considerably less smoke g"neration thiouqho*tt the teg .
At our request. General Electric has included provisions for a gas fire suppressien system (Italon 1301) in the floor sections for all PGCC confiriurations. Those system will be automatic for nor-Te(rel insultated cable based plants. The Halon supply systen external to the floor sections will be the responsibility of applicant referencing pGCC. GE GE bas provided has also l a'.ceptable intorf aces for the design of the Halon systen. .
I pe uvided quick opening covers on the floor sections for rianual fire !
fightinq purposes.
I Fire detection for the floor sections will consist of two theiwal and one Thesc ;
product of combustion detector in each longitudinal raceway. ,
J detectors will be connected in a zoned configuration te indicate location .
- of a fire by floor section. Applicants referencing MCC will be responsible I
.for the zoned system external to the PGCC.
For fire protection evaluation purposes, the terminatten c.abinets for the j j PGCC design are similar to' cabinets in other plant designs and were in not these ;
tested. We required that GE provide a neans for detecting fires GE Drnvided l te'9initir>n cabinets which contain re fundant safety divisions. '
l product of cm bustion detectors at the top of each cabinet whichEach contains catie i !
refundant saf"tv cable divisions to detect a fire in any bay. :
divisien will also be separated by a retal ' Ire barrier within tha l i
terminat ion cabinets, the derreters will alarm locally for t hm.o plants ,
- j whrre the cabinots are located in the control room, if the cabinets are j i Intated outs Me the control room, then the det.ectors will be connected {j tn the iaain tontiol room fire protection panel. I 8 j The termiriattori cabiriets will also be readily accessible for n'anual fire i -
fighting by nicans of hinged doors on the front of each cabinet that will ' i be kept unlocked for quick access. >
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REGUt.ATORY P_0_SIJ_IO#
All panels, consoles, and cabinets other than the termination cabinets will be evaluated for fire protection on a plant-by-plant basis since they are not included in PGCC Topical Report.
- Based on the overall review of the PGCC we conclude that the PGCC '
design provides sufficient independence between the Class IE circuits '
and between the Class IE and None IE circuits by utilizing barriers and physical separation.
I Therefore, the PGCC design is acceptable with respect to the physical independenca of electric systems and Raquiatory Guide 1.75. f i
- Bas id on our review of the design basis and criteria for fire protection. (
l't .rtace infor tation, and the fire tests a conclude that the proposed design is in accordance with Regulatory f u'".? 1.120 and General Design Criterion 3 and is, therefore, acceptabt; ;
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
PUBLIC. SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557
, AND WESTERN FARMERS ELECTRIC )
l COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
Testimony of Gary R. Engmann Concerning Contentions 7 and 8 (Fire Protection)
September 25, 1978
TESTIMONY OF GARY R. ENGMANN CONCERNING CONTENTIONS 7 and 8 (FIRE PROTECTION)*
My name is Gary R. Engmann. I reside at 11409 W. 90 Terrace, Lenexa, Kansas. I am Project Engineer-Electrical Systems for the Black Fox Station design project at Black &
Veatch, Consulting Engineers in Kansas City, Missouri, engineering firm retained by Public Service Company of Oklahoma. A statement of my background and qualification is attached as Attachment I to my testimony. My responsibilities as Project Engineer-Electrical Systems included preparation of Chapter 8 of the PSAR and preparation of the responses to the NRC Questions concerning the information presented in Chapter 8. My responsibilities also included providing technical advice pertaining to the description and analysis of electrical systems, equipment, and components f included in the preparation of the PSAR and PSAR Reference Reports. j Contentions 7 and 8 read:
- 7. Intervenors contend that in order for the Applicant to meet 10 CFR Part 50, Appendix A, Criterion 3, Black Fox, 1 and 2 must utilize cables with fire retardant insulation.
- 8. Intervenors contend that in order to meet 10 CFR Part 50, Appendix A, Criterion 3, the Applicant must separate the cable trays, including those in the cable spreading rooms so as to prevent a recurrence at i Black Fox, 1 and 2 of the type of fire which took place in the cable spreading room at Browns Ferry.
My testimony addresses 10 CFR Part 50, Appendix A, Criterion 3 -- Fire Protection and its applicability to the electrical cable constructions that will be used at BFS and cable qualification tests that provide adequate assurance that cable will meet the requirements of Criterion 3. My testimony also addresses 10 CFR Part 50, Appendix A, Criterion 3 -- Fire Protection and its applicability to the design of the cable l
tray installation in the BFS cable spreading rooms and the cable tray installation design criteria that provide adequate assurance that the design will meet the requirements of Criterion 3. My l testimony addresses cable and cable tray installation design that 1
is within the Black & Veatch scope of design. Messrs. Gang and l
Johnson of the General Electric Company address the Contentions in the context of GE's scope of supply, namely, the Power Genera-tion Control Complex.
Criterion 3 states in part:
" Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirement, the probability and effect of fires and explosions."
BFS structures, systems, and components important to safety are identified in BFS PSAR Section 3.2, and include cable and cable trays designated Class lE, as defined in IEEE Standard 380 - 1975,
" Definition of Terms used in IEEE Standards on Nuclear Power Generating Station." BFS cable designated as Class lE must be insulated with material which meets the requirement of Criterion 3.
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The cable spreading rooms at BFS are labeled " Division 1 and 4 Cable Spreading Area" and " Division 2 and 3 Cable Spreading Area,"
and are located at Elevation 575'0" in the Control Building (BFS PSAR Figure 1.2-19). The FBS Cable Spreading Areas contain Class lE cable and cable trays; and, therefore, the design of the cable tray installation in the BFS Cable. Spreading Areas must also meet the requirements of Criterion 3.
Flame retardant properties of Class lE cables can be adequately demonstrated by type testing in accordance with IEEE Standard 383-1974, "IEEE Standard for Type Test of Class lE Electric Cables, Field Splices, and Connections for Nuclear Power Genera-l ting Stations." Cables which meet this test have been accepted as j I
meeting NRC Branch Technical Position APCSB 9.5-1, Appendix A dated May 1976. All BFS cables within the Black & Veatch scope of design, 4 i
including any cables added to the PGCC, will be specified to meet IEEE 383-1974.
IEEE 383-1974 includes procedures for a flame test of an entire cable construction, including the cable insulation. This test demonstrates that, when the cable is installed in the vertical l
! l' tray configuration, the cable does not propagate fire even if the cable insulation and outer covering have been destroyed by fire in the area of flame source impingement. Successful completion of the type test of IEEE 383-1974 provides adequate assurance that the insulation and jacketing material of cable installed in a tray will not support the spread of a flame within that tray. It therefore provides adequate assurance that the cable constructions used for Class lE cables are designed to minimize the effects of fire as required by Criterion 3.
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Criteria for_the design of Class lE cable tray installa-tions are given in IEEE Standard 384-1974, "IEEE Trial-Use Standard Criteria for Separation of Class lE Equipment and
-Circuits." Section 5.1.3 of IEEE Standard 384 states criteria for the separation of Class lE cable trays in Cable Spreading Areas. That Section states in part:
-i "The minimum separation distance between redundant Class lE cable '
trays shall be. . .1 ft. between trays separated horizontally and '
3 ft. between trays separated vertically."
Section 5.1.1.3 of IEEE Standard 384-1974 states specific requirements for cable and cable tray construction, cable instal-lation, and hazard limits that must be met in order that the l I
l 1 ft. horizontal and 3 feet vertical separation distances can l
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be used as criteria for the design of the cable tray installation l
in Cable Spreading Areas. The BFS Cable Spreading Areas will meet these criteria.
Since the issuance of IEEE 383-1974 several cable fire l experiments have been conducted under the authorization of the NRC. The experiments have provided data for evaluation of the IEEE 383-1974 flame test and its adequacy in providing assurance that cable constructions that pass the test will meet the design g objective of minimizing the spread of a cable fire within a cable tray and thereby minimize the effects of fire. In an experiment conducted at Sandia Laboratories, cable of a construction that had passed the IEEE 383-1974 flame test was
installed in a horizontal run of cable tray, to a fill that appears to exceed the BFS cable-tray-fill design criteria, and ignited using two IEEE 383-1974 flame-test burners. From the reported test results, it appears that flame did not spread significantly beyond the area of burner flame impingement within the cable tray in which the ignited cable was installed.
In a series of six experiments conducted by Underwriters Laboratories, PVC insulated,' nylon jacketed cable was installed
.in a vertical cable tray and subject to a burner flame similar to that of IEEE 383-1974. (This cable construction probably meets IEEE 383-1974, but does not have as good flame retardant properties as the cable materials planned for BFS) . In five of the six experiments, separate burners were producing flame
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impinging on the front and rear surfaces of the packed cable mass for 20 or more minutes before flames had spread significantly beyond the immediate area of flame impingement. In two of the experiments, fire did not propagate beyond the immediate area of flame impingement, and the test was terminated 40 minutes after burner ignition. The results of these experiments are mixed.
However, these experiments do provide some data that support the adequacy of the IEEE 383-1974 20-minute flame test. That standard is a repeatable flame test that provides assurance that cable installed in nuclear power plants will meet the design objective of minimizing cable fire propagation, and therefore minimize the effects of fire. To my knowledge, the IEEE has no plans to revise the flame test provisions of this standard. In my opinion,
1 successful completion of the flame test of IEEE 383-1974 provides the required assurance that a cable construction is designed to minimize the effects of fire.
Since the issuance of IEEE 384-1974, several experiments have been conducted by Sandia Laboratories under authorization by the NRC. These experiments have provided data for evaluation 1
of the tray separation distance criteria of IEEE 384-1974, and the adequacy of these criteria in providing assurance that the cable tray installation will meet the design objective of minimizing the effects of an electrically initiated fire in one cable tray or adjacent cable trays.
