ML20064G349
| ML20064G349 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/30/1978 |
| From: | Weiss E SHELDON, HARMON & WEISS |
| To: | |
| References | |
| NUDOCS 7812110101 | |
| Download: ML20064G349 (82) | |
Text
~
- '8 NRC PUBLIC DOCU;INI ROOM #
gp UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSIhd APPEAL BOARD In the Matter of ) ,
)
VIRGINIA ELECTRIC AND POWER ) Docket Nos. 50-338 OL COMPANY ) 50-339,OL
) L (North Anna Nuclear Power ) T Station, Units 1 and 2) )
UNION OF CONCERNED SCIENTISTS REQUEST ,
FOR LEAVE TO FILE REPLY BRIEF h UCS has now received copies of both Vepco's and the[- '
Staff's responses to our brief amicus curiae in the ab'ove-captioned case. The Staff's response was not received until Friday, November 24th as a consequence of UCS not being included on the Staff's service-list.
UCS requests permission to file a b'rief response to the ;
Vepco and Staff positions. One week would be sufficient for this purpose. We would also like to direct the Appeal Board's attention to the following portions of the record which bear on the Vepco and Staff position that the plant-( '- -
specific analysis of the turbine missile issue for North Anna is sufficient to resolve the outstanding questions:
- 1. Limited Appearance of Robert D. Pollard,
, D N. Tr. 3010, g o -
yfgr 2. The ASLB's request to the Staff and Vepco e '
.i to respond, Tr. 3600, q
'il.7,"?: 'T," ~
0 lI 3. Applicant's response, letter of July 5,
.T 1977, (copy of relevant portions attached)
- 4. Staff response of Sept. 2, 1977, (copy of relevant portions attached) j 8121101C)I
6 ! ')It 9
3
- 5. Robert Pollard's comments on Staff and Applicant responses, Sept. 26, 1977, (copy attached)
- 6. Licensing Board questions to Staff, 4 Oct. 21, 1977, (copy attached)
- 7. Staff's response, Nov. 23, 1977, (copy of relevant portions attached)..
Our request for permission to respond stems substantially from the fact that the Staff's filing with the Appeal Board appears to differ in material fashion from its previous responses to the Licensing Board. As a prime example, in w,
its response to the ASLB dated November 23, 1977 (Number 7 ,,'
above), the Staff stated (page 1) :
It should be noted that the staff's conclusions
' in Section 10.2 of Supplement No. 2 to the North Anna Power Station, Units 1 and 2 Safety Evaluatin[Fic]
Report limit the acceptability of the risk to being sufficiently low for permitting the plant to operate until a generic study on the subject turbine missiles is completed. (Emphasis added) -
In contrast, the Staff has now stated to the Appeal Board:
. . . [T]he Staff does not consider the turbine missile risk analysis for North Anna Units 1 and 2 to be dependent on Task A-37, and concludes. ')
the matter to be resolved for North Anna Units 1 ./
and 2. (Emphasis added, NRC Staff's Response to UCS Brief Amicus Curiae, p. 3)
In contrast to its earlier position, the Staff now appears to be saying that the generic problem is simply irrelevant to North Anna. UCS would like the opportunity to respond.
Respectfully submitted,
( i s. / a. < lJ,- .
Ellyn R. Weiss Sheldon, Harmon, Roisman & Weiss 1025 15th Street, N.W.
Suite 500 Washington, D.C. 20005 (202) 833-9070 Counsel for Union of Concerned Scientists DATED: November-30, 1978
,Y #
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY'AND LICENSING APPEAL BOARD In the Matter of )
)
VIRGINIA ELECTRIC AND POWER ) Docket Nos. 50-338 OL COMPANY ) 50-339 OL
)
(North Anna Nuclear Power ) .
Station, Units 1 and 2) )
CERTIFICATE OF SERVICE I certify that I have served a copy of the Union of Concerned Scientists Reply for Leave to File Reply Brief on each of the persons named below by first-class mail, f
Secretary James N. Christman, Esquire U.S. Nuclear Regulatory Commission Hunton & Williams Washington, D.C. 20555 P.O. Box 1535 ATTENTION: Chief, Docketing & Richmond, Virginia 23212 Service Section Daniel T. Swanson, Esquire
- Michael C. Farrar, Esquire U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Washington, D.C. 20555 Appeal Board U.S. Nuclear Regulatory Richard M. Foster, Esquire Commission 1908-A Lewis Mountain Road Washington, D.C. 20555 Charlottesville, Virginia 22903 Anthony J. Gambardella, Esquire
( office of the Attorney General
- Suite 308 #
11 South 12th Street 53 Richmond, Virginia 23219 4, semm Alan S. Rosenthal, Esquire S .\ ,
- Atomic Safety and Licensing Appeal Board ; I-NOV3 01978 P ,-
U.S. Nuclear Regulatory Commission =
Washington, D.C. 20555 E' op;g,*; t""'
sun.
Dr. John H. Buck 4
- Atomic Safety and Licensing Appeal Board or U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i Atomic Shfety and Licensing Appeal-Board * ~
U.S. Nuclear Regulatory Commission -
l Washington, D.C. 20555 - . . / .uc
. . ., . j i
- Documents 3 through 7 enclosed Ellyn R. Weiss l Hand-delivered 11/30/78 l
i
i s ~,, '. r *
- SHEI. DON, HARMON, RorsMAx & WEISS so2s esa sTRECT, N.W.
SutTC soo MARIN P. sH CLDON WASHINGTON, D. C. 2ooos T C LE Pe*O N C (202) 833 9070 GAIL M. HARMON ANTHONY 2. ROISMAN CLLYN R. WCISs WILLIA M S , JO RD AN,111
- ADMITTED $N MICHIGAN ONLY '
November 30, 1978' '
Daniel T. Swanson, Esquire Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 RE: Docket Nos. 50-338 OL, 50-339 OL
]
Dear Mr. Swanson:
Apparently UCS has not been included on the Staff's service list for the North Anna proceedings, despite the Appeal Board's granting of our request to file an amicus brief. As a result, we only received word of the Staff's response to the brief through a third party and did not receive a copy of your filing until last Friday, November 24th.
This is to request that you have UCS placed on.your official service list to ensure that we receive all future filings relevant to the issues outstanding before the Appeal Board.
Service can be made to me as counsel for UCS.
Very truly yours,,
i ti3 ks O
fJ Ellyn R. Weiss G./l.ala eJ
- EsTE Sheldon, Harmon, Roisman & Weiss k 1025 15th Street, N.W.
-sI Suite 500
(;4- N0'/.7 e7 . , ,,,, ,_ 01978 Washington, D.C. 20555
> f (202) 833-9070
- <g;,te. !
\
q -,
/
- ' Counsel for Union of Concerned W M Scientists ERW/dmw cc: Service List
.. 9 ;,
3 Above information supplied by: y O. James Peterson, III (Vepco) Ts7TE
~
NOV3 01978 > r I c' *1 C S RESPONSE TO g y ,
MR. POLLARD o
Mr. Robert D. Pollard, in his limited appearance statement (Tr. 3010-23), suggested that a number of features of f North Anna 1 and 2 do not meet. NRC regulations. The ASLB asked Vepco and the Staff to list parts of the plant that do not comply with current rules and regulations and to' explain why they need not do so.
Vepco believes that all parts of North Anna Units 1 and 2 comply with the NRC regulations. Our responses to Mr.
~
Pollard's allegations are set out below. -
(
INDEPENDENCE OF REDUNDANT SAFETY EQUIPMENT D
1 .
Mr. Pollard's cor.cerns about the independence of redundant safety equipment (Limited Appearani:e Statement of Robert D. Po'lla rd 4-7) are the following:
- 1. . No discussion in North Anna SER of how the design 10/ Hereinafter " Pollard Statement." ,
E
t s, f ' f, ' '
o ,
e complies with the IEEE 279-1971, General Design Criterion 3, and General Design Criterion 21;
- 2. No regard for " associated circuits" as outlined in Regulatory Guide 1.75 and IEEE std. 384-2974;
- 3. No discussion of basis of separation distances in potential' missire-producing areas;
- 4. IE inspection Report No. 50-338/76-4 (cable tray fill); and
- 5. Cable separation in the main contcol' room.
The regulations Mr. Pollard cites are 10 CFR 50.55a(h)
(which he says requires conformance to IEEE 279-197111) and General Design Criteria 3 and 21 in Appendix A to 10 CFR Part f
~50. Section 50.55a(h) reads as follows:
Protection systems: For construction permits issued af ter January 1,1971, protection systems shall meet the requirements set forth in editions or revisions of the Institute of Electrical and Electronics Engineers Standard: " Criteria for Protection Systems for Nuclear Power Generating Stations," (IEEE-279) in effect on the formal docket date of the application. for a construction permit.
Protection systems may meet the rdquirements set forth in subsequent editions or revisions of IEEE-279 which become effective ( footnotes omitted) .
The two Design Criteria in Appendix A (which became effective *~' -
e May 21, 1971, 36 Fed. Reg. 3255) are as follows:
l 3. Fire protection. Structures, systems, and 11/ The doc'ket date for North Anna 1 and 2 was April 4, 1969, l
so the proposed IEEE 279 that became "in effect" on August 30, 1968, should apply rather than IEEE 279-1971, which became "in effect" on June 3, 1971, see n. 7 to 10 CFR'S 50.55. However, as noted below, Vepco has complied with the more recent of these two standards.
l 1
- ,s components'important to safety shall be designed I and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the uniti particularly in i locations such as the containment and control room.
Fire detection and fighting systems of appropriate- !
capacity and capability shall fbe provided and designed to minimize the adverse effects of fires - -
on structures, systems, and components important to !
safety. Fire-fighting systems shall be designed to -t assure that their rupture or inadvertent operation j does not significantly impair the safety capability ;
of these. structures, systems, and components.
~
- 21. Protection system reliability and testabilitv.
( The protection system shall be designed for high !
functional reliability and inservice testability commensurate with the safety functions to be i per formed . Redundancy and independence designed into the protection system shall be sufficient to l u
assure that (1) no single fail're results in loss !
of the protection function and (2) removal from ;
service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the !
protection system can be otherwise demonstrated. l The protection system shall be designed to permit i periodic testing of its functioning when the !
reactor is in operation, including a capability to ,
test channels independently to determine failures !
and losses of redundancy that may have occurred. i
( !
- The Introduction to Appendix A says that these Design Criteria .
/ are not to be ~ imposed blindly:
[T]here may be. vater-cooled nuclear power units for
- which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General ,
Design Cciteria must be identified and justified. l Vepco has provided a design that complies with IEEE 279-1971 and with General Design Criteria 3 and 21.- The NRC ;
l I
l c
l
r,,- v, ' -
and are completely isolited from the controls in the control' room. These panels were provided specifically to deal with the case of the " uninhabitable control room."
References:
North Anna Units 1 and 2 FSAR:
Section 3.5 Section 7.7 Section 8.3 Response to Comments in FSAR:
S3.2 S3.54 S3.55 S3.56 S8.1 D8.2 D8.3 S8.7 Letters to NRC: Serial Nos. 231 and 599 of 8-22-75 and 9-13-76 (separation problem correction)
Above information provided by:
J. Barnhart (Stone & Webster)
J. M. Davis (Vepco)
]
TURBINE MISSILES Mr. Pollard suggests (Pollard Statement 7-10) that North Anna 1 a'nd 2, may not meet General Design Criterion 4:
Environmental and missile desian bases.
Structures, systems, and components important to safety shall be designed to accommodate the effects of and to
based on the sires of the missiles that are considered credible in areas where redundant circuits may be routed. The design avoids the routing of safety related raceway in missile areas except where necessary.
The 'K' cable trays have always been'an allowable fill ,
of 50%. The ' K ' service class cables are defined as having no I2 R Losses, intermittent service, or cocontinuous service with a derating of 40%, as stated in.the FSAR and Specifications NAS-238. This provides a very conservative
(
power level loading for these trays (well below IPCEA allowable), and no significant heating of cable occurs.
The computer floor in the main control room has been reworked to ensure that separation, independence, and single failure concerns are satisfied. Although safeguards cables are t
r ou ted in this area, the cables are separated or barriered from one another. The potential for fire which "would make the l entire control room uninhabitable and limit or prevent initiation of safety action" is very remote, because this space i
under the floor is provided with an effective detection and '
automatic flooding fire extinguishment system. Moreover, an
" uninhabitable control room" would not prevent or even limit the initiation of ' safety actions, because redundant auxiliary shutdown panels are available, as described in FSAR S 7.7. The auxiliary panel controls do not pass through the control room
/
w ,' ' e
~1 to be on the order of 10 per year to satisfy Regulatory Guide 1.115 recommendations.
Mr. Pollard misquotes SER S 10.7 as saying that the probability of strike on essential systems is on the order of 2
-1 Actually, this probability represents the strike x 10 .
probability on a building that houses essential systems. The probability of strike on an essential system can be found by' multiplying 2 x 10 ~1 by the density of essential systems within each build ing. This probability is itself conservative, .
- because it neglects external and internal walls and barriers that would stop or slow the missile. .
Reorientation of the turbine with respect to the containment building is not a requirement of Regulatory Guide 1.115. The Guide (which, like all Regulatory Guides, is not itself a regulation) says simply. that reorientation is one i
means of mitigating the effects of low-trajectory turbine missiles. Finally, Mr. Pollard's assertion (Pollard Statement 9) i i that the rotational energy of the missile was ignored in the ,f r penetration calculation is simply not true, Westinghouse has !
~!
performed tests and analyses on the contair nent of disc burst i
- fragments. In those tests it was observed that the rotational {
component of the fragment neither helped not hindered , t l \
+
. i 1
be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, 'that may result from equipment failures and from events and conditions outside the nuclear power units. < In fact, turbine missiles have been a design consideration .on the North Anna plant in that the risk, taking l into account both the probability of generation and the probability of damage, has been found to be acceptably low. L Information about the missiles postulated inside the containment and the protection provided to . mitigate the effects of these missiles is in Section 3.5 of the North Anna 1 and 2 FSAR. Also, the matter was adequately discussed with the ACRS j Subcommittee on North Anna on . October 13, 1976. Regulatory Guide 1.115 says that the probability of damage to essential systems should be less'than 10 -7 per year. This corresponds to a probability of strike of less than
-3 10 per missile event, if the historical turbine failure rate is assumed. However, Westinghouse has performed f ault tree analysis- of the turbine overspeed protection system, giving due consideration to improvements in turbine materials, valving, and controls, and estimates a probability of turbine -
-6 failure on the order of 10 per turbine per year. Using this turbine failure rate, the strike probability must be shown
~
t t
m when one considers the extremely low probability of this specific accident scenario. The overall accident probability includes the probabilities of a double-ended pipe rupture at the BCP discharge, system restraint failure, and missile cause and its additional consequences. This matter was discussed by the Staff at the ACRS subcommittee meeting on R. E. Pollard's concerns on Thermal-Hydraulics, Vibration, and Pump Overspeed, March 31, 1976. The Staff concluded that the overall
-8 -11 probability was between 10 and 10 per facility-year and that this probability, compared to ot'aer . well-known )
low probability occurrences, discredits the accident scenario. Vepco and Westinghouse do not agree with Mr. Pollard's belief that, since generic studies are being performed on pump performance characteristics during a loss-of-coolant accident, it has not been demonstrated that the public health and safety is adequately protected. The g.eneric stud'ies are being per fo rmed for the purpose of substantiating or modifying current mathematical models that predict the pump performance - parameters during a LOCA. The Westinghouse topical report on this subject (WCAP-8163, " Reactor Coolant Pump Integrity in LOCA," September 1973) thoroughly' documents the calculation methods used to obtain RCP overspeed. These calculations were conservatively done. Westinghouse views the verification by the generic studies to be an additional conservatism on what is
.' ,o perforation and in fact was dissipated in heating and smearing of metal.
Above information provided by: ; i j J. L. Vota (Westinghouse) REACTOR COOLANT PUMP (RCP) FLYWHEEL MISSILES ' i Mr. Pollard -{inds f ault with the NRC Staff's treatment j of RCP flywheel missiles (Tr. 3018), saying that the SER gives no basis for accepting the current d.esign. In fact, the Staff's app' roved acceptance criteria for
~
i i RCP flywheels may be found in Regulatory Guide 1.14, Rev.1 i (August 1975), " Reactor Coolant Pump Flywheel Integrity." This : i Regulatory Guide provides details for acceptable material selection, flywheel fabrication and design, and preservice and l i ! l inservice inspections. ; Vepco and Westinghouse agree with the Staff's position i on RCP flywheels as stated in Section 5.4.1 of the North Anna 1 ; and 2 SER. The use of suitable construction material, adequate de sign , and inservice inspections of the flywheels provide reasonable engineer ing assurances. These engineering measures are particularly reassuring . i
3 As Mr. Pollard recognizes, the SER says that the subject of Class I equipment qualification is still open. vepco has supplied and is still supplying extensive information on the qualification of safety related equipment in order to satisfy all concern about its capability. At the time the equipment was being designed and purchased for the North Anna plant, the standards to which Mr. Pollard refers did not exist and so could not be used as acceptance bases; therefore the standards that were available, including IEEE 323-1971, were
,)
used. This does not mean, however, that equipment qualified to the older standards is inadequate. Mr. Pollard points out that NRC advisor Dr. Hanauer found IEEE 323-1971 to be general rather than specific, but that does not invalidate that standard, because its value depends on its intended use. IEEE 323-1971 would be dif ficult to use as a test procedure but was valuable in crecaring' test procedures. The standard is not a "how to" standard but a " test shall include or consider" i
/
standard. It is true that the 1974 version of IEEE 323 provides in more detail the environmental parameters to be considered in qualification, including aging. It is also true that the testing per formed "on the equipment required to operate in hostile conditions at North Anna 1 and 2 did not include several factors associated with long-term operation. However,
- ,' ,a O
already a very conservative design. Above information provided by: J. L. Vota (Westinghouse) SE'ISMIC AND ENVIRONMENTAL QUALIFICATION OF SAFETY EQUIPMENT Mr. Pollard suggests (Pollard Statement 10-12, Tr. 3013-15) that the seismic and environmental qualification of safety equipment at North Anna 1 and 2 does not comply with General Design Criterion 4, quoted above, and General Design Criterion 2, which reads as follows: f
~
Design bases for protection against natural phenomena. Structures, systems, and components - important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and' components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, .(2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be per formed. In particular^, Mr.. Pollard is concerned that Regulatory Guide 1.89, IEEE 323-1974, and IEEE 344-1975 are not being applied to North Anna.
= ..
conservative, would encompass present-day requirements.- Also, the Staff has very recently (June 14, 1977) established a regulatory position on assuring that the environment in operating nuclear stations is maintained within the temperature i range for which Class IE equipment has been shown to be qualified to operate. Vepco will be responding shortly to that position and will comply with the Staff's position as it applies to North AnG2. t the following information on this subject jRe fer ences has previously been submitted: FSAR Descriction S3.38-1 Environmental qualification of safety features for DBA operation , S3.70-1 Qualification test program for BOP IE equipment S7.15 Class IE equipment temperature transients S7.16 during MSLB conditions. Qualification of S7.17 BOP IE equipment outside containment. 58.5 Qualification tests performed - ~) on IE cable. Letters and Attachments Descriction Serial No. 341 Qualification of equipment required 1 (11-26-76) , , for LOCA. Ef fects of MSLB on ' - equipment. Qualification of BOP IE equipment for postulated environments Serial No. 100A Electrical penetration qualification i (4-25-77) repor ts seal material data, proposed areas to be temperature ' t l l t i l k _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
~
,. ,'. a the. goal of long-term availability was addressed by the qualification program for the. North Anna equipment; for example, transmitter / instrument testing included radiation dosage for 40 years' operation, and electric valve operators included radiation dosage sufficient to envelope 40 years'
~
operation plus accident condition, life cycling of the mechanical components, and thermal aging of the electric operator. , Mr. Pollard's comparison of IEEE 344-1971 to IEEE 344-s 1975 is likewise wihhout merit. It is a fact that the 1971 version of IEEE 344 regarded the sine beat testing methods as ; the preferred choice. However, Mr. Pollard misrepresents IEEE 7 344-1975 as requiring multi-axis, multi-frequency testing. In fact, the 1975 version also allows single-axis sine beat testing, as well as other methods. Westinghouse has submitted a topical report (WCAP-8373) which documents the conservative nature of Westinghouse single-axis sine beat testing (testing at resonant f re quenc ie s) . Moreover, the seismic testing recently performed by Westinghouse was of a multi-frequency, multi-axis nature; this testing demonstrated the operability of < the equipment tested. I
~
The NRC Staff has conducted extensive reviews of the t seismic qualification data on Class IE equipment to verify that , the tests and design methods used, whici are quite WCAP demonstrates the conservatism of the single axis-sine beat method of seismic testing employed by Westinghouse. This topical report is referenced in North Anna FSAR, Section 3.10. August 1974 -- Westinghouse submitted to AEC WCAP-8234,
~
" Topical Report, Seismic Testing and Functional Verification of By-Pass Loop Reacter Coolant RTD's."
This topical report is referenced in North Anna FSAR, Section 3.10. j December 1974 -- AEC removed WCAP-7817 from the list of , accepted reports and requested additional information concerning electrical operability of equipment (Vassallo to Eicheldinger) . July 1975 -- Westinghouse proposes supplemental seismic and environmental qualification program to NRC (NS-CE-692,
~
Eicheldinger to Vassallo) and initiates program. For
.)
North Anna, the applicable portions include environmental testing of instruments and environmental analysis and evaluation of valve operators; seismic testing of Westinghouse 7300 Process bistables and Nuclear Instrumentation System bistables and providing additional information regarding the solid state protection system testing and transmitter testing. . July 1975 -- As part of NS-CE-692 (Eicheldinger to Vassallo) Westinghouse provided additional information concerning . ' seismic qualification of the solid state protection system and transmitter testing and additional information regarding the environmental qualification , of valve operators. November 1975 -- NRC accepted Westinghouse proposed program ' /) outlined in NS-CE-692 as being satisfactory for resolution of NRC concerns on WCAP's 7817 and 7744 (vassallo to Eicheldinger) . November 1975 -- NRC formally approved WCAP-8234 concerning ! seismic qualification of by-pass loop RTD's (Vassallo [ to Eicheldinger) . j November 19'7 5 '- ' Westinghouse submitted , as par t o f NS-CE-692, listing of Westinghouse supplied safety related motor operated valves located inside containment for plants committed to the program, including North. Anna l ) (NS-CE-847, Eicheldinger to Vassallo) . f f l i l ; t w . . _ _ - .= _ - - - . _ .
monitored Serial No. 223 Cl. ass IE cable qualification data. (6-6-77) Chronologv l The following chronology shows the extensive . ! Westinghouse qualification activities that are applicable to l North Anna Units 1 and 2: : 1969 - Westinghouse initiated seismic and environmental [ qualification programs for Westinghouse scope safety l related electrical equipment. September 1971 - Westinghouse submitted WCAP-7744, l
" Environmental Testing of Engineered Safety Features ;
Related Equipment" to the AEC. This re. port described ; testing of safety related electrical equipment which could be exposed to the effects of postulated , acc iden ts . For the North Anna application, the equipment is limited to pressure and differential pressure transmitters and valve operators. This topical report is referenced in North Anna FSAR, Section 3.11. January 1972 -- Westinghouse submitted WCAP-7817, " Seismic Testing of Electrical and Control ~ Equipment", including initial ^ supplements to the WCAP, to the AEC. This report and its supplements describe the seismic ( qualification testing of Westinghouse supplied safety related electrical equipmont. North Anna FSAR Section 3.10 references the WCAP and appropriate supplements : for qualifications. January 1973 -- WCAP-7817 and supplements 1 through 4 formally ; approved by AEC Letter dated January 12, 1973 (DeYoung to Salvatori). March 1974 '-- AEC. identified need for additional information to complete review of WCAP-7744 (Vassallo to Salvatori) . August 1974 -- Westinghouse submitted to AEC WCAP-8373, !