These experiments were conducted on cable tray installa- l tions that met the requirements of IEEE Standard 384-1974 for cable and cable tray construction, cable installation, and hazard limits. The cable tray separation distances used in these experiments were less than the IEEE Standard 384-1974 design criteria for minimum separation distances between cable trays in Cable Spreading Areas. Seven separate full scale experiments were conducted, and the experiments are described in detail in Sandia Laboratories reports. In all of these experiments, electrical ignition of a cable fire was accomplished in an
" ignition" tray, and data was gathered on the effect of such 1
l a fire on cable in trays below, above, and to the side of the i
ignition tray. As stated on page 32 of the Summary and Conclusions of " Cable Tray Fire Tests," Sand 77-ll25C, July 1977:
"At no time did the cables in trays displaced from the ignition tray begin i to burn. All circuits in these trays remained functional and elongation measurements taken of the insulation closest to the fire showed no major
( < 10 % ) change."
These experiments provide significant assurance that Cable Spreading Area cable tray installations designed in accordance with the criteria of IEEE Standard 384-1974 will minimize the effects of fire as required by Criterion 3.
An additional experiment was conducted at+Sandia Laboratories in July 1977. In this experiment, a cable tray fire was I initiated by two burners, each of which satisfied the heat input l
specifications of IEEE 383-1974. In other words, the experiment ,
used an exposure fire. It also employed stacked' cable trays and cable tray separation criteria of IEEE 384-1974. Therefore, this experiment does not. constitute a verification of the cable tray separation criteria of IEEE 384-1974 for cable spreading rooms, since the design basis fire for cable spreading rooms is an electrically induced fire rather than an exposure fire.
Nevertheless, this experiment provided data useful in analyzing ;
ca'ble tray installations that have an exposure fire as a design basis.
The first attempt to ignite the cable with a five minute exposure to the dual burners did not result in a fire in the exposed cable tray. In order to obtain a fire, a steel bar was used to move the cables about at the ignition area to increase
8-air flow through the trays. Such an air flow is ordinarily l precluded by the industry standard practice of specifying installation of cable in a parallel, stacked fashion, with no j i
slack. This practice can be an effective means of reducing the probability of cable fire as a result of an exposure fire, and probably ought to be more rigorously enforced.
At least one aspect of the experiment tends to compromise I the potential value of the data. The cable trays simulating a redundant division spearated by five feet vertically could not be observed throughout the experiment. Since these cable trays collapsed into the lower trays, it is not possible to positively i 1
conclude that the top trays were set aflame by the fire in the j lower trays.
i On the other hand, the lowest three cable trays separated I horizontally from the exposed cable tray by 8 inches, were not i
affected by the fire and the cable in those trays remained j 1
functional. In addition the cable tray simulating the redundant division, and located three feet horizontally from the top of i
l the 14 trays simulating the other division, was not damaged and j l
l the cable in that tray remained functional. This provides some 1
evidence that the horizontal spearation distance criteria of IEEE 384-1974 may be more than adequate even if an exposure fire is admitted as a design basis for cable spreading areas. l l
With the issuance of IEEE 384-1974, the IEEE has established l l
cable tray specific distance criteria for cable tray in cable l spreading areas.
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Subsequent Sandia Laboratories experiments with electrically initiated fire have provided evidence that these specific separation distance :riteria are adequate. Although the IEEE issued a revision to IEEE 384-1974 in 1977, the cabl'e tray separation distance criteria were not revised; and, to my
< knowledge, the IEEE has no plans to revise these separation criteria.
In my opinion, cable: tray installations in cable spreading rooms designed in accordance with IEEE 384-1974 will provide the required assurance that the cable tray installation will minimize the effects of fire.
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ATTACHMENT I i PROJECT DESIGN ENGINEER-ELECTRICAL: Gary R. Engmann EDUCATION:
BS Electrical Engineering, University of Kansas, 1966 MS Electrical Engineering, University of Kansas, 1970 MBA Business Administration, University of Kansas, 1971 Additional Education and Training Westinghouse, PWR Technology Seminar, 1974 I-T-E Imperial Corporation, Electrical Switchgear for Nuclear Power Stations, 1975 Reliance Electric, Qualification of Motors for Class lE Service, 1976 Rotork, Inc., Valve Motor Operator Design and Application, 1976 United Technologies-Essex Cable, Fabrication and Testing of l Power Cable, 1974 I PROFESSIONAL REGISTRATION:
1 Professional Engineer, Oklahoma, P.E. 9995, 1975 Professional Engineer, Kansas, License No. 6750, 1973
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EXPERIENCE:
As Project Design Engineer-Electrical, I am responsible for the design of the auxiliary electric system including the preparation of economic analyses, engineering calculations, drawings, j and electrical equipment procurement specification for BFS. I am also responsible for the preparation of electrical construction l l
technical specifications and drawings.
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Since my assignment to BFS in 1974, I have supervised and contributed to the auxiliary electric system design and analysis, participated in the preparation of the PSAR chapter on the auxiliary electric system, supervised the detailed design of electrical raceway and cabling, contributed to the scheduling of electrical design activities, and coordinated the electrical design effort with other engineering disciplines required for station design.
Throughout 1974, I served as the Project Design Engineer-Electrical for a two-unit 960 MW coal-fired generating scation and participated in the preparation of studies, analyses, and !
l design criteria for electrical system configuration and equipment !
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ratings. I also participated in and supervised the preliminary i
design of the auxiliary electric system.
From 1972 to 1976, I served as the Project Design Engineer-Electrical for a two-unit 960 MW fossil-fueled generating station.
I participated in and supervised the final design of the auxiliary electric system, preparation of engineering calculations, prepara- ;
i tion of equipment and construction specification and drawings, and l I preparation of the electrical control schematics.
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Upon joining Black & Veatch in 1971, I was assigned to a l
design project for a 480 MW fossil-fueled generating station and !
I performed engineering calculations for auxiliary electric system j i
design. l I am presently serving on the Power Generation Committee and the Nuclear Power Subcommittee of the IEEE Power Engineering !
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Prior to joining Black & Veatch, I was employed by the Center for Research in Engineering Science, a'nd participated in the design, fabrication and testing of radar units for remote sensing application.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC 3AFETY AND LICENSING BOARD I
In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556
)
STN 5J-557 AND WESTERN FARMERS ELECTRIC COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
[?fTE.D CORRESPONDDiCE Testimony of Dr. E. L. Cox Concerning Contention No. 9 l (Fire Protection) j N
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September 25, 1978
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1 TESTIMONY OF E. L. COX CONCERNING CONTENTION 9 My name is E. L. Cox. I reside at 8209 Linden Drive, Prairie Village, Kansas. I am a Project Engineer for the Black l
Fox Station project at Black & Veatch, Consulting Engineers in l Kansas City. A statement of my education and experience is attached as Attachment I to my testimony. My responsibilities at Black & Veatch include the direct supervision of the detailed Fire Hazards Analysis and the design of the fire protection systems for Black Fox Station within the scope of supply for Public Service Company of Oklahoma (PSO), as the lead Applicant.
My testimony is prepared in response to Intervenor Contention i i
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- 9. / The purpose of my testimony is to demonstrate that the Applicants have designed an in-depth fire protection system for the Black Fox Station which complies with 10 CFR Part 50, Appendix A, Criterion 3. The fire protection systems for Black Fox Station are briefly described in Section 9.5.1 of the Black Fox Station Preliminary Safety Analysis Report and, in addition, are more fully described in an extensive Fire Hazards Analysis Report which has been performed and submitted to the NRC Staff on November 28, 1977, as Black Fox Station Reference Report 16.
-1/ Contention 9 reads:
Intervenors contend that the Applicant has not designed an in-depth fire protection I
system for Black Fox 1. and 2. which complies with 10 CFR Part 50, Appendix A, Criterion 3.
9 9 f PSO addressed. the positions of Appendix A to NRC Branch Technical Position 9.5-1 " Guidelines for Fire Protection for Nuclear Power Plaints Docketed Prior to July 1, 1976," in particular, those positions specified for applicants whose application has been docketed but whose construction permit had not been received as of July 1, 1976.
The Fire Hazards Analysis designates 15 fire areas associated with safety.related equipment and structures in each unit. :iany l
of these areas are further subdivided into fire zones. The 1
l' Fire' Hazards Analysis evaluates the types and quantities of combustibles contained in each area, hypothesizes a design basis fire which can be considered to-cause the most damage and evaluates this fire to determine the impact on station safety based upon
.the design of structures and the fire detection and fire suppres-sion systems-provided.- Inadvertent operation or malfunction of the fire system is also evaluated. Where indicated by the Fire L-
' Hazards Analysis, changes to the fire protection systems were made.
The fire protection systems have been designed with an in-depth approach. This approach uses a balance of the following:
- a. Design for fire protection,
- b. Prompt fire detection.
- c. Prompt fire suppression systems.
- d. Separation of essential systems to reduce the consequences of a fire.
- e. Station personnel. trained in fire protection methods.
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- f. Administrative procedures and controls.
The design of structures, systems and components important to l
safety considers the importance of fire protection and prevention.
Fire walls and fire doors are provided to contain fires depending upon the identified fire hazard. The propagation of fires is minimized by the separation of safety systems, the placement of fire walls and the isolation of combustible materials. Non-combustible materials have been used wherever practical throughout the station, particularly in locations importan' to nuclear safety. The Fire Hazards Analysis gives the location and type j i
of fire detection and suppression systems currently being designed for Black Fox Station.
Safety related equipment is located, or other provisions are made, such that water used to extinguish fires does not affect their operational status. The design specifications require that fire protection systems be designed for proper capacity.
The systems are designed such that the effects of pipe rupture and inadvertent operation will not impair the safety function of systems, structures and components. The fire protec tion systems are designed and are to be constructed in acevrdance with acceptable fire protection codes and standards. The fire water system at Black Fox Station is dedicated to fire protection only and redundancy is provided for the Reactor Building from a safety
! related water supply. Sufficient fire water stations with hoses are located inside and outside buildings to provide a means for l
manually fighting any fire if necessary. Portable fire r
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f extinguishers are to be located as needed at~the station.
I Detectors are located in areas and ventilation ducts within I
Estructures where any significant fire potential exists. A sophisticated fire detection system is provided to give prompt l
alarm in the event of a fire and provides detail on fire location.
L Fire detection and actuation systems are powered by a reliable power source. The system is designed to minimize the potential for false alarms and yet provide a reliable fire detection system.
Automatic suppression systems are located in areas where the potential of a fire hazard exists. Charcoal filters are equipped with detectors and manual fire suppression systems.
Areas of high cable concentrations are protected by suppression systems, including the cable spreading. rooms and the floor of
.the control room. For the PGCC floor sections, a fire suppression system is provided in accordance with the General Electric criteria accepted by the NRC staff, as described in the testimony of Messrs.
Willian Gang and R. B. Johnson on Contentions 7. and 8.
PSO will provide a trained fire brigade at the station to provide fire fighting capability 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day. A detailed fire plan is being formulated by the applicant as part of the station procedures; and fire prevention is included as a part of the station administrative procedures. PSO has designated a corporate officer to be responsible for fire protection and has implemented l
special quality assurance procedures on the procurement of fire protection equipment and construction of fire systems.
1 The in-depth design of the fire protection system for Black Fox Station considers the experience of other nuclear stations as reported to the NRC and, 4.n particular, the experience gained from the fire at TVA's Brams Ferry nuclear station. Subsequent to the fire at TVA's Browns Ferry nuclear station, I ordered a compilation of all fire related incidents reported to the NRC since 1969. This compilation was sufficiently detailed to permit me to evaluate the capability of the Black Fox Station system to prevent or suppress similar occurrences. In addition, I have reviewed the "Hanauer Report" ! on the Browns Ferry fire and have considered this information in the design of the Black l Fox fire protection system.
During development of the specifications and detailed design for the Black Fox fire prevention system, I am continuing to review and consider NRC reportable occurrence reports relating l to fires or fire protection equipment. Therefore, the design criteria for the Black Fox Station fire prevention system is 1
based on the most current information available. !
As further evidence of the in-depth measures toward fire protection for Black Fox Station, Black & Veatch, as directed by PSO, transmits drawings and specifications for Black Fox fire protection systems to an independent fire protection consultant l 1
I for review and comment. The consultant's comments are incorporated
-2/ Hanauer, et al., Recommendations Related to Browns Ferry Fire, Washington, D. C.: U.S. Nuclear Regulatory Commission, 1976.
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3 into the design of Black Fox as directed by PSO. l I conclude that the Applicants have designed an in-depth fire protection system for Black Fox, 1. and 2., which complies with 10 CFR Part 50, Appendix A, Criterion 3.
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f ATTACHMENT I I ' POSITION AT BLACK & VEATCH Project Engineer - Licensing - Special Studies for Black Fox.
Station Project- j EDUCATION BS Mechanical Engineering, Case Institute of Technology, 1955 MS Nuclear Engineering, University of Missouri, 1967 l
PhD Nuclear Engineering, University of Missouri, 1970 PROFESSIONAL QUALIFICATIONS
. Registered Professional Engineer - Missouri and Ohio EXPERIENCE:
At Black.& Veatch I have responsibility for' licensing ]
activities pertaining to Black Fox Station including the technical l
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' i review of the content of the Environmental Report and Preliminary 1 i
I Safety Analysis Report, the preparation of the two documents and for coordination of communications between B&V and the NRC, I PSO and GE. Two systems in particular fall under my responsi-l bility as Project Engineer for Special Studies. These are Fire Protection and Security. I supervised the preparation of the 1 l
Fire Hazards Anal2 sis for BFS and I am responsible for coordinating L
the review of BFS design with PSO and the PSO fire protection consultant. I am responsible for the security design of Black Fox
. Station including associated design and procurement specifications.
I directly supervised the technical portions of the BFS Security
-Plan.
For ten years prior to joining Black & Veatch, I was employed by the University of Missouri in Columbia, Missouri. I served in the capacity of Director of the University's 10 MW E__ _
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Research Reactor Facility, also as Reactor Supervisor and j Associate Reactor Supervisor at that facility. I held a senior reactor operations license for the facility.
From 1959 to 1963, I was employed by General Electric Company and worked at the Va'llecitos Atomic Laboratory located near Pleasanton, California. I served in the capacity of Irradiations Engineer, Plant Engineer, Shift Supervisor and i
Senior Plant Engineer at the General Electric Test Reactor. I l held a senior reactor operators license on the General Electric Test Reactor.
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UNITED STATES OF AMERICA 4 c'o13 s#I.
[3 c'j NUCLEAR REGULATORY COMMISSION A / N!
\[a. d BEFORE TIIE ATOMIC SAFETY AND LICENSING BOARD NW ' . -
\ .t _..25 '
In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 j ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 l l
AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
Testimony of Dr. Gerald M. Gordon and Mr. William G. Gang Concerning Question 10-1 l l
September 25, 1978 l
TESTIMONY OF WILLIAM G. GANG AND GERALD M. GORDON CONCERNING QUESTION 10-1 My name is William G. Gang and I reside at 6428 Paso Los Cerritos, San Jose, California. I am the Project Manager for the supply of nuclear steam supply system components for the Black Fox Station working within the Nuclear Energy Projects !
I Division of General Electric Company. A statement of my qualifications is attached to my testimony on Contention 3.
My name is Gerald M. Gordon and I reside at 41523 Fordham Court, Fremont, California. I am the Manager of the Plant Materials Engineering Group for the Nuclear Energy Engineering Division of the General Electric Company in San Jose, California.
A statement of my background and qualifications is attached as Attachment I to this testimony.
This testimony addresses the following question posed by the Licensing Board in its Order of September 8, 1978.
10-1 Did the cracking of feedwater nozzles, con-trol rod drive return nozzles and a collet l
cylinder tube mentioned in the MHB affidavit arise because of faulty Q/A? (Specific faults in the 0/A programs at the reactors at which cracking occurred should be pointed out.) Do the same faults exist at present in the BFS quality assurance proposal?
The cracking which occurred in some feedwater nozzles, control red drive nozzles and a collet cylinder tube mentioned in the MHB affidavit was not due to faulty O/A. The feedwater and control rod drive nozzle cracks were the result of a long term, thermal hydraulic fatique phenomenon. Collet cylinder cracks occurred in a few drives because of a combination of chemical, metallurgical, and thermal cyclic factors. Neither type of cracking has any connection with the Q/A programs at the reactors where cracking occurred. The Black Fox plant draws on 300 reactor years of experience. Some low frequency phenomena not obvious at the time earlier reactors were designed have surfaced in this experience. The Black Fox plant is the beneficiary of this increased design knowledge inasmuch as the General Electric Company has made design changes to mitigate the occurrence of similar cracking at Black Fox.
l General Electric has a Quality Assurance Program which
, includes provisions for design verification. The design verifica-t ton process is sed to check design adequacy by the use of des gni review, alternate or simplified calculational methods, or per-formance of suitable qualification test. The designs were properly verified using the state-of-the-art engineering knowledge
! available at that time. However, in the case of nozzles, the cracks were initiated by a totally unanticipated operating l condition (high cycle thermal fatigue) which could not have been expected to be ascertained by a design verification. Similarly, I
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the Intergranular Stress Corrosion Cracking phenomenon, which caused the collet tube cracking, was not included in the state l
l of the design art at the time and could not have been expected to be detected by design verification. After a Design Verifi-cation Review, the design is considered frozen, and "... changes to the design documentation will only be made for the following reasons... feedback from recent plant startup testing or operat-i ing plant experience." (See NEDO-ll209-4A, " Nuclear Energy Business Group BWR Quality Assurance Program Description,"
March 1978, Section 3.10, p. 3-5.)
i The cracks discussed above were discovered in operating j
.]
plants. The GE design process specifically provides a mechanism j whereby remedies can be instituted to mitigate or eliminate the problems. The GE QA Program as set forth in NEDO-ll209-4A describes this mechanism and its control.
The feedwater nozzle cracks mentioned in the MHB affidavit were initiated by high cycle thermal fatigue. The cyclic strcases causing the cracking resulted from rapid thermal fluctuations on the nozzle blend radius inside surface. These temperature variations occurred because cool feedwater leaked between the thermal sleeve and nozzle inside surface and impinged on the hot nozzle.
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- j ll jl Plants where this leakage occurred had-thermal sleeve j 1
designs in which the seal between the thermal sleeve O.D.
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nozzle I.D.-was. formed by an interference fit. This inter- ,
l forence fit unanticipated 1y relaxed slightly after several reactor startups and~ shutdowns, allowing the leakage described and the unpredicted complex thermal mixing. This understanding of the causative mechanism is based on extensive experimental and analytical work performed by General Electric immediately l after the problem occurred in 1976. In the Black Fox design, this problem has already been corrected by the use of a piston ring between the thermal sleeve and nozzle; a configuration i qualified by full scale testing at the Moss Landing Power Plant.
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Cracking of the control rod drive return was-also due to ;
j a thermal cycling resulting from non-equilibrium flow effects after long' periods of-time. The control rod drive return line nozzle has been eliminated from the Black Fox station design, l
thereby eliminating the concern.
! Stress corrosion cracking occurs when three contributing conditions are present: (1) susceptible material, (2) stress,
,and (3) environment. Based on extensive General Electric research, it is now known that the cracks observed in the collet cylinder tubes were caused by high, thermally induced cyclic residual stresses in a part sensitized during processing and exposed to an environment conducive to intergranular stress corrosion cracking.
A modified collet cylinder tube has been developed which addresses all three considerations. The following changes are incorporated in this new design which is used for the Black Fox plant.
1 A. Elimination of a section change in order to reduce discontinuity stresses.
B. A change in the material to type CF3 castings with controlled ferrite to provide resistance to stress corrosion.
C. A change in equipment to provide low oxygen coolant for the drives during operation so that the aggressive-ness of the environment is reduced.
Detailed qualification testing has demonstrated that these changes to the Black Fox design have eliminated the concern that collet cylinder tubes will experience cracking.
In summary, none of the three cracking conditions mentioned had any relationship to plant Q/A programs. Further, the actual causes of all the incidents have been determined and design changes qualified and implemented to minimize further occurrences.