" Qualification of Westinghouse Testing Procedure for :
Electrical equipment Tested Prior to May 1974." This f 1 i
, . _ . , -_-g--_... _ , _ _ .
e ,
and Stone and Webster met with NRC Staff to review 3 environmental conditions inside containment resulting l from postulated pipe ruptures and the environmental i capability of equipment. Virginia Electric Power Company indicated it would replace instruments which are determined unacceptable by the Westinghouse testing program. October 1976 -- Status of seismic and environmental programs applicable to North Anna were presdnted at the ACRS North Anna subcommittee meeting in Washington, D.C. October 1976 -- To satisfy NRC request at September 1976 meeting, Westinghouse provided to NRC, for information, WCAP-8541, " Topical Report Seismic and Environmental Testing of Foxboro Transmitters" (NS-CE-1251, Eiche1dinger to Stolz)'. Foxboro transmitters are used 4 to monitor pressurizer pressure in the North Anna plant.
~
).
November 1976-January 1977 -- Final topical reports covering multi-frequency, biaxial testing of bistables submitted ' to NRC by Westinghouse. Those applicable to North Anna are:. WCAP-8831, " Seismic Operability Demonstration Testing of the Nuclear Instrumentation System Bistable Amplifier" and WCAP-8829, " Seismic Operability Demonstration Testing of the Westinghou.se ISD 7300 Series Process Instrumentation Bistables." December 1976 -- Westinghouse submitted, under NS-CE-1315, a report entirled " Sensor Qualification Program, IEEE 323-1971 Demonstration Program; Interim Test Report for the Barton/ITT Modified Differential Pressure Tr a n s= i tte r . " This transmitter is used by North Anna , to monitor pressurizer level. , ) January 1977 -- Westinghouse submitted , under NS-CE-1323, a report entitled " Sensor Qualification Program, IEEE 323-1971 Demonstration Program; Interim Test Report for the Barton/ITT Modified Pressure Transmitter." This transmitter is'used on North Anna to monitor wide range reactor coo.lant system pressure. March 1977 -- Westinghouse submitted, under NS-CE-1384, final' test reports for the above' mentioned Barton/ITT Modified Differential Pressure and Pressure Transmitters. March 1977 -- Westinghouse submitted WCAP-8937, " Environmental . l 4 i
.' l February 1976 -- Westinghouse submitted, as par t of the program, WCAP-8695, " General, Method of Developing Multi-Frequency,Bi-Ax.ial Test Inputs for Bistables."
This WCAP describes the method for developing test , j . inputs for the Westinghouse 7300 process bistables and > Nuclear Instrumentation System bistables applicable to' i i North Anna. , 1 ,, . March 1976 -- Westinghouse submitted, as part of the program, ! WCAP-8674, " Multi Frequency and Direction Seismic ~ ' [ Testing of Relays." This WCAP provides an . analytical evaluation of multi-frequency, multi-direction inputs " i to relays used in Westinghouse safety related electrical equipment such as the Westinghouse 7300 , Process Control Equipment. 1 March 1976 -- Virginia Electric Power Company formally i committed to Westinghouse program, described in , NS-CE-692, in response to Regulatory Positions 3.2 and 3.3. This commitment was made in the North Anna FSAR. ! March 1976 -- Westinghouse initiated supplemental multi- l l frequency biaxial seismic testing of equipment, . 2 including the Westinghouse 7300 process bistables and
- Nuclear Instrumentation System bistables. ;
June 1976 -- Westinghouse initiated supplemental environmental ! i testing of instruments to current Westinghouse , r equ ir ements .
- June 1976 -- Westinghouse met with NRC Staff to review and~
i discuss results of multi-frequency biaxial seismic j tests. 1 July 1976 -- Westinghouse submitted, under NS-CE-ll27, draft ; reports on the above-mentioned multi-frequency biaxial l seismic testing. ! August 1976 -- Westinghouse met with NRC Staff to review and - ! discuss results of environmental testing of instruments ; l completed at that time. September 1976 -- Virginia Electric Power Company submitted to ; NRC Staff a report entitled " Environmental Qualification of Westinghouse NSSS Scope ~ Safety-Related Instrumentation for North Anna Units 1 and 2." This-report was the subject of the following meeting. ] September 1976 -- Virginia Electric Power Company, Westinghouse f B __ _ _ ---.__. -. . . . . - . , =~- .-- ..
T -
- o. , . .
. b *e SER when referring to compliance with regulations. ,
He cited three examples: .
- 1. Containment heat removal system (SER S 6.2.2)
- 2. Containment Isolation system (SER S 6.2.4)
- 3. Combustible gas control system (SER S 6'.2.5)
Mr. Pollard discussed deviations of the conPainment isolation system criteria on another plant as supporting his view that regulations are being circumvented by the NRC Staff. ) The NRC Staff itself is the approp'riate,one to say what it means in the SER. Vepco's position is that North Anna 1 and 2 comply fully with the ' NRC regulations, and Vepco has supplied extensive information on the three systems listed above to demonstrate compliance. Here is a list of applicable ; t references: System Pr imary PSAR . Reference Additional FSAR Refers is Containment. Heat 6.2.2 Numerous responses in Removal System 3.1.34 thru 3.1.36 Supplement II, Volume II ! r Containment 6.2.4 S6.54 thru S6.58 ! Isolation System 3.1.47 thru S6.60 thru S6.64 ; i 3.1.50 S6.122 and S6.124 3A.ll ; Combustible Gas 6.2.5 S6.65 thru S6.68 ! . Control System 3.1.37 thru 3.1.43 3A.7 i l' l
Qualification-Instrument Temperature Transient Analysis." This report presents an -analysis investigating the thermal response of instrument transmitters to severe environmental temperature transients and demonstrates the conservatism of the 3estinghouse test conditions. Most of the above information has' been supplied by J. L. ;Vota,
~
with the help of W. R. Sugnet (both of Westinghouse) . The , following Stone & Webster personnel can also provide information in the indicated fields: John Freeman (Seismic) A. Murphy (Control Equipment Qualification) . ; s E. Benabourg (Electrical Equipment Qualification) J. Barnhout (Electrical Equipment Qualification) i
~
INTEGRITY OF STEAM GENERATOR TUBES t Mr. Pollard addresses the question of steam generator [ t tube integrity (Pollard Statement 12-14), which we have already j i discussed in our response to Mrs. Allen's limited appearance. ! We might add that the steam generator issue has been considered l [ by the Appeal Board in the Prairie Island 1 & 2 proceeding, In the Matter of Northern States Power Co. (Prairie Island Nuclear f Generating Plant, Units 1 & 2), 4 NRC 169 (1976). 'I t I CONFORMANCE TO THE INTENT OF THE REGULATIONS' ! Mr. Pollard expr'essed concern (Pollard Statement 14-16, Tr. 3022) over the NRC Staff's use of the word " intent" in'the f l [ t h i I I
Regulatory GuiGes are issued to describe and make available to the public methods acceptable to the NRC Staff of implementing' specific parts of the Commission's regulations, to . delineate techniques used by the staff in evaluating specific problems or postulated accidents, or to provide guidance to applicants. Regulatory Guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions different from those set out in the guides will be acceptable if they provide a basis: - for the findings requisite to the issuance or continuance of a permit or license by the Commission. - Vepco has in some cases used " methods and solutions. I different from those set out in the guides" and has documented those differences in the FSAR. The NRC Staff has explored l '} those differences with Vepco, as documented in. the FSAR. Above information provided by: W. H . House, II (Vepco) Heat Removal System - S. Frank (S&W) - Analysis l R. Bradbury (S&W) - Hardware Combustible Gas Control'- R. Bradbury (S&W) - Hardware C. Changelian (S&W) - Filters Containment Isolation - R. Bradbury (S&W) R. Bone (S&W) ,') REACTOR CCOLANT SYSTEM OVERPRESSURE PROTECTION Mr. Pollard says (Tr. 3016, lines 4-16) that there is no apparent reason why equipment and design changes to ' prevent overpressurization of the reactor coolant system should not be required at North. Anna before the plant begins operation. Mr. i k
g* %- a x . General-Design Criteria ~ s The regulation applicable to this discussion is 10 CFR Part 50, Appendix A, which contains the General Design Criteria. As' no ted above, the plain words of.c the Introduction to appendix A itself al[ow sem'e , latitude in the $ application of' s s' the criteria. In addition, 'some of the Criter.ia themselves g- ' allow for alternative m~ethcds of satisfying the objectives of
'the Criteria.
Criteria 55;and 56, for. example, include the s s s.' g s . . 1 words "unless it can be demonstrated that the containment isolation provisions for a' spec'ific class of lines, such as instrument lines, are : acceptable on some other defined basis."
>Vepco has supplied appropriate information and y' justification for all case's.where - '
the design of North Anna 1 ' s s . ~ y- and 2 di'f #er's from, the ' specifics of tha' General Design
.s -
,, Cri.teria. In each'such car 6 the design meets the fundamental concern em?codied "in the scriteria'hy an alternate defined basis, 3
s' ' ' e s ~, 30' 1 h the NRC. Staff has .revi'ewed and accepted. ~
~
sq - 1 7 is',, , .
,\
'Regulatorv Guides 3 ,
s V' .
\d .
-Py. Pollard frequently mentions failures to comply with
_,, Reg ulatory Gul6es . ' A Regulatory Guide, of course , is not a
' '~~
> t; _ regulation, as'each one makes clear on its face: g \_ i 5 ' -
\
g As. ' s ; - g +: 34 12 3 .
,, , g .
," g' s g
, 4
' \.
. , m
.k-_ ,
. 6 1- ,3 .1, _
e . ,- .. , During cold water solid conditions When the reactor coolant system is not in a hot , operational condition, on the other hand, - the allowable pressure limits are more restrictive. These pressure limits are dependent upon system temperature as related to re' actor vessel mdterial failure in the non-ductile mode and are i determined in accordance with 10 CFR Part 50, Appendix G. If ' the reactor coolant system is relatively c.old (less than or equal to 350 degrees F) and a suf ficient " cushion" is not 3 :
/
available in the RCS, there is no automatic method to relieve
- reactor coolant system pressure should a transient occur. The :
concern, then, is that overpressurization events might occur (and indeed have occurred) during cold water solid conditions, and appropriate means must be used to prevent such an event. In July 1976 Vepco attended a meeting with the NRC Staff at which the overpressurization issue was discussed. The Staff expressed their concern about the overpressurization ,) events that had occurred on many occasions and said that licensees would have to provide a method to prevent such events ; at their licensed facilities. The NRC issued a letter on August 11, 1976, that formally outlined the Staff's concerns ; l and required responsive action by Vepco. Vepco letter serial , no. 196/081176, dated September 7, 1976, to the NRC clearly outlines the Company's position on overpressurization. The
t - Pollard also implies' that administrative changes are inadequcte to prevent reactor coolant system overpressure.
~Vepco presented some testimony on this issue at the ;
hearing June 2,1977 (Tr. 3590-91). The following discussion
- i elaborates on that testimony. .
During ooeration North Anna is equipped with several systems designed to ; prevent reactor coolant system (RCS) overpressurization when the reactor coolant system is at operating pressure and temperature. One of these systems is the pressurizer, which . automatically maintains a relative.'" constant pressure by means of a steam bubble or " cushion" to absorb .the expansion of coolant. Should the pressurizer- be unable to maintain 'the proper pressure, two independent relief valve systems are ; available to protect the plant. One of these systems consists ! of two redundant power-operated relief valves that operate when system pressure reaches a given setpoint. Should these not be sufficient to mitigate the pressure increase, three j self-actuated code safety valves are available to relieve - system pressure. Thus the problem of RCS overpressurization - under normal ocerating conditions, where the' possibility and magnitude of overpressure is the most significant, has been r greatly mitigated by the station design. . l
. 1
}
minimize the equipment available that 'might -lead to an - overpressure event. Above information provided by: B. Ralph Sylvia (Vepco) LOCA EFFECTS ON FUEL AND PIPE BREAK SPECTRUM Mr. Pollard refers-(Tr. 3017) to Section 2.9.4 of the ] Safety Evaluation Report, which he says mentions "the - deformation between the spacing of the fuel (elements)" but then fails to discuss "whether or not there will be adequate i spacing between the elements." The statement to which Mr. j Pollard refers discusses some of the results of a scoping analysis that was performed to obtain an upper bound for the consequences of the loss-of-coolant (LOCA) ' accident. This scoping analysis, which assumed failure of all reactor vessel . supports, demonstrated that the integrity of essential structures, systems, components., and connected piping, and the functioning of the control rods, are assured, even under such ( extreme circumstances. When extended to the fuel assemblies, however, this analysis showed the margin between calculated impact forces on . l the fuel assembly spacer grids and the experimental crush l 1 I I S i
--m .
.._.,.,,,_w- __ , _.. , _ _ _ _ _ . . _ , . . . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _
letter states: "vepco is committed to providing physical protection against overpressure transients in its reactor vessels. We believe that hardware changes will be necessary to provide assurance that the integrity of a reactor vessel is safeguarded." Vepco has pursued this commitment actively since that time. vepco letters serial nos. 196A/081176, 196B/081176, and 005/010677 to the NRC outline the progress made on this project. In a letter dated Mai 5, 1977, the NRC Staff stated that the proposed overpressure protection system (OPS) to-be ( installed at Surry is conceptually satisfactory. The design of the overpressure protection system to be used at Surry is also applicable to North Anna, as Vepco letter serial no. 346, dated November 30, 1976, explains. Vepco continues to pursue development and procurement of the final design for the OPS. We have committed to install the system at Surry before the snd of 1977- and at North Anna
. Unit 1 before the startup following the first refueling. The OPS will be installed at North Anna Unit 2 during construction. ;
i In the meantime Vepco has instituted interim measures to reduce ' the chance of having an overpressure transient. At North Anna , an existing pressure-sensing system is temporarily reset when the plant is shut down; this provides low-pressure relief valve i protection for the cold water solid condition. Procedures are also in effect to minimize water solid operations and to ! t
(which are such things as actual results of grid strength tests and design stresses in fuel assembly components) are considered to be of competitive advantage to Westinghouse, and consequently Westinghouse requested that vepco ask that they be withheld. The information is available, under the rules of the Nuclear Regulatory Commission, for inspection by persons
" properly and directly concerned." Furthermore, the assumptions used in the analyses and the resulting final safety margins can be found in Appendix 5A to the FSAR.
In order to establish the pipe break with the most - severe potential consequences for the reactor vessel supports, a spectrum.of pipe ruptures was postulated during the preliminary analysis, including both horizontal splits and double-ended (guillotine) ruptures, on both the hot leg and cold leg of the reactor coolant loop. For the RPV support system, the maximum load occurs -for a guillotine break at the reactor pressure vessel inlet nozzle; thus this break was used i as the limiting case for further analysis of the support . system. As detailed in Appendix SA, similar limiting cases were used for the reactor internals and fuel assemblies. In every case, therefore, the most limiting pipe break was used to analyze the, effects of the postulated accident. This analysis, detailed in Appendix 5A, demonstrates that the integrity of the reactor vessel supports, fuel, and reactor internals will be I l l
~
strength to be smaller than the NRC Staff considered acceptable. To develop a more accurate idea of fuel assembly performance to quantify the actual margins available in the reactor vesscl supports, additional analyses were performed using the Westinghouse MULTIFLEX computer code. ~ The results of these analyses for the fuel elenents are documented in Section 5A.9 of the FSAR. Using the assumptions P and the version of the MULTIFLEX code approved by the NRC Staff, the analyses demonstrate that substantial margin exists between.the applied loads and the load at which the fuel grids subject to maximum loading would sustain measurable permanent , d e fo rma tion . (It should be noted that these analyses deal with safety margins of one particular component of the fuel , assemblies, .namely, the spacer grids; the actual deflections of the fuel elements themselves are small and have not been an item of concern.) These analyses demonstrate that no deformation of fuel element spacers can occur and that there is no impact on ECCS (emergency core cooling system) performance. Discussions of the assumptions and results for the analyses of the reactor vessel support, internals, and fuel can be found in Appendix S A to the FSAR. In a few cases, however, Vepco requested that some numerical values in these discussions be withheld from public disclosure as proprietary in accordance with the Commission's regulations (10 CFR 2.790) . These values
~.
overpressurization whild the system is in operation is provided by the ASME Code relief valve; which is located in the discharge side of the system. This Code relief valve, which has a set pressure of 600 psig and a capacity of 1133 gym, is capable of maintaining the RHR system pressure within ASME Code limits for any credible overpressurization transient. > The most limiting potential causes of RHR overpressurization can occur only when the RER system is in , operation and open to the Reactor Coolant System (RCS).
- Ther e for e , although the Code relief valve provides adequate RER overpressure protection by itself, additional redundant RER overpressure protection will be provided by the system that is to be adopted for RCS overpressure protection (see above) . .
The " prevent open" and " auto closure" interlocks on the RHR suction isolation valves, on the other_ hand, are not intended to provide conventional overpressure protection of the RHR system (since this is fully provided by the Code relief ) valve discussed above) but are provided only as backups to the plant operator, who is required by procedures not to open the RHR suction isolation valves if RCS pressure is greater than 425 psig an.d to close them before increasing RCS pressure above 4 50 psig . Since two suction isolation valves are provided in ! series, the design purpose of the " prevent open" and " auto l closure" interlocks is to prevent a situation f rom occurr ing ; [ ) j
. -. ~ ' . . maintained. A tabulation of limiting break locations for the entire reactor coolant system can be found in the response to , NRC Staff Comment 5.62 in the FSAR. This table gives further , information regarding the spectrum of breaks considered for various analyses. The NBC Staff, on page 3-17 of the Safety Evaluation Report, state that they have found the reactor. vessel support system acceptable and note that the particular pipe break r
, analyzed is of low probability. This should not be interpreted as indicating that the consequences of such a pipe break would i be unacceptable. On the contrary, the pipe break loading leading to the most severe potential consequences has been analyzed to demonstrate the acceptability of the system.
Above information provided by: , J. L. Vota (Westinghouse) RHR SYSTEM OVERPRESSURE PROTECTION Mr. Pollard finds f ault with the NRC Staff's treatment of overpres.sure protection for the Residual Heat Removal (RHR) i System (Tr. 3019). Protection of the RHR System against potential , 9
cperation and for prevention of exposure of the RHR system to normal RCS pressure), gross' rupture of the RER system 'is not a credible event. If leakage from the RHR system should occur dur ing RHR system operation, isolation could be provided by the RHR suction isolation valves, which are designed to operate within two minutes against a seat differential pressure of 700 p s ig . Contrary to what Mr. Pollard seems to think (Tr. 3019, lines 9-12), the North Anna RHR system is located inside the reactor containment, and thus an RHR rupture, even if it could occur , would only occur inside the containment. Also, the RER system i*s not part of the Emergency Core Cooling System (ECCS), ad so its failurv cotcld in no way affect the post'-LOCA- - .- performance of the ECCS. Above information provided by: J. L. Vota (Westinghouse) ) SAFETY RELATED DISPLAY INSTRUMENTATION . Mr. . Pollard criticizes the NRC Staff's discussion of safety-celated display instrumentation (Tr. 3020, lines 2-23). His chief criticism seems to be that the Staff have not specified the criteria for judging what instruments are needed.
~80-
where two operator errors or one operator error plus one single failure could result in the opening of the low pressure RHR system to the high normal operating pressure of the RCS. To achieve this protection, each of the two suction i isolation valves has both a " prevent open" interlock and an .
" auto closure" interlock, which are separate, independent and diverse f rom the " prevent open" interlock and " auto closure" interlock of the _other suction isolation valve. Each " prevent o pen" interlock prevents its associated suction isolation valve f rom being opened if RCS pressure is greater than 425 psig. ,
Euch " auto closure" interlock automatically closes its associated suction isolation valve if RCS pressure reachas 750 psig during a reactor startup and the valve has not already been closed by the plant operator. ; Diversity between the two " preven.t open" interlocks and between the two " auto closure" interlocks is achieved by using a Foxboro pressure sensor (which operates on the force balance principle) for one channel (one suction isolation valve) and a Rosemount pressure sensor (which operates on the capacitance r principle) for the other channel (the other suction isolation valve). , , In view of the overpressure protection provided the RER system (both for conventional overpressure transients during
, g,l, UNITED STATES [ . .
p*g NUCLEAR REGULATORY COMMISslON ! , a
< 0 WASHINGTON. D. C. 20S55
;% yE
- t. Y
'% ,,,,+j September 2, 1977 r
Frederic J. Coufal, Esq., Chairman Dr. Paul W. Purdom, Director q Atomic Safety and Licensing Board Environmental Studies Institute , p U.S. fluclear Regulatory Commission Drexel University Q Washington, D. C. 20555 32nd & Chestnut Streets .3 , Philadelphia, Pennsylvania 19104 jS Mr. R. Briggs 110 Evans Lane d j) Oak Ridge, Tennessee 37830 . i.t_. In the Matter of Virginia Electric and Power Company . (florth Anna fluclear Power Station, Units 1 and 2) 3
- Docket N.; . 50-338 OL and 50-339 OL ;
Gentlemen: Enclosed are the following documents related to the North Anna 1&2 operating license proceeding: (1) Affidavit of Alexander W. Dromerick, in respense to a limited appearance of Robert Pollard; il79 (2) Affidavit of James W. Shapaker, in response to Mr. . ce,,, Pollard's limited appearance statement; & *$"'c 4 (3)' HRC Actions On GA0 Recommendations Regarding the ! NOV3 01978 > E Investigation of Allegations of Poor Construction U e/"- .% - Practices at the florth Anna fluclear Power Plants, s U'"* with cover letters, dated August 16, 1977, to the 4 y respective Chairmen of the U.S. Senate Committee g > on Governmental Affairs and the U.S. House of Representatives Committee on Government Operations signed by Lee V. Gossick, Executive Director for Operations (the flRC response to the GA0 Report on florth Anna) l (4) Letter f' rom R. F. Fraley, Executive Director for N iie the Advisory Committee on Reactor Safeguards, to 67: Mrs. June Allen, dated July 28, 1977 (the July 20, iy 1977 memorandum from R. F. Fraley to Lee Gossick, bl which was attached to the original July 28 letter, was submitted to the Board by letter, dated August 10, M% ! 1977, and is thus not included in this submittal); d u 1 1 1____________________________-_______________--__-.____-_.._________ _ _ _ . _ _ _ _ _ _ _ . . _ . - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - . - _ - _ - _ _ _ _ _ _ _ _ _ ,
e'
?
Due to the sig6ificant variations in nuclear power station systems and design , it would be impractical to have a single criterion specifying the exact numbers and types of safety-related display instruments to be provided. This must. be determined instead on a case-by-case basis for each - individual station; this was done by Vepco for the North Anna Power Station, taking into account the NRC Staff's comments and concerns. Indeed, concerns about proper and sufficient display
; instrumentation have received special attention by both Vepco and the Staff. The subject has been reviewed and discussed at several meetings with the Staff, and a presentation was made to the ACRS in August 1976.