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p ATTACHMENT I GERALD M. GORDON PRESENT POSITION
~ Manager, Plant Materials Engineering, Nuclear Energy Engineering Division, General Electric Company.
EDUCATION B. S. - Metallurgical Engineering, Wayne State University, 1956
(- Ph.D. - Metallurgical Engineering, The Ohio State University, 1959 l
TECHNICAL ASSIGNMENTS' Prior to joining General-Electric,.I was a Senior Metallurgist at Stanford Research Institute, Menlo Park, California from 1959-63. I served as a Project Leader on a number of govern-ment and commercially sponsored programs in the areas of high temperature oxidation and mechanical performance of refractory metal alloys.
Ijoined the General Electric Company Nuclear Energy. Division in 1964 as a Senior Metallurgist in the Reactor Materials Development Group at Vallecitos. I became Manager of the Metallurgy Development Component in 1969. This group had materials research and development responsibility for physical metallurgy, fracture toughness and radiation damage of
i i l l l 1 i
reactor materials and aqueous corrosion and stress cracking e of nuclear reactor pressure boundary and internals materials' .
In 1973, I became Manager of the Zircaloy Performance Group 1
l with responsibility for. development and evaluation of nuclear- l I
fuel cladding and channel materials. I also served as l
Manager, Plant Component Behavior Analysis and was respon-1 l' sible for implementation of laboratory. developments in design of reactor plants. 'I assumed my current position as Manager, Plant Materials Engineering in 1976 and am currently l
responsible for evaluating and specifying BWR plant materials ,
l as well as materials surveillance and identification and solution of potential or actual stress corrosion cracking problems.
I am a Registered' Professional Engineer in California and a Fellow of the American Society of Metals. I have authored numerous publications and patents and I have been an invited lecturer or Session Chairman at several International Confer- ,
ences on Corrosion & Stress Corrosion Cracking. I am. currently Chairman of the National Association of Corrosion Engineers ,
Committee T-llA on Corrosion in High Purity Power Plant Water.
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NUCLEAR REGULATORY COMMISSION 9 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD d>I In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )-
4 Testimony of Mr. J. B. Perez Concerning Questions 10-2 and 10-3 l
i september 25, 1978
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TESTIMONY OF J. B. PEREZ CONCERNING QUESTIONS 10-2 and 10-3
)
I reside at 5922 E. 48th Street, My name is-J.'B. Perez.
Tulsa, Oklahoma. I am the Manager of Quality Assurance for Public Service Company of Oklahoma and my responsibility includes.the QA Program for the Black Fox Station Project.
I. report directly to the Vice President of Power Generation, who is. independent.of BFS' Project Management. This organiza-tional structure provides me with the authority and organiza-tion freedom to perform my duties in accordance with the j requirements of Criterion I of 10 CFR Part 50 Appendix B. I l have held the position of Manager, Quality Assurance since December, 1977. Prior to that, I was the Quality Control Supervisor for the Black Fox Station project and I have-been l employed with PSO Quality Assurance since 1974. A statement l
of my background qualifications is attached as Attachment I l
.to my testimony.
My testimony addresses Questions 10-2 and 10-3 posed by
! the Licensing Board in its Order of September 8, 1978.
L , -Question 10-2 reads as follows:
l Is WASH-1309 undergoing revision?
Is-applicant committed to conforming to the latest revision?
I l
-Response PSO is committed to the " Green Book", WASH-1309, " Guidance on Quality Assurance Requirements During the Construction Phase of Nuclear Power Plants", May 10, 1974. WASH-1309 consists mainly of regulatory guides and ANSI Standards in draft form.
Implementation of the regulatory guides included in WASH-1309, including any revisions to them, is addressed in Section 1.9 of the PSAR. We have verified through the staff that WASH-1309 is not undergoing revision and that the 1974 issue is still in l effect.
Question 10-3 reads as follows:
What experience in the nuclear quality assurance area do the members of Applicants' Q/A staff have?
Response
The experience of the PSO Quality Assurance staff in the nuclear quality assurance area is listed below:
l Jesus B. Perez, Manager, Quality Assurance l I have 7-1/2 years experience in quality assurance activ- ;
1 ities. For six of those years, I was directly involved in )
nuclear quality assurance. I was a Metallurgical Engineer j in the Quality Assurance Department of YUBA Heat Transfer Corporation for 2 years (1971-1973). After that, I spent 1-1/2 years as the Quality Control Manager for Con-Rad Divi- l sion of U.S. Industries. I was the PSO Quality Control l I
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Supervisor for the BFS Nuclear Project for approximately 3 years (1974-1977). I have been in my present position as Manager, Quality. Assurance, for PSO since December, 19./. j i~
Harry H. Eller - Site Quality Assurance Superintendent.
Mr. Eller has 20 years' experience in quality assurance in manufacturing, construction, and installation activities.
For eight (8) of those 20 years of experience, Mr. Eller has I been directly involved in nuclear quality assurance activities.
He has had 3 years as a Quality Engineer for Peter Kiewit Company during construction of Fort Calhoun and Cooper Nuclear l
Station (1970 - 1973). After that, Mr. Eller spent 3 years i
with JELCO involved in the Liquid Metal Fast Breeder Reactor i
program, where he held the positions of Quality Control Engineer, Site QC Manager, and Corporate QA Manager; and 9 months as Site Quality Control Supervisor for Gibbs and Hill at'the Fort Calhoun 2 Project. Mr. Eller has been the Site QA Superintendent in the BFS Project since April, 1977.
Wayne J. Kropp - Quality Control Supervisor Mr.'Kropp has had 6 years of experience in the U.S. Navy
! Nuclear Power Program plus 7 years of nuclear quality assur-ance experience in the power industry. From 1971 to 1974, Mr. Kropp was performing procurement QA activities for Duke Power Company. From 1974 - 1975, he developed QA procedures l
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-for Fluor Pioneer and from 1975 to 1977 he was Source Sur-veillance Supervisor for Brown and Root on the Comanche Peak and. South Texas projects. .He started working in the PSO QA I l
organization in March, 1977. l i
Robert D. Smither - Quality Surveillance Supervisor Mr. Smither worked as an Electrical Engineer for Public
~ Service Company of Oklahoma for 14 years and has been in the BFS Project for 5 years. He has been in the QA organization since April, 1977.
A. W. Pingry - Senior Source Surveillance Specialist Mr. Pingry has had 11 years' experience in the quality assurance field with 8 years being in nuclear quality assur- ;
ance. Prior to joining PSO Quality Assurance in 1976, Mr.
Pingry worked for Conam Inspection at Dresden 1 and 2, Three Mile Island, Zimmer, Indian Point, and LaSalle 1 and 2.
i R. F. Casey - Qu'.lity Control Specialist Mr. Casey has had 17 years' experience in QA/QC activ-ities with 2 years' experience in nuclear quality assurance.
L Prior to joining PSO, Mr. Casey was a Senior QA Specialist
.with Brown and Root in the South Texas Project.
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J. C. Acosta - Source Surveillance Specialist Mr. Acosta has had 10-1/2 years' experience in quality )
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assurance with 4 years' experience in nuclear quality assur-ance. Mr. Acosta has worked for U.S. Testing, Universal i
Testing Labs, and Quality Industries at Three Mile Island l and Oyster Creek nuclear projects. Prior to joining PSO in January, 1978, Mr. Acosta was a QC Inspector for. Daniel Inter- j national on the Wolf Creek Project (1977 - 1978).
W. J. Ferguson - Senior Quality Surveillance Specialist Mr. Ferguson has had 5 years in quality assurance with 3 years' nuclear QA experience with YUBA Heat Transfer (1974 -
1977) prior to joining the PSO Quality Assurance organization in November, 1977.
l G. R. Kimmell - Senior Quality Assurance Auditor Mr. Kimmell has had 7 years' experience in the U.S. Navy Nuclear Power Program plus 1 year and 9 months of nuclear QA experience with *he PSO Quality Assurance organization.
G. W. Geren - Site Quality Engineer - Civil Mr. Geren has had 11 years' QA/QC experience with 1-1/2 years of nuclear QA experience in.the PSO Quality Assurance organization.
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l M. W. Johnson - Site Quality Engineer - Audits Mr. Johnson has had 9 years in the U.S. Navy Nuclear
[ Power Program. He also has three (3) years of experience in nuclear QA with Metropolitan Edison at Three Mile Island Nuclear Plant (1973 - 1976).
W. E. Hagan - Quality Control Specialist Mr. Hagan has had 24 years' total QA/QC experience with i
4 months of Nuclear QA experience in the PSO Quality Assurance {
i organization. I 1
J. P. Phillips - Quality Control Specialist Mr. Phillips has had 13-12/ years' total QA/QC experience with 1 year of Nuclear QA experience in the PSO Quality Assur-ance organization.
L. Sanford - Source Surveillance Specialist Mr. Sanford has had 8 years' total QA/QC experience with 8 months of Nuclear QA experience in the PSO Quality Assurancc organization.
D. L. Vavra - Quality Assurance Specialist Mr. Vavra has had 6 months of Nuclear QA experience with PSO. He has 6 years' Nuclear experience in the U.S. Navy and 2 years' nuclear experience with the Omaha Public Power District.
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As shown in the above list, the Quality Assurance staff i
has extensive experience in the nuclear quality assurance area (50 years). The staff is also very highly experienced in the overall Quality Assurance area doing QA/QC activities I
directly related to their present job function (143 years). I 1
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ATTACHMENT I MANAGER, QUALITY ASSURANCE: J. B. Perez EDUCATION:
B.S., Metallurgical Engineering, 1967 M.S., Metallurgical Engineering, 1969 University of Texas at El Paso Additional Education and Training Ultrasonic Inspection - ASNT, 1971 i
Applied QA Audit Techniques - L. Marvin Johnson, 1974 Radiography - OSU Extension Course, 1975 i
Quality Control - OSU Extension Course, 1976 J 1
Training and Qualification Seminar - ASNT, 1977 Registered Professional Quality Engineer in California (QU-3317) ]
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EXPERIENCE:
As Manager, Quality Assurance for Public Service Company of Aklahoma, I report to the Vice President of Power Generation and I am responsible for the development and implementation of a Quality Assurance program during the design, construction, and operation of Black Fox Station (BFS). Prior to this assign-ment, I was the Quality Control Supervisor and was responsible for implementation of the QA Program for procurement of equip-ment and services for BFS. My duties as Quality Control I
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Supervisor included-the QA review of procurement documents, evaluation of. suppliers' QA programs, and source surveillance l
to verify ~the quality of purchased equipment and services. l Prior to employment with PSO, I was Quality Control Manager l
1 at Con-Rad, Division of U.S. Industries (Heat Exchanger l
}
L manufacturer for the power and petro-chemical industry).