The North Anna Power Station is provided with safety-related display instrumentation of the proper parameters and in sufficient cuantity to ensure that the operator can follow the course of an accident and determine the , effectiveness of safety systems. The following information on this subject has been previously submitted: 4
- 1. North Anna FSAR:
Section 7.5 Safety Related Display Instrumentaion Section 11.4 Radiation monitoring ; 4 Section 13A Emergency Plan '
- 2. Document No. CR9362 -- Transcript of the proceedings of the ACRS on August 11, 1976, pages 248 through 262.
f
* * 'f ' * / ),] f),I,, e lu) $ { }. 0w s '. c
~?!, ,, ;., . . ;.s !o w N :~ !~"'
I. . y , r?,,, /C) havn ~ ;; ~lNp j)b lpv ,y
t
~
results of the earlier investigation are reported in I&E Report Nos. 50-338/77-14 and 50-339/77-11 referenced- in Staff Exhibit 6A. As indicated by I&E Witnesses in testimony during the recent hearings, I&E confirmed that the welds were made by qualified welders using properly qualified weld procedures, but that a non-conformance occurred when a welde.r p failed to properly identify a weld he had performed (see Tr. 3424-25; , m I&E Testimony, supra, at page 147). The NRC did not require these j specific welds to have the welders' identification number applied and tj did not consider the identification of the weld to be the type of inci- g dent which could affect the public health and safety. g ti
. i/d (3) Allegation of Poor Quality Control Regarding Welding p_
Progra n In a letter, dat'ed June 22, 1977, counsel for the NRC Staff informed the Board about another allegation of poor quality control at North Anna'. . 3 ~ I&E contacted the alleger, who claimed that the manager of the welding ) department and one supervisor are not qualified to manage the North Anna ' welding program, and that if changes were not made, cost and quality of the welding would be adversely affected. I&E investigators did not find any evidence to substantiate the allegation that the management of the welding program was unqualified. I&E investigators indicated to the alleger that.the impact of the alleged improper welds on the cost of the construction of the facility was not a safety concern that the NRC had jurisdiction over. As indicated at the recent hearing, I&E has con-cluded an intensive investigation of the overall welding program at North Anna, and concluded that'certain management and quality assurance controls were deficient, but that none were of serious safety concern (see I&E Testimony, supra,.at pages 136-146). The alleger also claimed that gang box meeting and training programs, which were initiated by yEPC0 as part of a commitment to the NRC, were no longer held on a regularly scheduled basis, and that the welding _T manager had someone forge his signature in the attendance records of / these meetings. I&E determined that there was no evidence to confirm the allegations and that the meetings were held on a regularly scheduled basis and constituted a viable program to m'eet the needs of the crafts-men. I&E also determined that the attendance records were not intention-ally falsified. ,, p The alleger finally claimed that, as a result of a lack of proper manage-ment, the weld.ing rejection rate has increased by a significant amount 4 in the last few months. I&E informed the alleger that there was no 9 evidence to substantiate the allegation that the weld rejection rate was , K increasing. 8 N d
(5) NRC Staff's Proposed Transcript Corrections (covering evidentiary hearings held on November 30 through December 3, 1976 and May 31 through June 2, 1977); and (6) Supplement No. 7 to the NRC Staff's Safety Evaluation - Report related to the operation of North Anna Nuclear Power Station, Units 1 and 2. ;
~
The Office of Inspection & Enforcement (I&E) has completed investiga- h tions of allegations made during, and subsequent to, the evidentiary hearing which concluded on June 4, 1977. The results of this effort are - as follows: (1) Allegations by Frank D. James in Limited Appearance ' Statement (Tr. 3563-67) Mr. James alleged that, while employed as a painter at the North Anna site, he carried a large bag into the site several times without being searched, and that he was not required to surrender his identity card ' when he was fired from his job at the site. The NRC does not consider access to a nuclear power plant during the construction phase to be a serious safety consideration, and does not have any requirements re-lating to physical security at plants under construction. Mr. James would not be able to gain access to the operational areas of +'.ie facility using his construction identification badge, once an operating license . is issued for North Anna 1 and 2. Mr. James also alleged that workers slept on the job, drank alcohol, smoked marijuana, and generally avoided work. This allegation also does not relate to matters of plant safety and, since the alleged activities do not affect the public health and '
- safety, the NRC has no jurisdiction over this matter. Accordingly, I&E i . did not conduct a fonnal investigation of the allegations.
(2) Allegation Identified by VEPC0 Counsel to Hearing Board During the Evidentiary Hearing On June 2,1977, counsel for VEPC0 informed the Board of an allegation p-by a welder at the North Anna site that he was instructed to perform a weld incorrectly, and to put a stamp on that weld (Tr.3576). I&E com- q'5 pleted ari investigation of that allegation and determined that this % allegation was related to the previously investigated allegations re- C6 garding the proper identification of welders to the welds they had per- . formed (see Supplemental Testimony of I&E on Board Questions 1 and 2 9,W (hereinafterI&ETestimony),followingTr.3037,atpage147). The
1 1 l cc w/ enclosures: cc w/ Affidavits Only: Michael W. Maupin, Esq.* R. Pollard Anthony Gambardella, Esq. Mrs. June Allen * - Mrs. Margaret Dietrich* Mr. Dean P. Agee i, Dr. Kenneth A. McCollom William H. Rodgers, Jr., Esq.* j John J. Runzer, Esq.* , p
~
Mr. James M. Torson* Richard Foster, Esq.* - . Mr. Bradford Whitman " t Mrs. James C. Arnold
- Mr. William Warren Atomic Safety and Licensing Board Panel Atomic Safety and Licensing ') /
Appeal Panel Docketing and Service Section . Those persons indicated with a (*) have previously been sent a copy of Supplement 7 to the SER, and are not being sent a copy with this letter.
=
y. g g
. k d
m H t--. 9
4 4 The I&E investigation of these allegations related to the welding pro-gram management is documented in I&E Report No. 50-339/77-23. A copy of this .repo. t is in the Public Document Room, and car. be supplied to the Board or any of the parties if it is so desired. (4) Allegations Related to Falsification of Records of Test Instrument Calibrations y A final set of allegations was made regarding the alleged falsification of documentation by the field quality control regarding calibration of Q test instruments. I&E conducted an investigation.into the allegations, a and determined that the allegations were not substantiated. I&E deter- )~ mined that the alleger admitted that quality control did not actually encourage him to falsify documents, but merely put pressure on him to get the instruments calibrated. I&E further determined that, although ~ certain calibers were documented as calibrated against a standard which - - was not at the facility, the calibers were within calibration, and therefore no safety significance could be attached to the incorrect record. Various other pieces of equipment, which were alleged to be improperlv calibrated, were found to be within the proper calibration levels. I&E did determine, as a result of its investigation, that certain records regarding calibrations were back-dated in an attempt to conceal what was though to be an overdue calibration. I&E found that the calibrations in question were not overdue, and that the back-dating was ,not o# safety significance to the facility. I&E understands that the individual who back-dated the records has had his employment terminated with Stone & Webster. I&E also found one instance of improper follow-up of out-of-tolerance reports, but that this item had been previously identified in - an audit performed by Stone & Webster in Feoruary, 1977. Adequate corrective action had been taken and no safety problem was identified regarding this item. I&E reported the results of its investigation in Report'No. 50-338/77-
- 28. A copy of this r. port is in the Public Document Room and can be supplied to the Board or any of the parties if it is so desired.
- c. Sincerely, b 0~.
L1
- Daniel T. Swanson a 3
Counsel for NRC Staff " d(%
Enclosures:
As stated ~. l R cc: See page 5 L l r -
I
~
- l RESPONSE TO R. POLLARD'S LIMITED APPEARANCE STATEMENT
- l This response addresses certain questions raised in a limited appearance statement by Mr. Robert D. Pollard (Tr. 3010) .
The Board requested that in regard to Mr. Pollard's statement, the staff list the parts of the plant that do not comply with the rules and regulations ,,_ of the Commission and state why they need not comply (Tr. 3600). , in Section 22.0 of the Safety and Evaluation Report we state that based on u'l our analyses of the design of the North Anna Power Station, Units 1 and 2, j we have determined that upon favorable resolution of the outstanding matters [ set forth in Section 1.7 of that report (as modified in Supplements 1 through ;j 7), we will be able to conclude that: M
. M
" Construction of the North Anna Power Station Units 1 and 2 (the facility) has proceeded and there is reasonable assurance that it will be substantially completed, in conformity with s
- Provisional Construction Permit Nos. CPPR-77 and CPPR-78, the I application as amended, the provisions of the ACT, and the rules and regulations of the Commission".
The staff had the cogni:: ant Branch Chiefs (see list of names attached) responsible for the technical areas addressed in Mr. Pollard's statement review the respective sections of the Safety Evaluation Report and its Supplements that they prepared, The purpose of this review was to determine whether the parts of the North Anna Power Station, Units 1 and 2 discussed by Mr. Pollard still meet the NRC rules and regulations. Each of the Branch Chiefs has, on the bases of his review, reconfirmed that the North Anna Power Station Unita.1 and 2 satisfy the rules and regulations of the Commission. Mr.' Pollard also contends that stating that a system design conforms to the 3
" intent" of the regulations leaves doubt as to whether the specific require- )
ments of the regulation have been met. Th word " intent" was used in concluding statements in Sections 6.2.2, 6.2.4, and 6.2.5 of the Safety Evaluation Report, pertaining to the containment heat removal system, j containment isolation system and the combustible gas control system,
~
l ' respectively. In the enclosed Affidavit of James Shapaker, the staff l concludes that, with respect to the three areas mentioned, the respective 7 i rules and regulations are complied with. ' l . Q l In conclusion, with the exception of the recirculation spray pumps, which . C are under re-review by the staff and the applicant, and subject to the l, satisfactory resolution of the outstanding item listed in Supplement No. 7 , to the Safety Evaluation Report, the North Anna Power Station, Units 1 and 2 i has been constructed in accordance with the Commission's regulations.
,, ._ .. l *- UllITED STATES OF AMER 1CA l g ,
NUCLEAR REGULATORY C0iMISSION g p. E BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
- e VIRGINIA ELECTRIC AND Docket Hos. 50-338 OL POWER COMPAllY 50-399 OL (North Anna Power Station, Units 1 and 2) )
AFFIDAVIT OF ALEXANDER W. DROMERICK , i I, Alexander W. Dromerick, having been duly sworn, state that: , I am the Project Manager for the North Anna Power Station, Units 1 and 2 of the Division of Project Management in the Office of Nuclear Reactor Regulation with the U. S. Nuclear Regulatory Commission. A copy of my professional qualifications was received into evidence following Tr. 3037. I authored the attached document in response to Mr. Pollard's limited appearance statement. This document addresses Mr. Pollard's concerns regarding the North Anna Power Station, Units 1 and 2 meeting the Commission's regulations. I believe the attached document to be accurate r~ to the best of my knowledge and belief. ,]
, N
# f W!. do '
Alexander W. Dromerick d Sworn to and subscribe'd before 3 6"p./f77
~
me this J / '*- ~ bed., $ Nothry Public 7
%} e My Commission expires: 34 /7 7#
7 /
UllITED STATES OF AMERICA A'l1 '. .' NUCLEAR REGULATORY COMMISSION g *!Ei"c'
~ 1
~ "
NOV3 01978 > BEFORE THE ATOMIC SAFETY AND LICENSIt;G BOARD [ e,,,,,,,,,, 3
- -~ g; y .
- In the Matter of og J r VIRGINIA ELECTRIC AND POWER Docket Nos. 50-338 OL 4
- 3 COMPANY 50-339 OL -
[d ri (North Anna Power Station, d Units 1 and 2) ) l,t m 4 2. AFFIDAVIT OF JAMES W. SHAPAKER I, James W. Shapaker, having been fully sworn, state that: - I am a Section Leader in the Containment Systems Branch of the Division of Systems Safety in the Office of Nuclear Reactor Regulation with the U. S. Nuclear Regulatory Commission. A copy of my professional qualif f-cations is attached. I authored the attached document in response to Mr. Pollard's limited appearance statement, and believe it to be accurate to the best of my knowledge and belief. This document addresses Mr. Pollard's concerns regarding the North Anna Power Station, Units 1 and 2 meeting the Com-mission's regulatio. . r-
" k" ;3 James W. Shapaker g
Sworn to and subscribe h efore me this 3 /2 7-ff77
- d ig h
nk/If L&
~
Nota h Public // i My Commission expires: -
// /I
,, - . i.
LIST OF BRANCH CHIEF NAMES Names Branch I D V. Benaroya Auxiliary Systems ! I R. Bosnak Mechanical Engineering 5 ' D. Bunch Accident Analysis [ P. Check Core Performance . J. Collins Effluent Treatment System ( T. Ippolito Instrumentation and Control Systems T. Novak Reactor Systems S. Pawlicki Materials Engineering I. Sihweil Structural Engineering i ( w L 1 i i
testing. The system design satisfies the redundancy requirements for continued operability, assuming a single failure, and analysis has been presented to show that the system is capable of preventing the hydrogen W concentration from exceeding prescribed limits following a loss of cool- Ni l 4 ' ant accident. Test results have been presented which demonstrate the Id functional capability of the hydrogen recombiner. Design provisions h l y$ have been made to permit periodic surveillance roid testing of essential , Q system components and equipment. With respect to the containment isolation system, most of the contain- ]) ment isolation arrangements conform to the explicit requirements of the present General Design Criteria; i.e., General Design Criteria 55, 56 and 57. Of the remaining lines penetrating the containment that do not meet the explicit requirements of the General Design Criteria, those governed by General Design Criteria 55 and 56 were found acceptabic on some other defined basis, as allowed by General Design Criteria 55 and
- 56. Standard Review Plant 6.2.4, Containment Isolation System, contains guidelines for acceptable alternate containment isolation provisions for ,
)
certain lines; e.g., instrument lines and lines that are safety rela ted, regarding the number of valves, valve location and method of valve actuation. All other lines which do not meet the explicit requirements r -- are governed by General Design Criterion 57. As to these, a departure N
$m from the explicit reqtrirements of Criterion 57 was justified pursuant to the Introduction to Appendix A of 10 CFR Part 50. The bases for our jjW ki conclusions are discussed below. a"
- - - - . ,v. _-. ,&
. t RESP 0 rise TO R. POLLARD'S LIMITED APPEARAtiCE STATEMEllT In a limited appearance statement, R. Pollard contended that the state-ment that a system design conforms to the " intent" of the regulations F' leaves doubt as to whether the specific requirements of the regulations have been met (Tr. 3010 et seq.). This response addresses Mr. Pollard's ,
concerns. , b
~
The word " intent" was used in concluding staterr.ents in Sections 6.2.2, - 6.2.4, and 6.2.5 of the Safety Evaluation Report, pertaining to the con- - tainment heat removal system, containment isolation system and the com-bustible gas control system, respectively. The Staff used the expression to indicate that there could be several ways that the criteHa could be met. In the three cases identified above, the general design criteria are satisfied. The bases for the conclusion for each of the three , sections stated above are discussed below. The containment heat removal systems satiisfy the design requirements of General Design Criteria 38, 39 and 40 for performance, inspection and testing. The design provides assurance of continued operability under postulated accident conditions, and that the performance requirements for containment depressurization will be satisfied, assuming a single r failure. Design provisions have been made to permit periodic surveil- 9
- 4 lance and testing of essential system components and equipment.
fy a The combustible gas control system satisfies the design requirements of General Design Criteria 41, 42 and 43 for performance, inspection and {
defined basis from the requirements of General Design Criteria 56 regard-ing the number of isolation valves and the actuation provisions for them. I I The four containment leakage monitoring lines and the two containment ;y F vacuum pump suction line.s each contain two automatic isolation valves in [ g series outside containment. The containment leakage monitoring lines cannot have isolation valving inside containment since the containment pressure senscrs, used to actuate the engineered safety features in the event of an accident, are located outside containment between the auto- ) matic isolation valves and the containment Because of the safety function the pressure sensors must perform, they are located outside containment to prntect them from accident environments. With respect to the containment vacuum pump suction lines, both isolation valves are located outside containment to improve their reliability following an accident. These lines have a post-accident safety function; i.e., they are the suction lines for the combustible gas control system. We, there-fore, concluded that the isolation provisions for these lines provide an acceptable other defined basis from the requirements of General Design j Criterion 56 regarding the location of the isolation valves. F The 1/8-inch diameter line leading to the pressurizer dead weight pres- p' c sure calibrator contains two normally closed, administrative 1y controlled, j M . manual isolation valves in series outside containment. This line is an . O instrument line that is only used when the plant is shutdown; it has no E L
c Fourteen lines penetrating the containment have isolation valves which - are not automatically closed following an accident but rather are opened or remain open to accomplish post-accident safety functions. In each I case, however, redundarrt valving is provided to permit isolation of the lines, if necessary. The safety injection system, quench spray system
- and recirculation spray system are in this category. These closed :
safety grade systems also provide containment isolation and make it . unnecessary to automatically isolate the lines. We, therefore, con-cluded that the isolation provisions for these lines provide an accept- .. able other defined basis from the automatic isolation requirements of General Design Criteria 55 and 56. The recirculation spray system and the safety injection system each have two lines penetrating the containment with only one isolation valve in cach line located outside containment. These lines, which are th'e sump suction lines for these systems, have post-accident safety functions and must be opened following an accident. System reliability is improved with only one isolation valve in these lines. Furthermore, the valve is located outside containment since it would be under water following a pipe break accident if located inside containment. Containment iso-lation is also provided by the closed, safety grade piping systems out- i-D side containment; i.e., these systems constitute a second isolation [ q barrier in the absence' of a second isolation valve. Consequently, 3 double barrier protection is maintained. We, therefore, concluded that q the isolation provisions for these lines provide an acceptable other b i
e e' , I even though they will close rapidly upon loss of flow. Therefore, we ; have concluded that the check valves are suitable containment isolation valves for this service. In addition, other valves are available in the I system piping upstream of the check valves to further isolate these [ p lines; i.e., manual shutoff valves and remote manual power operated i valyes. The remaining six lines are associated with the secondary plant system; i.e., the three auxiliary and three main feedwater lines. The auxiliary
~
feedwater lines join the main feedwater lines outside containment but dowristream of the main feedwater line isolation valves. The auxiliary feedwater lines are of seismic Category I design and contain. remote manual valves which can be used as backup isolation valves. The main feedwater lines contain safety feature grade flow control valves and isolation valves, upstream of the check valves. These valves are fast acting (5 second closure times) and automatically close; they also pro-vide backup isolation capability.
\
Because of the importance of assuring secondary plant system integrity inside containment S the event of a LOCA, the system is seismically designed and analyzed to assure that provisions are made for adequate r~ pipe whip and missile protection. Consequently, rupture of the secondary 5 plant system is not po.stulated to occur either concurrent with or as a 6] C result of a LOCA. With the secondary plant system design providing , t assurance of system integrity, automatic isolation of the lines is not ]%
=.
safety function. The isolation provisions for this line were accepted on the basis of the recommendations in Regulatory Guide 1.11, Instrument Lines Penetrating Primary Reactor Containment. I *
. 1 There are eleven lines penetrating containment that are governed by 4 r.: ;
d ; General Design Criterion 57. They are asso.ciated with the secondary 4 ; W system and the component cooling water system, and have closed safety-
~
grade systems inside containment and single check valves in the lines outside containment. The containment isolation arrangements for these i lines differ from the explicit requirements of General Design Criterion 57 with respect to the method of valve actuation; i.e., simple check valves are used which rely on reverse flow of the process fluid to close the valve. However, we have found the use of the check valves to be acceptable since there are other system design considerations, discussed below, which justify a departure from the explicit requirements of General Design Criterion 57 for these particular systems. Five of the eleven lines in question are associated with the component i cooling water system, which is a closed, water filled, intermediate cooling system; i.e., the supply lines to the resid.:1 heat removal heat exchangers, the~ excess letdown heat exchanger and the coils of the 7 containment air recirculation' coolers. Since the system is closed outside containment, and seismically designed, automatic isolation of )j' q the lines is not necessary following a LOCA; i.e., the simple check , 1:.
- t valves are not relied upon to function as automatic isolation valves, t J
L
JAMES WILLIAM SHAPAKER PROFESSIONAL QUALIFICATIONS CONTAINMENT SYSTEMS BRANCH 0FFICE OF NUCLEAR REACTOR REGULATION I am a Section Leader in the Containment Systems Branch, Office of ( N Huclear Reactor Regulation. In this position I am responsible for the M r technical supervision and review of the activities of personnel in a section of the Containment Systems Branch. I supervise the safety review of associated containment systems and features of the proposed . design, and operating procedures, to assure maximum safety to the health
~
]
and safety of the general public. I graduated from Marquette University in 1963 with a Bachelor of Mechanical Engineering degree. I then acepted a position with the Ato:nic Energy Commission in the Division of Reactor Licensing. I attended the Oak Ridge School of Reactor Technology at the Oak Ridge National Laboratory in Tennessee during the 1963-64 term. Upon returning to the Atomic Energy Commission in 1964, my responsibilities included the review and evaluation of license applications for university,
)
industrial and government-owned research and test reactor facilities. My responsibilities wlso included the review and evaluation of the safety aspects of proposed experiments and changes to the technical - E specifications for operating research and test reactors. }n b In 1972 when the Atomic ' Energy Comission was reorganized I was assigned 4 9 2 to the newly formed Containment Systems Branch as a Systems Analysc. My d responsibilities in this position, and as a Systems Engineer and Senior l
b g .0 - , t; - 1, s ,
~
necessary from.a containment isolatidn standpoint following a LOCA. We have, therefore, concluded that' the check valves are acceptable contain-me[t isolation valves for this service, 'and that the power operated .. s- valves in the respective lines, provide adequate backup isolation capa- ! t bility. Based on the foregoing disc 3tsion of the isolation provisions for lines __ in closed systeras penetratingJthe containment; we have concluded that < adequate containment isolation capability exists for the component cool- . . ing water supply lines and secondary plant system lir[s. Basically, the simple check valves, which are not relied upon to function as automatic isolation valves, in conjt.nction with other favor:ble aspects of the ~ systemdesigns,formlhebasisforourconclusion. Furthermore, we have 4 \ '? concluded that even though the isolation provisians differ from the
- explicit requirements of General Design Criterion 57, a departure from ,
Criterion 57 is justifi6d in accordance with the Introduction to Appendix s , A to 10 CFR Part 50. *de ' therefore, concidde that the containment iso ' - lation prt,visient for these lines provide isolation. protection equi-
~ .
~
valent to that provided by systems designed to meet the explicit' require-ments of General Design Criterion 57. F-o k 0 I e
*- d 4
A _>
. agg . .
t 3 . _ . . . . . .- NOV3 01978 > --
- r. -
W
- '~
E c; " ; = C. O' s-a- i '
& ? '
on & L ' UNION OF CONCERNED SCIENTISTS *I# 1025 15th Street, N.W. Washington, D.C. 20005 September 25, 1977 Frederic J. Coufal, Esq., Chairman Dr. Paul W. Purdom, Director Atomic Safety and Licensing Board Environmental Studies Institute U.S. Nuclear Regulatory Commission Drexel University Washington, D.C. 20555 32nd 6 Chestnut Streets Philadelphia, Pennsylvania 19104 Mr. R.B. Briggs 110 Evans Lane Oak Ridge, Tennessee _ 37830 In the Matter of Virginia Electric and Power Company ., (North Anna Nuclear Power Station, Units 1 and 2) ',' I Docket Nos. 50-338 OL and 50-339 OL Gentlemen: On May 31, 1977, I made a limited appearance in the subject proceeding (Tr. 3010). The Board requested the applicant and the staff to respond (Tr.3600). The responses to the Board were sent by the applicant's letter of July 5, 1977, and the staff's letter of September 2, 1977. I am writing to provide comments on those responses. In addition, I am taking this opportunity to provide another internal NRC memorandum that identifies another hazard in the North Anna design. The responses from the applic' ant and the staff reinforce my conclusions that the evidentiary record is incomplete and that the subjects I identified deserve additional inquiry by the Board. The Affidavit of ,, Alexander W. Dromerick is devoid of substantive information and the info' that is provided is inaccurate. The Affidavit of James W. Shapaker acknt jtio ledges that the design of North Anna Units 1 and 2 does not meet the specific requirements of the Commission's regulations and advances disingenuous reasons for this being acceptable. Detailed comments on these staff responses and examples of the misleading nature of the applicant's response are attached (Enclosure 1). I am also providing a copy of an August 18, 1977.menorandum written by Stephen H. Hanauer, Technical Advisor to the Executive Director for Operations of NRC (Enclosure 2). This document was apparently.not intended to be made available to the public and I do not know who provided my copy. The Board will note that Dr. Hanauer believes that the Westinghouse design is unsafe. In addition to identifying a " design defect" in Westinghouse plants, the memorandum is significant for other reasons. 1208 Massachusetts Avenue . Cambridge. Massachusetts 02133 . Telephone (617) 547-5552
Systems Engineer, including the review and evaluation of the safety ' I aspects of containment systems for commercial nuclear power plants. I --- 1 l was responsible for the review of the construction permit or operating j licensing applications for the following nuclear power plants: Indian f L Point 3, Rancho Seco, Arkansas 1, Greenwood 2 and 3, Fort Calhoun 1, [ Davis Besse 1, Beaver Valley 1, Beaver Valley 2, and Duke Project 81. - - In addition to the above plant assignments, I was assigned to draft the , Stand Review Plan for the Containment Systems Branch. I have also
- t. . .
served on committees engaged in developing standards'for the nuclear industry. In 1974 I was selected to be a Section Leader in the Containment Systems Branch. O
\
?