As Quality Control Manager, I developed and implemented a QA program to meet ASME Section VIII requirements. My duties also included the supervision of the shop Q.C. inspectors, l development of inspection plans, qualification of welders and weld procedures, and training of nondestructive testing personnel.
From 1971 to 1973, I was a Metallurgical Engineer in the QA j
department of YUBA Heat Transfer' Corporation. At YUBA Heat, I was responsible for testing purchased material against ASME requirements, testing of weld procedures, writing Q.C. proce-dures,'and maintenance of a calibration control program. I also assisted in the development of a QA program to meet ASME Section'III requirements. (YUBA Heat Transfer manufactures heat exchangers for the nuclear industry).
Prior to joining YUBA Heat, I was involved in metallurgical research work for the Army at Frankford Arsenal in Philadelphia, Pennsylvania.
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/ / 9 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD / lg.vg j
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3 In the Matter of )
)
PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
COOPERATIVE, INC. ) ;
)
(Black Fox Station, Units 1 and 2) )
i Testimony of Mr. Donald G. Long t
Concerning Question 10-4 4 l
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i l September 25, 1978
l TESTIMONY OF DONALD G. LONG CONCERNING QUESTION 10-4 My name is Donald G. Long and my business address is l 175 Curtner Avenue, San Jose, California. I am the Manager, Quality Systems in the Product and Quality Assurance Opera-l tion of the General Electric Company. A statement of my
[
qualifications is attached as Attachment I to my testimony. ,
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My testimony addresses the_following question posed by 4
the Licensing Board in its Order of September 8, 1978:
10-4 Would there be a substantial improvement in quality ssurance for the components listed l I
in Contention 10 if the quality assurance j program required formal identification of !
each process which is to be treated as a "special process" within the meaning of I
Criterion IX of Appendix B to 10 CFR Part 50?
' General Electric Company supplies the following items from the hardware listed in Contention 10:
- 1) Pressure Vessel
- 2) Control Rods l
- 5) Gas radwaste equipment
2-L 10CFR Part 50, Appendix B, Criterion IX lists three processes as "special processes" but it does not define the term "special process". The NRC has not provided industry i l !
l with a formal definition of "special processes". Conse-quently, in order to comply with Criterion IX, GE has exer-cised its judgment in determining whether a particular process l
in a "special process". The starting point for determining l
whether a process was "special" was an evaluation of those processes identified in Criterion IX for comnion elements. It
]
was then necessary to develop criteria which would identify 1
processes for which the Criterion IX requirements were appro-l
! priate and necessary. GE has concluded that a process is "special" within the meaning of Criterion IX if it meets the l
following criteria: l l
- 1) The process affects or measures safety-related-functional aspects of an item.
- 2) The results of the process are highly dependent on the control of the process or the skill of j the operator, or both.
~
- 3) The specified quality cannot be readily deter- l l
mined by inspection or test. 1 The first criterion is based on Appendix B requirements applying only to the safety-related functions of structures, I systems, and components. The second and third criteria are l \
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consistent with the definition of "special process" contained in the latest draft of ANSI /ASME NQA-1, " Quality Assurance Program Requirements for Nuclear Power Plants."
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i The GE QA Program identifies as "special" those processes L
! which meet the above criteria. These processes include those J
identified in Criterion IX, as well as brazing, for'certain applications.
GE agrees that it is a necessary element of-the QA Program to identify "special processes" in order to comply with the requirements of Criterion IX. This GE has done. The'GE QA Program requires that those processes identified as "special" be controlled and accomplished by qualified personnel using qualified procedures as required by Criterion IX.
Processes identified by Mr. Hubbard such as neutron sensor plating and crimping of control cables are not "special processes" in that they no not meet the above defined criteria.
For example, inspection and testing of the results of these two processes are feasible, and, in fact, the'results are inspected and tested.
l- Because a-process is not designated as "special," and therefore'not subject to Criterion IX requirements, does not I
mean that a process is uncontrolled. Necessary procedural controls, qualification of processes or personnel training l-
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is performed commensurate with the importance of.the activity.
or process to product quality. Sensor plating and cable l l-l crimping, as examples, .are controlled processes and the neces-i sary controls to assure product quality are specified in the GE QA Program. The arbitrary designation of such processes as "special" might serve to further formalize and extend their documentation, but such designation would not substantially
. improve product quality.
In summary,.GE feels that-it is necessary to identify l "special processes" and to comply with the requirements of criterion IX relative to processes so identified. This GE does. Other processes not identified as "special" are con-trolled commensurate with the importance of the process to product quality. 1 l
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! l l ATTACHMENT I DONALD G. LONG I am employed by the Nuclear Energy Business Group of I
the General Electric Company in San Jose, California. I am i
currently Manage -Quality Systems in the Product and Quality
( Assurance Operation. In this capacity I have responsibility for planning and documenting the overall structure of the quality system within which the quality assurance program for l
l nuclear power plant systems and components, supplied by the Nuclear Energy Business Group, is developed. As Manager, Quality Systems, I was also responsible for preparing the l
General Electric portion of the quality assurance program t description contained in the Black Fox Project PSAR.
I I received a B.S. degree from Oregon State University in 1956 and, following graduation, joined the General Electric Company. I graduated from the General Electric Company's th ree-year Manufr aturing Management Program in 1959 and have been associated with the nuclear field since that date.
During the three-year period from 1956 to 1959, while on 1
( the Manufacturing Management Program, I held a number of posi-tions in manufacturing engineering, production control, facilities engineering, and manufacturing administration. l l
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i Since 1959, I have been associated with nuclear power plant ,
i l l quality assurance-related activities at General Electric i '
facilities located in San Jose. From 1959 to 1962, I held l
the position of Specialist-Quality Review and Quality Control j i i Engineering, responsible for quality control planning for {
l nuclear fuel manufacturing activities. In 1962, I became Manager, Quality Control-Fuels, responsible for planning and <
i implementing quality assurance and quality control activities associated with the manufacture of Boiling Water Reactor nuclear fuel. I held the position of Manager, Quality Control-Fuel, until 1966 when I was appointed Advance Quality Control 1
Engineer, responsible for quality system planning for Atomic Power Equipment Department manufactured and procured items.
In 1968, I was promoted to the position of Specialist-Quality Systems for the Nuclear Energy Division-BWR business activities.
In 1972, I was promoted to the position of Consultant-Quality Systems, for the Nuclear Energy Division. I have been in my present position as Manager-Quality Systems since 1974.
I am a member of the American Society for Quality Control and I have presented papers to the Society in the general field of quality assurance. I am a registered professional engineer in quality engineering in the State of California. For the l past nine years, I have been intimately involved in the develop-ment of ANSI N45.2 series standards covering the field of quality i
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- l. i assurance of nuclear-power plants. I was a member of the I
original N45 Ad Hoc committee on quality. assurance program l l
requirements which first met in 1969, and I have continually I i
1 served on ANSI & ASME committees responsible for the develop-i' ment of N45.2 series standards since that date. Currently I'am a member of the ASME Committee on Nuclear Quality Assur- J l
'ance-and a member of the Subcommittee on General Requirements.
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- i UNITED STATES OF AMERICA C
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t}a I NUCLEAR REGULATORY COMMISSION I#' / . igV" l
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD L In the Matter of )
I. )
l PUBLIC SERVICE COMPANY OF OKLAHOMA, )
- ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 l
AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
l i I
l Testimony of Mr. C. J. Ross Concerning Questions 12-1, 12-4 and 12-5 September 25, 1978 I
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TESTIMONY OF C. J. ROSS CONCERNING QUESTIONS 12-1, 12-4 and 12-5 My name is C. J. Ross, I resideRat 10114 Pawnee Lane, Leawood, Kansas. I am Manager-Design for the Black Fox Project (BFS) at Black & Veatch, Consulting Engineers in Kansas City.
As Manager-Design for Black Fox Station, I am responsible for coordinating and directing project design activities. A state-ment of my education and experience-is attached as Attachment I to this testimony. My testimony is prepared in response to the Licensing Board's questions relating to intervenor contention 12.
Question 12-1 reads as follows:
Is inspection of the rack anchors and hold down bolts necessary to insure structural capability, and, if so, have provisions been made for such inspection?
As explained below, inspection of the rack anchors and hold down bolts is not necessary to insure structural capability because the materials and design used are more than adequate to accommodate expected stresses. The spent fuel storage racks including the anchors and hold down bolts, as currently described l
l 'in the PSAR, will be conservatively designed to Seismic Category I requirements. The stresses on the anchors and bolts emanate l from two sources, namely, static loads and loads resulting from the design basis safe shutdown earthquake. When the stresses from these loads are combined (the maximum expected load), the stresses imposed on the rack anchors and bolts will be less than
i one-half of the allowable stress for the material used, namely, stainless steel.(ASTM A193).
However, inspection of the rack' anchors and hold down bolts can be accomplished, in place, should it be-desires using
'means such as underwater television, boroscope, etc. q It should also be noted that.the Applicant intends to install at some future time high density rather than low density racks. Some designs of high density racks are free standing and I
do not require anchors and hold down bolts. Other designs may require anchors and/or bolts. If the latter design is selected,
! .the Applicant will assure that the structural design will meet the same structural capability as described above.