6 9
d Enclosure 1 COMMENTS ON STAFF AND APPLICANT RESPONSES The staff's response is contained in the affi' davits.of. Alexander - i W. Dromerick and James W. Shapaker. Mr. Dromerick provides no substantive information and the information he does provide is inaccurate. He states, ' in part, that "the cognizant Branch Chiefs . . . . review (ed) the respective sections of the Safety Evaluation Report and its Supplenents that they r '3 , prepared." (emphasis added) Knowing that Branch Chiefs do not prepare SERs, I telephoned Mr. Dromerick. He stated that he meant that the identified Branch Chiefs supervise the technical experts who reviewed the design and prepared the SER and its Supplements. Thus, Mr. Dromerick's affidavit would have been equally accurate or, depending on one's point of view, inaccurate had he stated that the Director of Regulation prepared-the SER and its supplements. The point'is that the people best qualified to state whether the North Anna plants meet the Commission's regulations.are
)
the technical experts who reviewed the application and prepared.the SER and its Supplements. I suspect the reason management officials, i.e., Branch Chiefs, rather than reviewers,were asked to respond to the' Board, is the Board's statement that ".....if additional hearings were held ' concerning these matters, we would expect that these people would very likely be witnesses in any such hearing." (Tr.3601). It would certainly be an unmitigated disaster, from NRC management's point of view, if the 4 technical experts who actually reviewed the North Anna design were, under - the pressure and protection of testifying under oath, to actually tell the whole truth. On May 31, 1977, I discussed the reasons reviewers are unable to do this when preparing the SER.
1 i l l According to another internal NRC document I have, interaction l between control systems and protection systems is currently classified by the staff as a Category A generic problem. Staff testimony in this pro-cooding concerning " generic items" can be characterized as: the plant is
- safe and generic items are being studied only to confirm and quantify the safety margins. Such testimony is directly contradicted by Dr. Hanauer's-memorandum and the other previously secret NRC documents I provided. In the face of these documents, it seems not only aapropriate but also neces-sary for the Board to question the veracity of the staff's testimony.
Dr. Hanauer's memorandum is also significant in light of the many pronouncements made by NRC officials within the last year or so. According to this litany, staff members are obligated to bring forth matters of safety concern and disagreements with official positions and are free to fulfill these obligations withcot fear of reprisal. As noted above, the staff's testimony does not even hint at the existence of staff views that the North Anna design is unsafe. Furthermore, the fact that Dr. f Tauer's memorandum was provided anonymously indicates that someone
- ares his views, believes the public should be informed, and is in fear of reprisal if identified. i I recognize that the preceding discu'ssion may be of limited interest to the Board in view of its statement that " investigation of the internal operations of the Nuclear Regulatory Commission (are) heyond the scope of this proceeding. . ." (Tr. 3 600) . However, I have not asked the Board to investigate NRC's internal operations, per se. Rather I recom-mended that the Board conduct additional inquiry in order to have a complete evidentiary record. I have discussed NRC's internal operations and provided internal NRC documents to illustrate why I believe such inquiry is needed.
I made a limited appearance in this proceeding as part of my efforts to have the current deficiencies in the licensing process corrected. If the Board decides that it need not conduct some kind of independent i yestigation, it would be helpful if the bases for that decision were fted. With such information available I might be more successful the (n 6xt time I try to have a board consider all the facts. Sine rely, W Robert D. Pollard Union of Concerned Scientists 1025 15th Street, N.W. Washington, D.C. 20005 (202) 347-5800
Enclosures:
(1) Comments on Staff and Applicant Responses l (2) Memorandum from Stephen H. Hanauer to E.G. Case, ; August 18, 1977 ! cc w/ enc 1: d Michael W. Maupin, Esq. Anthony Gambardella,Esq. Richard Foster, Esq. Daniel T. Swanson, Esq. l RDP/1m
v reliable, I have difficulty accepting the inplication that in these instances the absence of another containment isolation valve results in a net improvement in safety over a design that meets the regulations. Does Mr. Shapaker believe that the regulations should be amended to require no more than one containment isolation valve in each line? I think not
~
for he has accepted lines with two valves. But if so, why stop there? Just imagine how reliable the systems would be if no containment isolation valves were provided. With regard to a valve inside containment becoming submerged, this should present no insurmountable technical obstacle. Ing When it was discoveren fact the staff has already approved such designs. that operating plants designed by Westinghouse would have other safety related valves submerged following an accident, the valve operators were simply relocated above the calculated post-accident water level. Finally, as I noted on page 15 of my limited appearance statement, the staff has previously determined that the General Design Criteria do not recognize a closed system outside containment as'an acceptable barrier. It should also be noted that even if these four lines met the specific requirements of theregulations,closedsafetygradepipingoutsidecontainmentwouldst._) be required. Thus the North Anna design has no compensating features over a design that does meet the regulations. There are many other easily detectable flaws in Mr. Shapaker's affidavit which suggest that this Board needs to conduct further inquiry. On page 4 he states that both isolation valves on the containment vacuum pump suction lines are located outside containment to inprove their reliabilis following an accident. Given the sorry state of affairs of the environmental qualification of safety-related equipment in the North Anna plants, this statement is no doubt true. But there is also no doubt that the Commission considered the effect on reliability before adopting a regulation that
Mr. Dromerick also suggests that by reviewins the SER and its Supplements one can determine whether the North Anna units satisfy the rules and regulations of the Commission. That is a specious argument which evades the thrust of my statement and is unresponsive to the Board's . request. Except for the use of the word " intent", which was addressed by Mr. Shapaker, the SER and its Supplements clearly state that the rules and regulations are met. It is the relation of these documents-to reality
+' , t is in dispute.
While arguing that the word " intent" was used in the SER ". . . . to indicate that there could be several ways that the criteria could be met," Mr. Shapaker confirms the suspicion I held that the word was used to avoid discussing the fact that the design does not comply with the specific re-i quirements of the regulations. If his interpretation were correct the word " intent" would have been used throughout the SER. The reasons he advances for finding the design acceptable, even though the specific requirements of the regulations are not met, are similarly disingenuous.
/
Consider, for example, the discussion beginning on page 3 of Mr. Shapaker's affidavit. We learn that four pipes penetrating the containment have only one isolation valve. The regulations require two valves on each pipe, one inside and one outside the containment. Mr. Shapaker's reasons for nevertheless finding the existing design acceptable apparently are:
- 1) system reliability is improved with only one valve in these lines,
- 2) the valve is located outside containment since it would otherwise be under water following an accident, and 3) a second containment isolation barrier is provided by the closed, safety grade piping outside containment. _
While, in general, it is true that a system with fewer valves'is more
l in the control room." (Page 71, Report to the Director, Office of Nuclear Reactor Regulation, Concerning R. Pollard's Allegations, U.S. NRC, February 28, 1976.) Therefore, before accepting the a'pplicant's unsupported state-mont that the auxiliary panel controls are " completely isolated" (I assume they mean electrically as well as physically) the Board should require additional testimony on this aspect of the design. Otherwise, the NRC inspector's statement that a fire would limit or even prevent the initiation of safety actions for both units should be presumed true. [') Pages 57-59 of the applicant's response addresses the problem of reactor coolant pump flywheel missiles. In evaluating this response the Board should consider the following:
- 1. " Overspeed of the pump rotor assembly during a transient increases both the potential for failure and the kinetic energy of the flywheel. The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or other equipment or systems important to safety." (Regulatory Guide 1.14)
- 2. "ECCS cooling performance shall be calculated ... for a number of postulated loss-of coolant accidents of different -
i sizes, locations, and other properties sufficient to provide assurance that the entire spectrum of postulated i _) loss-of-coolant accidents is covered." (10 CFR 50.46)
- 3. "Also, some of the specific design requirements for structures, systems, and components important to safety have not as yet been suitably defined. Their omission does not relieve any cpplicant from considering these matters in the design of a specific facility and satisfying the necessary safety requirements. "***These matters include:
...(3) Consideration of the type, size, and orientation of possibic breaks in components of the reactor coolant pressure boundary in determining design requirements to suitably protect against postulated loss-of-coolant accidents." (Introduction to Appendix A,10 CFR Part 50)
- 4. ... the Commission considers that neither consequences nor probabilities should be considered alone..." (Commission interim policy statement of August 1974)
requires one valve to be located inside the containment. Another flaw appears on page 2 when Mr. Shapaker references the Standard Review plan. But the Board requested references to rules, regulations, and policy statements of the Commission. The Standard Review plan is none of.these. I have left my comments on the applicant's response to the end for . two reasons. First, my limited appearance statement focused primarily on the staff's testimony. Second, because of the conflict between applicant's financial interests and its safety responsibilities, the applicant's , r.ements should be given substantially less weight than those of the staff. \ Hoivever, since the Board must consider the applicant's response, I will give a few examples of why I believe its response should also move the Board to conduct the additional inquiry I recommend. On pages 53 and 54 of its response, the applicant addresses the information on pages 6 and 7 of my limited appearance statement. Applicant states that because the 'K' service class cables are defined as having no I R losses, intermittent service, or continuous serviceit.ith a derating of (,40%, this provides a very conservative power level loading and no significant heating of cable occurs. Applicant fails to address the point. In eval-unting the potential for fire, one must consider the power capable of being carried by the cables should an electrical fault occur, i.e., one can not assume that everything will always operate according to-definition. With regard to the routing of all engineered safety feature control cables for both units in a confined space beneath the floor of the common control room, applicant disagrees with the statements the NRC inspector made (see enclosure 5 to my limited appearance statement). I remain unconvinced. For example ,- in evaluating compliance with that portion of General Design Criterion 19 concerning an uninhabitable control room, the staff has testified before Congress that . . . no maior damage is assumed to occur to the equipment
g g uwlTED STATES Enclosuro 2 p ". 4et NUCLEAR REGULATORY COMMISSION
?