Question 12-4 reads as follows:
Does the spent fuel pool design provide for an adequate source of water to ';
1 fill the pool and maintain its level during operation?
Yes, the' design does provide for an adequate source of make-up water. The water level in the spent fuel pool will be maintained through the addition of make-up water supplied from
- the Condensate Storage and Transfer System. Additionally, redundant loops of the Standby Service Water System (which are both Seismic Category I) can be used as a source of make-up water i should the Condensate System not be available.
Question 12-5 reads as follows:
Are there off-normal conditions under which the design of the Spent Fuel
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l Filter and Demineralizing System would permit an undue hazard to arise?
There are no off-normal conditions under which the design of the Spent Fuel Filter and Demineralizing System would permit an undue hazard to arise. The purpose of the system is to main-tain water clarity and minimize corrosion and fission product buildup. Water clarity is needed to conduct fuel transfer operations. If the system fails and the water becomes murky, fuel transfer operations will be halted. This action presents l no undue hazard to the health and safety of either the public or station personnel.
Radioactivity is released to the spent fuel pool as a l
l result of the solution and dispersal of corrosion and fission products from the fuel. The dispersal of this radioactivity i throughout the pool water will create a " shine" dose rate which t l
l 1s limited to the personnel located around the pool. This dose rate under normal conditions (when the system is operating) will be less than 0.3 milleram per hour. The absence of a cleanup system will increase the dose rate to approximately 3.3 millerem per hour for personnel standing directly adjacent to the pool l railing. This dose rate together with operating procedures which limit the exposure time, does not constitute an undue hazard to the health and safety of station personnel. This low dose rate is localized to the area directly adjacent to the spent fuel pool and does not affect the site boundary dose rate to the public.
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ATTACHMENT I !
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PROJECT MANAGER: Charles J. Ross EDUCATION:
BS Mechanical Engineering, San Jose State College, Nuclear Power Option, 1962 MEMBER: ANS PROFESSIONAL REGISTRATION:
Registered Professional Engineer: Kansas, Oklahoma EXPERIENCE:
I am presently assigned as Manager-Design on the Black Fox Station project with responsibilities in management of the firm's design engineering activities and coordination of engineering activities with the Public Service Company of Oklahoma.
I have more than 15 years experience in nuclear power
. plant work, having participated in NSSS evaluations and contract negotiations, reactor operations, plant startup, overall review of plant systems, safety analysis reports, project management, and conferences with the AEC-DRL staff and the ACRS.
From 1962 to 1966, I was a reactor engineer and later senior reactor engineer at Phillips Petroleum Company's National Reactor Testing Station. I was responsible for operation of the reactor I
l and various associated engineer experiments.
I joined General Electric Company in 1966 and was assigned shift supervisor of the EVESR at General Electric Company's Vallecitos Nuclear Center (Licensed Senior Reactor Operator) where I was responsible for safe, efficient operation of the reactor, related facilities, and experiments. In 1967, I was named mechanical design engineer in General Electric's Atomic Power
r Equipment Department where I was responsible for the design and development of reactor servicing equipment.
In.1968, I joined Northern States Power Company as a nuclear engineer and participated in engineering activities associated with the Monticello and Prairie Island nuclear genera-ting plants. In-1972, I was named project engineer for the Tyrone Energy Park where I was responsible for overall coordination, I
direction, liaison and preparation of engineering activities concerned with two 1100 MWe nuclear units. While associated ;
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'with Northern States-Power Company, I served as a member of the Monticello Safety Audit Committee as well as chairman of the Technical Committee for the Standardized Nuclear Unit Power Plant Symbol (SNUPPS).
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NUCLEAR REGULATORY COMMISSION VA M- x' q4' #
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD k\ %~AT'U;. -
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l In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC )
i COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and ; )
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Testimony of Dr. Gerald M. Gordon l Concerning Question 15-1 I
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l September 25, 1978 u
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I f i TESTIMONY OF GERALD M. GORDON CONCERNING QUESTION 15-1 My name is Gerald M. Gordon and I reside at 41523 Fordham Court, Fremont, California. I am the Manager of the Plant Materials Engineering Group for the Nuclear Energy. Engineering i l
Division of the General Electric Company in San Jose, California.
My statement of qualifications is attached to my testimony in answer to Question 10-1.
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l -My testimony addresses the following question posed by the Licensing Board in its Order of September 8, 1978.
l L 15-1 Will General Electric be committed to remedial measures in parts of the Black Fox system where very recent (or future) experience indicates IGSCC may occur, as well as in parts of the system where such cracking has occurred in the past 10-15 years?
l t 1 General Electric is committed to remedial measures in the
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parts of the Black Fox system where past experience, including j very recent experience, indicates intergrcnular stress corrosion cracking ("IGSCC") may occur. Recall that only 132 out of 17,000 l
pipe welds within the coolant pressure boundary have experienced l IGSCC in any BWR. Of these, the bulk has been in the recircula-tion bypass line, the core spray line, and the control rod drive l^
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hydraulic return line. Counter measures have been identified !
and qualified for these lines. Since the bulk of all history
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of IGSCC cracks has been addressed in operating plants, the already small probability of future IGSCC occurrences has been l further diminished. i General Electric's commitment to identify, develop and l I
qualify practical IGSCC countermeasures is clearly demonstrated j l
by the numerous well known programs undertaken in the past l several years for parts experiencing IGSCC in operating plants.
- 1 General Electric's maintenance of an extensive ongoing sur- j l
veillance program and large qualified technical staff and test facilities testifies to our desire for early detection of any future occurrence of IGSCC and to our intent of identifying auitable countermeasures for these components.
The material changes specifically applied to Black Fox for stainless steel piping are the result of programs suggested by a special interdisciplinary General Electric Task Force investigation conducted in 1975 to determine the cause of cracking in stainless steel piping lines. Potential improve-ments were identified and extensively tested. Only after being proven were they implemented into Black Fox and other plants.
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The improved collet cylinder tube is an example of response to cracking identified by General Electric's surveillance pro-gram. This cracking was shown not to be a safety concern. It i l
was thoroughly investigated and an improvement identified.
Implementation in plants under construction as well as operating plants was done to enhance plant availability by reducing inspection and repair requirements over the long term.
l More recent changes made to the Black Fox plant to further l
avoid IGSCC incidents include the use of feedwater spargers which are solution heat treated after welding, and Control Rod Drive Housings which are fabricated from Type 316L stainless steel, one of the newly qualified alloys in the programs men-tioned above.
o of remedies in operating plants has con-Implementati'n sistently been based on cooperation between General Electric and utilities with the joint aim of improving plant availability.
In summary, General Electric's past performance demon-strates a comprehensive program to minimize the probability of the occurrence of IGSCC in the Black Fox plant. Many changes have been made in the Black Fox plant due to General Electric's past efforts in preventing IGSCC. General Electric plans to continue surveillance and test programs to identify areas of incipient IGSCC and to qualify and implement countermeasures where appropriate.
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UNITED STATES OF AMERICA I '
I, NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELEC'PRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPERATIVE, INC. )
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(Black Fox Station, Units 1 and 2) )
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Testimony of Mr. Dwane R. Glancy Concerning Question 18-1 September 25, 1978
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TESTIMONY OF MR. DWANE R. GLANCY CONCERNING QUESTION 18-1 l
My name is Dwane R. Glancy and my business address is Box 1
201, Tulsa, Oklahoma. I am treasurer of Public Service Company of Oklahoma (PSO) and have the direct responsibility of the external financing, cash management, financial budgeting and ,
i forecasting, and rate area at PSO. The financial information l
provided to the Nucle r Kegulatory Commission in Exhibit III to
" Black Fox Station, Application for Licenses, Construction Permit Stage" was prepared under my direct supervision. A statement of my qualifications is attached to my testimony as Attachment I.
My testimony addresses the following question posed by the Licensing Board in its Order of September 8, 1978:
18-1 Has PSO provided different data on coverage ratios for bonded debt to NRC and OCC, and, if so, what is the reason for the difference?
On May 5, 1977, the NRC forwarded to PSO a request for specific additional information regarding the financial qualifica-tion of PSO and its partners in the Black Fox Station. The requested information was collected and provided to the NRC as Exhibit III to the " Black Fox Station, Application for Licenses, l
Construction Permit Stage." The following coverage ratios for 1
bonded debt were requested frem PSO and were provided:
i a) Projected coverage ratios for each year through completion of Black-Fox Station Unit 2 based on the Company's Indenture requirements and coverage ratios based on the SEC method of calculation.
l b) The actual Indenture coverage ratio for the year ended August 31, 1977.
c) Actual " Times total interest earned before Federal Income Taxes," " Times long-term interest earned before-Federal Income Taxes" and " Times interest l
and preferred dividends earned after Federal Income Taxes" for year ended 1975 and 1976 and 12 months ended August, 1977, all calculated as requested by the NRC.
The coverage ratios were provided in the exact form and format requested by the NRC whether or not the Company uses the same information in the same form for other corporate purposes.
The Oklahoma Corporation Commission (OCC) does not require that any particular coverage ratios be furnished in connection with a rate request. PSO furnished the OCC coverage ratios relevant to bonded debt based on the Company's Indenture require-ments. The Indenture coverage ratios have been furnished to the OCC because if adequate coverage, as defined in the Indenture, is not maintained, PSO cannot continue to issue bonded debt under l
its first mortgage indenture, and this impairs PSO's ability to provide service to its customers at a reasonable cost. PSO also
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furnished the OCC " Time charges earned before taxes" based on the Moody's method of calculation. This coverage ratio is relevant to preferred stock issuances. Moody's method of calculating coverage ratios is used by the Company for many corporate purposes. All coverage ratios provided to the NRC and the OCC were identified by the method of calculation used.
The only difference between the Moody's " Times charges earned before tax" and the SEC coverage ratio is that the SEC's formula includes the interest portion of financing leases. This amount is so small that the resulting change in coverage is in-significant. The SEC coverage would be slightly lower. Generally, coverage ratios furnished to the OCC are on a historical or test year basis, while many of the coverage ratios furnished to the NRC are on a forecast basis. Wherever the historical calculations were presented to the OCC and NRC, they were identical provided the same type of coverage ratio was used. If a different type coverage ratio was used for the same time period, the base data used for developing the ratios was the same although the methodology of the ratios would differ.
Therefore, the coverage ratios furnished to the NRC and the OCC were based on the same data base and were, therefore, f consistent; each coverage ratio was identified for what it was, and each coverage ratio was appropriate for the purpose for which l it was filed. Any slight differences between the ratios resulted l
f from the fact that the NRC did not request the precise information that PSO furnishes the OCC.
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ATTACHMENT I QUALIFICATIONS OF DWANE R. GLANCY I am Treasurer of Public Service Company of Oklahoma. I received a BSBA with a major in accounting from the University of Missouri in 1964 and a Master of Arts in Accounting in 1968 from the same university. From 1964 to 1966 I was employed by i
Arthur Young & Company in Tulsa, Oklahoma, as a staff auditor w.th the majority of my assignments being related to the oil and gas industry. In 1968 I joined the firm of Arthur Andersen &
Company in Kamas City, Missouri. I was associated with Arthur Andersen eight years all of which time I worked in their regulated industries division dealing with the regulatory, financial and accounting problems of utilities, truck lines, banks and bank I holding companies and other transportation related industries.
I became an audit manager with Arthur Andersen & Company in 1973.
My primary responsibility was in engagements concerning the public 1
financial statement reporting to shareholders and reporting to the SEC, FPC, ICC and FRB. During the last five years I spent over half my time conducting and supervising special engagements concerning registration statements to the SEC and ICC for the sale of securities, the economics and indenture considerations of leasing and trust financing arrangements, and rate cases before state commissions, the FPC and the ICC.
l I am Certified Public Accountant in the states of Oklahoma and Missouri. I joined Public Service Company of Okli>homa in 1976 as Assistant to the Senior Vice President, Finance and was elected Treasurer in 1977.
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UNITED STATES OF AMERICA f '
NUCLEAR REGULATORY COMMISSION b s i >jW BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
l In the Matter of )
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PUBLIC SERVICE COMPANY OF OKLAHOMA, ) Docket Nos. STN 50-556 ASSOCIATED ELECTRIC COOPERATIVE, INC., ) STN 50-557 AND WESTERN FARMERS ELECTRIC -)
COOPERATIVE, INC. )
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(Black Fox Station, Units 1 and 2) )
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l Testimony of Mr. Edward D. Fuller Concerning Contention 67 l (Anticipated Transients Without Scram) l September 25, 1978 l
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TESTIMONY OF MR. EDWARD D. FULLER CONCERNING CONTENTION 67 (ANTICIPATED TRANSIENTS WITHOUT SCRAM) J My names is Edward D. Fuller and I reside at 132 Teresita Way, Los Gatos, California. I am the BWR Licensing Manager for the Safety and Licensing Operation of the Nuclear Energy Projects Division for General Electric Company in San Jose, California, the supplier of the nuclear steam supply system for the Black Fox Station. A statement of my background and qualifications is-attached as Attachment I to my testimony.
My testimony concerns Intervenor Contention 66 regarding the anticipate ( t.ransients without scram-(ATWS) issue. Mr.
John'Zink of Public Service Company of Oklahoma will address in his ,
i testimony the manner in which the design of the Black Fox I
Station will account for ATWS remedies advocated by the NRC's 1 Division of' System Safety.
I. BACKGROUND ATWS is composed of two postulated events: a transient I
occurring sometime during the plant lifetime that could conceivably
- i Contention 67 reads f The analysis by the Applicants and the Staff of the facilities' response I to certain anticipated transients with simultaneous failure of the scram system (ATWS) have underestimated ob'th the consequences of such events and their
' likelihood, to such an extent that the facilities present an undue hazard to the health and safety of the public.
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- lead to severe consequences; and the failure of the scram system to function on demand. For an ATWS situation to occur, both postulated events must occur concurrently. Accordingly, the probability of the ATWS event occurring is the product of the i
individual probabilities of each of the events.
Both the General Electric Company and the customer / utilities I
have been carrying on a continuing dialogue regarding the likeli- i hood of such events, and the prospective systems that would ease the severity of such postulated events. These discussions have gone on as the BWR/6 product line has evolved.
In 1969, the ACRS first raised the concern of ATWS events.
Since that time, the General Electric Company has performed a great deal of work in this area, with the result that several topical reports have been issued. In response to a request by the Atomic Energy Commission (AEC) in 1971, GE supplied an analysis of ATWS events.
In 1973, the AEC issued WASH-1270 requiring mitigation of I
postulated ATWS events and classifying mitigation by plant vintage, i.e., differing schedules for mitigating actions were suggested depending on whether the plant was operating, undergoing construction or in the initial design stage. WASH-1270 led to the development by GE of substantial additional information on ATWS which was furnished to the NRC Staff in 1974 and 1975. In 1975, the Staff I issued " status reports" to each vendor requiring further analysis and defined the ATWS safety goal. These reports were heavily I
I l I criticized by industry and numerous meetings were held with the Staff to attempt to resolve these differences. In 1976, General Electric issued two reports to respond to the 1975 Status Report.
In 1977 as a response to all of the criticism by the vendors I and utilities, the Division of System Safety (DSS) within the Office of Nuclear Reactor Regulation (NRR) initiated an extensive reevaluation of all of the information available on ATWS. This 1
reevaluation led to the issuance of NUREG-0460 in 1978. j l
NUREG-0460 represents a DSS position as noted by the dis- l claimer on the report. It is not a position of the NRR nor of j the NRC Commissioners. Presently the NRR is reviewing NUREG-0460 1
but is awaiting a recommendation letter from the ACRS prior to l l
making its recommendation to the Commissioners. It is likely )
l that rulemaking and possibly hearings will be required prior to the promulgation of a rule or regulation. In the meantime, no NRC regulations exist concerning ATWS. NUREG-0460 is still only a DSS report.
NUREG-0460 has been reviewed and evaluated by General Electric. The resultant evaluation and GE position on ATWS was then sent to NRC management in July 1978, presented to the ACRS subcommittee in August 1978, and will be presented to the full ACRS in October 1978. This presentation which includes an evalua-tion of event probabilities, a realistic ATWS safety goal, and a l value impact assessment is set forth below.
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l II. EVALUATION OF-ATWS ISSUE A. Frequency and Consequences. Two elements that contribute to the likelihood of ATWS occurring are the frequency of transient occurrence and reliability of the scram systam. Regarding the reliability of the BWR scram system, the General Electric position is based on the results of a comprehensive study performed by the General Electric Company. The results of the study which is the basis for the General Electric position is that existing
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scram system unreliability is about .8 x 10 per demand, the major contributor or scram rmreliability being the scram logic sensors. This is as opposed to the value of 3 x 10 per demand as specified in NUREG-0460.
The frequency of transients per reactor year is the subject of a continuing study by Electric Power Research Institute (EPRI). j I
The EPRI study is directed toward data collection and analysis of 1 transients which have occurred in operating nuclear plants. The data collection, which started late in 1975, is based on data provided by the utilities directly to EPRI.
This study considered generic scram system design both with relay and solid state electrical systems and considered, as well, the control rod drive mechanical systems. This study, which consumed eight manyears of effort, analyzed all of the related scram systems; reactor protection system relay logic, mechanical components, hydraulic control units, scram air headers and scram discharge volumes. A large number of failure modes effects analyses were developed through this study. A number of potential common cause failures were examined and a number of reported individual component abnor-malities were factored into the study.
1 In July 1978, EPRI issued report No. EPRI NP-801 describing the results of their findings to date which considered data from a significant number of operating plants. The results of that study, which have been confirmed by GE, indicate a frequency l of occurrence for significant transients of 3.5 events / year for BWR's. This is as opposed to a frequency of 6 events / year as stated in NUREG-0460.
When one combines the probability of a significant transient occurring (3.5 events /yr) along with the probability of failure
-6 to scram (.8 x 10 /yr.), one arrives at an ATWS probability of 3 x 10~ /yr. This differs by almost 2 orders of magnitude from the NUREG-0460 value. It is General Electric's position that !
any potential ATWS fix should be consistent with this low probability
-6 of 3 x 10 / year.
Several features of the Black Fox Station design serve to minimize the consequences of an ATWS event should one occur. The i
recirculation pump trip feature has the capability to reduce system power to 30% of full power thus providing overpressure protection and aiding nuclear shutdown. The main steam safety / relief valves provide additional overpressure protection. The capability for l complete nuclear shutdown is available through manual initiation of the standby liquid control system. Adequate cooling of the cores while shutdown is occurring is achieved by existing ECCS ,
Through a l system inventory as well as by normal feedwater. I
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combination of these and other features, the consequences of l postulated ATWS events involving turbine / generator trip do not i
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1 that most transients when combined with failure to scram lead j 1
to core melt. At least half of the expected transient events lead to insignificant-core damage.
B .- Safety Goal. The ATWS safety goal should be consistent.
with the accepted frequency of other accidents with similar consequences. As noted in WASH-1400, the ATWS event frequency of
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1 x 10 / reactor year is at one-fifth the frequency of other events postulatud to lead to core melt. It should also be noted i
that the ATWS consequence is not as severe as that from the ]
types of core melts considered in the WASH-1400 study. Consequently, an ATWS safety goal of 10 / year is a reasonable limit based on 1
J other risks _already being taken. The current scram system meets this reliability goal. This compares with an ATWS safety goal-
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of 10 as set down in NUREG-0460.
C. Value and Impact. Considering the ATWS event from the standpoint of value and impact, it is General Electric's position that the extremely low probabi2ity of the event does not justify modifications to existing nuclear power plant designs. However, it is of interest to view the value-impact assessment in NUREG-0460 l using realistic values for the probability of the ATWS event. In assessing the value of avoiding the ATWS event, the upper band benefit is $0.3-0.7 million as opposed to the NUREG-0460 value of $42-70 million. With regard to impact, consideration must be given to potential costs associated with installation of mitigating
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features. For example, one such cost'is the spurious initiation of an automatic boron injection: system which has been predicted to occur once in every-10 years per plant. These spurious injections could occur from operator error.during surveillance testing or' equipment failures. It has been determined that such spurious injections could result in a downtime of at least 1 month per injection for.the present 86 gpm system. -This would result-in a cost impact of $60-million/ plant due to decreased plant availability.
D. Conclusion. The frequency and consequences do not justify designating ATWS as a design basis accident. The.ATWS
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safety goal should be 10 / year based on other already accepted risks. Finally, a realistic assessment of value impact is required in order to determine 'the cost basis for designating ATWS as a design basis accident. As part of this assessment, one must consider the potential cost impact of ATWS mitigating features such as that associated with spurious boron injections
, 4 from an automated boron injection system. I 1
g III. RELATIONSHIP OF ATWS TO THE BLACK l FOX CONSTRUCTION PERMIT PROCEEDING.
L The purpose of this testimony is to apprise the Licensing
-Board and the parties as to the present status of a technical' )
l issue that currently is the subject of intense review and considera-tion by-the ACRS, NRC and the nuclear industry. This testimony
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clearly indicates areas of sharp disagreement between the NRC i
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Division of System Safety's report NUREG-0460 and GE's position.
These disagreements will be finally evaluated and reconciled by the NRC Commissioners after receiving the advice of the ACRS and public comment during the process of rulemaking. It would, of course, be: premature and inappropriate for the Black Fox Licensing Board to decide these matters since there is neither a licensing requirement for ATWS nor any requirement for implementa- !
tion of NUREG-0460.
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l ATTACHMENT I BWR LICENSING MANAGER: Edward D. Fuller i EDUCATION: BA, Physics, 1962 San Jose State University MS, Nuclear Engineering, 1964 Stanford University i ADDITIONAL EDUCATION AND TRAINING:
Survey of Reactor Design, General Electric, 1962 Advanced Engineering Courses, General Electric, 1965 - 1967 i .
Probabilistic Systems Analysis, General Electric, 1969 BWR Quality Assurance Training Course, General Electric, 1976 l
l Professional Business Management, General Electric, 1969 1
Management Development Course, General Electric, 1974 PROFESSIONAL RECOGNITION:
Registered Professional Engineer in California, No. 1934 - Nuclear Chairman N'1 clear Fuel Cycle Division, American Nuclear Society, l
1975-76, Member Power Division Executive Committee, American Nuclear Society, Author of many papers on BWR Fuel and System Performance EXPERIENCE:
As Manager of BWR Licensing for General Electric since June of 1976, I am responsible for General Electric's licensing support to each domestic BWR project, including defining required engineering analysis and supporting documentation to meet NRC requirements for Construction Permits and Operating Licenses. My responsibilities include the direction of GE's engineering program I and licensing action concerning the ATWS matter.
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UNITED STATES OF AMERICA SD s NUCLEAR REGULATORY COMMISSION Ne BEFORE THE ATOMIC SAFETY AND LICENSING BOARD h )/
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PUBLIC SERVICE COMPANY OF OKLAHOMA, )
ASSOCIATED ELECTRIC COOPERATIVE, INC., ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPERATIVE, INC. )
)
(Black Fox Station, Units 1 and 2) )
i e i Testimony of Dr. John C. Zink 1 Concerning Contention 67 (Anticipated Transients Without Scram)
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September 25, 1978 ,
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TESTIMONY OF DR. JOHN C. ZINK CONCERNING CONTENTION 67*
(ANTICIPATED TRANSIENTS WITHOUT SCRAM)
My name is John C. Zink and I reside at 12518 E. 134th -
1 Street, Broken Arrow, Oklahoma. .I became an employee of the l t
Public Service Company of Oklahoma in 1975 when I was assigned
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to the Black Fox Station Nuclear Power Project as Supervisor of l
Nuclear Engineering. On October 1, 1978, I will be promoted to j i
the position of Manager, Nuclear Fuels. Prior to my employment l \
with the Public Service Company of Oklahoma, I was an Assistant Professor of Nuclear Engineering at the University of Oklahoma from 1970 to 1975. I received B.S. and M.S. degrees in mechanical engineering -- nuclear option from the University of Notre Dame, !
and I received my Ph.D. in Nuclear Engineering from the same University in 1970. My testimony addresses the manner in which l l the design of the Black Fox Station has or will account for ATWS remedies advocated by NRC's Division of System Safety.
Public Service Company of Oklahoma has carefully followed I the ongoing industry-wide discussions and regulatory activities l
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regarding Anticipated Transients Without Scram (ATWS). I have l-l Contention 67 reads:
! The analysis by the Applicants and the Staff of the facilities' response to certain anticipated transients with simultaneous failure of the scram system (ATWS) have underestimated both the consequences of such events and their likelihood, to such an extent that the facilities present an undue hazard to the health and safety of the public.
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l 'l been. assigned the responsibility of following the status of .
ATWS and of considering the possible implications of the ATWS I issue for purposes;of' assessing its impact on the Black Fox q I
Station design. Since May of 1977, I have attended seven meetings l concerning this issue. These meetings ranged from presentations
.by General Electric to meetings involving larger industry groups, I meetings with the NRC Staff, and meetings with the Advisory Committee on Reactor Safeguards'.
Due to the high reliability of current. scram systems, Public Service Company does not believe that ATWS is a concern.
Our position is in agreement with the position held by other utilities,-Architect-Engineers and reactor manufacturers. It i i
1 is clear that ATWS is a generic issue which affects all light l- water reactors, both those currently in operation and those which I
will become operational ~in the future. Public Service Company has committed to. incorporate into Black Fox Station the generic resolution that is finally determined by the NRC Commissioners.
We recognize that the resolution of this issue, may ultimately incorporate some of the systems described by the NRC Division of System Safety (DSS) in NUREG-0460. One of the mitigation systems referenced is the recirculation pump trip feature discussed on pages 5 and 6 of Mr. Fuller's testimony. The recirculation pump trip feature is desirable for BFS Nuclear Steam Supply System because it accommodates the end-of-cycle scram reactivity changes, i
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a phenomenon unrelated to the ATWS event. For this reason, it has been incorporated into the Black Fox Station.
A high capacity automatic boron injection system, incor- !
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- porating larger pumping capacity and additional sodium pentaborate solution storage volume, was also considered by DSS in NUREG-0460. Should the ultimate generic resolution of ATWS require the installation of this system, adequate space has been provided in the Black Fox Station to accommodate it.
Thus, of two potential " fixes", one -- the recirculation pump trip feature -- is included in the design of the Black Fox Station, and inclusion of the other -- the high capacity automatic l boron injection system -- will not be foreclosed by the construction i of Black Fox Station.
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e 1S UNITED STATES OF AMERICA y [/ ,
NUCLEAR REGULATORY COMMISSION l m 6 -3
!~ BEFORE THE ATOMIC SAFETY AND LICENSING BOA In the Matter of the Application of ) N l Public Service Company of Oklahoma, )
! Associated Electric Cooperative, Inc. ) Docket Nos. STN 50-556 i and ) STN 50-557 j Western Farmers Electric Cooperative )
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i (Black Fox Units 1 and 2) )
CERTIFICATE OF SERVICE l
l I hereby certify that a copy of the foregoing NOTICE OF FILING APPLICANTS DIRECT TESTIMONY AND IDENTIFICATION OF EXHIBITS was mailed, postage prepaid, in the United States l Post Office, this 25th day of September, 1978, to the following:
Sheldon J. Wolfe, Esquire L. Dow Davis, Esquire William D. Paton, Esquire Atomic Safety and Licensing Board Panel Colleen Woodhead, Esquire l U.S. Nuclear Regulatory Counsel for NRC Staff l Commission U.S. Nuclear Regulatory l Washington, D.C. 20555 Commission j Washington, D.C. 20555 l Mr. Frederick J. Shon, Member
! Atomic Safety and Licensing Mr. Clyde Wisner Board Panel NRC Region 4 U.S. Nuclear Regulatory Public Affairs Officer Commission 611 Ryan Plaza Drive, Suite 1000 Washington, D.C. 20555 Arlington, Texas 76011 Dr. Paul W. Purdom Joseph R. Farris, Esquire i
Director, Environmental Studies Green, Feldman, Hall & Woodard l Group 816 Enterprise Building Drexel University Tulsa, Oklahoma 74103 32nd and Chestnut Streets Philadelphia, Pennsylvania 19104
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Mrs. Ilene H..Younghein- Mr. Vaughn L. Conrad' l I
3900 Cashion Place Public Service Company of oklahoma City, Oklahoma 73112 Oklahoma P.O. Box 201 Atomic Safety and Licensing Tulsa, Oklahoma 74102 l Appeal Board Panel U.S.. Nuclear Regulatory
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Mr. T. N. Ewing Commission Acting Director Washington, D.C. 20555 Black Fox Station Nuclear Project Public Service Company of Atomic Safety and Licensing Oklahoma Board Panel P.O. Box 201 U.S.-Nuclear-Regulatory Tulsa, Oklahoma 74102 l
- Commission l Washington, D.C. 20555 Mrs. Carrie Dickerson
! Citizens Action for Safe Docketing and Service Section Energy, Inc.
Office of'the Secretary of P.O. Box 924 the Commission Claremore, Oklahoma 74107 U.S. Nuclear Regulatory Commission Mr. Maynard Human Washington, D.C. 20555 General Manager (original only -- the 20 copies Western Farmers Electric will be delivered by messenger Cooperative on 9/26/78) P.O. Box 429 Andarko, Oklahoma 73005 1 Mr. Lawrence Burell l Route 1, Box 197 Dr. M. J. Robinson Fairview, Oklahoma 73737 Black & Veatch j P.O. Box 8405 l Mr. Gerald F. Diddle Kansas City, Missouri 64114 General Manager Associated Electric Cooperative, Inc.
l P.O. Box 754 Springfield, Missouri 65801 l
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I Jos ~ Gaflo' Ond/@df tNe Attorneys l
for Applicants i