WASHINGTON. D. C. 20555 , August 18, 1977
\...,/
MEMORANDUM FOR: E. G. Case Acting Director Office of Nuclear Reactor Regulation . FROM: Stephen H. Hanauer Technical Advisor to Executive Director for Operations
SUBJECT:
INTERACTION BETWEEN CONTROL SYSTEM AND PROTECTION T SYS.EM The Zion incident of July 12,19'/7, apparently shows a design defect as well as the obvious gross management deficiency. .The 31 dumy signals disabled the primary system level control, which initiated a transient involving decreasing level. Concurrently, the same sequence of events disabled portions of the protection functions associated with the same level. Thus a single sequence of. events } caused the tra,'nsient and paralyzed the safety provided for th,at very transient. Westinghouse designs are characterized by the large number and types of interactions between control systems and related safety systems. They think this is great. I think it is unsafe. This feud has been going on for years. I have not so far been able to find out whether a single signal or group of signals went to both control and safety, or whether the interaction was more obscure. It almost doesn't matter. I also-don't know (and don't much care) whether the interaction, whatever its nature, is allowed by the various meticulously crafted clauses in IEEE-279. For existing be taken plants, to heart andI believe acted onthe lesson of theThe constructively. Zion factincident that, this shou'
. ')s time, nothing bad happened is a tribute to good operator action and .
defense in depth, and should not keep us from learning the lesson. All interactions between control functions and safety function should be reviewed in the light of this experience. A statement that no such dummy signals are allowed is not to the point; nexttime, some different and not now foreseen sequence of events may start the ball rolling. t What is needed is adequate independence of control functions from safety functions that provide against control malfunctions. I i
b 4 l 1 , 6- i l 5. "The argument ( of low probability) is equally impressive when used by industry to say that the LOCA is not a , legitimate design basis accident. I think it's equally _ l wrong, too. The double-ended break is a design basis for ~ i millions of dollars' worth of diesels, pumps, shock absorbers, pipe whip control hardware, etc., and is - appropriate for. pump missiles, too. '(PUMP OVERSPEED PATCHES, ; , memorandum from S.H. Hanauer, DRTA to John F. O' Leary, , Director of Licensing, August 9, 1973)
)
Regarding environmental qualification of safety equipment, I ( tect the Board's attention particularly to pages 60 and 61 of.the applicant's respnse. Applicant acknowledges "...that the testing [ performed on the equipment required to operate in hostile conditions-at North Anna 1 and 2 did not include several factors associated with [ i long-term operation." -Applicant then states "However, the goal of l l long-term availability was addressed by the qualification program for ! j , the North Anna equipment," as if this somehow ameliorates the deficiencies. l ! Note that even in the example given by the applicant, there is no f sertion that thermal aging over 40 years' operation has been included i in the testing program. Therefore, I reiterate my previous recommendation: 1 th.e Board should conduct additional inquiry to determine _for what period of time the qualification tests performed remain valid for demonstrating-conformance to the regulations and limit any operating license issued so i that it is not valid for any longer than that period. :
-i l In summary neither the applicant's nor the staff's response j 4 ,
i i obviates the need for the Board to conduct additional inquiry. i I 1
,, . - . . . - .- - - , 1, - --- n--- -me yw.- ., ,,,n g,e-- - ,, ,gm
,. ?
~ E#pesan % UNITED STATES NUOLEAR REGULATORY CO.1 MISSION , ,
E? l ATOMIC SAFETY AND LICENSING BOARD PANEL
;(.. b:/,'cfl 1 f
wAsmscron. o.c.2osss
..... October 21, 1977 Daniel T. Swanson, Esq.
5 y-
%~
N g'D - p' U.S. Nuclear Regulatory Commission /,{. g,5 Washington, D.C. 20555 (, 's In the Matter of ch h Virginia Electric and Power Company - (iforth Anna Power Station, Units 1 and 2 % ,# Docket Nos . 50-338 and 50-339
Dear Mr. Swanson:
]
We have reviewed the responses of VEPCO and of the Staff to the limited appearance statements made by Mrs. Allen and by Mr. Pollard. The Staff's response does notfappear to fully respond to Mr. Pollard's statement. This may be true because the Staff agrees with positions taken by the to Mr. Pollard's statement. If Applicant that is so, with respectthere is no nececsity for the Staff to dupli-cate that part of the Applicant's response with which it agrees. We do request replies to the following cuestions:
- 1. Does the Staff fully agree with the Appli-cant's response to the limited appearance stntement made by Mr. Pollard?
- 2. To what extent did the Staff consider the 'l
./
question of " associated circuits" in its review of the routing and independence of safety cables in the North Anna Units 1 and ' 2 installations?
- 3. Is it the Staff's conclusion that the separa-tion of cables and other measures taken by '
the Applicant to insure the independence of.
=
redundant safety cables at North Anna provides is a degree of protection to the public that 6 4 essentially the same as would be provided if eK82 g full compliance with Regulatory Guido 1.75
- and the applicable portion of 1.120 had been 8
I NOV 3 0197B > r required? L m :... g.: ,- pr3
' hs , ,
e
E. G. Case 2 August 18, 1977 7
.For future plants, we haVe RESAR-414, with a new " Integrated .
Protection System " which includes interactions between safety channels and control"and between (PSAR, safety and "non-safety systems for monitoring
- p. 7.1-27).
Such interactions seem to be on a scale far beyond present practice and involve a complexity (multiplexing,
,enco.untered.. data links between computers) not previously The philosophy (old and new) is, " Westinghouse j considers it advantageous to use certain information derived .
from protection channels to control the plant" (PSAR, p. 7.1-62). The acceptability of all systems, Westinghouse and n'on-Westinghouse, old and new, needs to be reviewed in the light of the Zion event and any unacceptable interactions removed. f-( .
\
~
T ,l l g.I. x . < @ m ) r., ce r. . .
- .qcx .
- ./O' Stephen H. Hanauer Technical Advisor to Executive Director for Operations cc: L. V. Gossick S. Levine E. Volgenau R. Minogue a e .
t . 0 s I
p #" "##4 '* UNITED STATES ,
= . NUCt. EAR REGULATORY COMMISSION I b WASHINGTON, D. C. 20555
..... December 9,1977 t
r.- V Robert D. Pollard 102515th Street, N.W. t i, Washington, D . C. 20005
Dear Mr. Pollard:
In response to your letter to me dated October 31, 1977, I am enclosing the ]' NRC Staff's response to a letter from the Atomic Safety and Licensing Board to me dated October 21, 1977, regarding North Anna. Sincerely , J 4 2 5. % Daniel T. Swanson Counsel for NRC Staff
*:! 2
- 0 NOV3 01978 > ~ -
L- ;- c, . . . p
-~ q
- a t 4 .= I{.!
a M d i l
e . . .
, :. .s
- 4. To what extent were the departures from the General Design Criteria, discussed in the Affidavit of James W. Shapaker, reviewed with and approved by the ACRS. In particular, did the ACRS review the use of check valves as isolation valves in lines covered by Criterion 57? ,
Please supply to the Board copics of the document entitled
" Analysis and System Modification for Recirculation Spray Pump Net Positive Suction Head" which is described on the first page of the letter of Sam C. Brown, Jr, to Edson G.
Case dated September 16, 1977. Please advise the Board as to the present status of the Staff's plan to provide a resident NRC inspector at North Anna Units 1 and 2 ery truly yours, , Lu A - k Frederic J. Confal, Esq. Board Chairman e 9 9 4 ) e O
- G 9
e O 9 9
~
2 o, _ i 5 thare are two qualifications with respect to tnis conclusion. -- First, tne applicant will oe requiced to adnere to the turoine valve inspection, maintenance, anc testing procedures ] y 2 1 out)ined in Section 10.2 of the Stancard Review Plan (SRP, NUREG-75/Uu7) and to the turoine inspection provisions indicated in Section 10.2.3 of the same Standard Review G Jf _ Plan. Secondly, the applicant will be required to provide - additional protection if the results of the generic study 8 warrant it. d.
^
In reference to the applicant's statement on page 56 tnat =
"tne probability of strixe og an essential system can to -
found oy multiplying 2 x 10-3 by the density of essential systems witnin each ouilding," the staff concludes that the S'
, 7-statement is only valid for primary missile effects. However, turoine missile impact on concrete barriers can produce numerous secondary missiles (scabbing) such that the density of essential systems would not be a valid indicator of the strike probaoility. Unless antiscabbing shields are present, tne staff uses tne conservative assumption tnat a turoine C missile impact producing secondary concrete missiles will result in a pronaoility of one of striking safety related systems oehind the impacted barrier.
In reference to tne applicant's statement on page 56 tnat "nr. Pollard misquotes SER 510.7 as saying that tne probagility j of strike on essential systems is on tne order of 2 x 10 ," it appears to the staff that tnis ir-an inaccurate reference 5 to Mr. Pollard's statement. Mr. Pollard refers to a strike on an " essential part of tne plant," whicn is not necessarily i limited to " essential system," but could, as the applicant g later indicates, include plant structures. (R. Pollard's .'
~'
-7 written Limited Appearance Statement, following Tr 3022, p. 8).
In reference to the applicant's discussion on page 56 regarding turoine missile rotational energy, it is not clear wnetner the d discussion is directed at turoine casing penetrations or to 5 carrier penetrations beyond the confines of tne turbine. The r' 7
- applicant's reference to Westingnouse tests is meaningful if turoine missile-turoine casing interactions are being consicered. -,
b,; = However, it does not readily follow that tne same Westinghouse test 'cata snowing negligiole rotational energy-penetration Q .'
'" 2 correlation for tne turbine casing extend to barrier inter-actions in general . However, the conclusions stated in f ,
Section 10.7 of Supplement No. 2 to tne Safety Evaluation Report 'i
;J remain unchanged.
~
~
I R n 4
I e STAFF RESPONSES TO BOARD'S REOUEST FOR c ADDITIONAL INFORMATION [
~
LETTER DATED OCTOBER 21, 1977 1 t Qucstion 1: Does the Staff fully agree with the applicant's response to the L limited appearance statements mace by Mr. Pollard? - Response: The staff has reviewed VEPCO's response to Mr. Pollard's limited appearance statement cated July 5,1977, and agrees with its ' responses witn the following additional comments: ,e. (1) 'On page 49 of tne applicant's statement VEPCO states tnat "it believes that all parts of North Anna Units 1 and 2 comply with NRC regulations.' It snould be noted tnat in a letter dateo August 9,1977, VEPCO requested relief from certain preservice inspection requirements involving the reactor coolant pressure boundary. In a letter dated OctoDer 11,1977, VEPC0 provided additional supporting information to justify tneir request for relief. As a result of our review of tnis information we have recently determined that tne preservice inspection program for North Anna Units 1 and 2 does not fully comply with 10 CFR Section 50.55a(g)(2), and have also determined Inat an exemption regardi~ng this matter for Unit 1 is j usti fied. (2) In refereace to the applicant's statement on page 55 that "the [ turbine missile] risk ... nas Deen found to De acceptably dow," the staff believes that some qualification is necessary. \ It should De noted that the staff's conclusions in Section 10.2'of Supplement No. 2 to the North Anna Power l Station, Units I and 2 Safety Evaluatin Report limit the acceptability of the risk to being sufficiently low i for permitting tne plant to operate until a generic study I .. on the subject of turbine missiles is completed Further r 1 1
.. i I
From this sentence, and the context of the rest of the page, it could ue inferred that MULTIFLEX calculations F7 using the North Anna conditions (i.e., a break area of 7
- ia square inches and a creak opening time of 10 milli- f '
seconds) were done. In fact the analytical work consisted FA , of a calculation done with the approved code version, at a M J oreak area of one square foot and a time of one milli- 1 second, in combination with an estimation of the load reduction ?i using otner calculations. The staff approval of HULTIFLEX, j in its Safety Evaluation Report of 6/17/77, did nr' includ,e
- g MULTIFLEX at break opening times other than one mr.lisecond.
In order 1;o furtner explore the effect of Dreak opening time,' oreak area, and the flexible wall assumption, tne staff and VEPCO recently performed adattional calculations with these 3 ' parameters as variables. These studies, although not yet fully documented, confirmed our earlier judgment expressed in Section 4.2.4 of Supplement 7 to the North Anna Power Station, Units 1 and 2 Safety Evaluation Report, wnich was arrivea at oy the combination of an approved MULTIFLEX calculation in concert with approximate sensitivity stucies. We understana that VEPC0 wi11 document tne results of tnef r calculations soon. With respect to page 76, tne limiting model for the reactor pressure vessel support used a limited aisplacement break, modelea as a. slot, not as a guillotine break as stated by VEPCO. Tnis model is acceptaole and does not change our conclusions regarding this matter. With respect to page 75, tne words, " substantial margin .* ) exists" were used in the discussion relating to the fuel rod , spacing reduction during a postulated accident. We believe
- tnat " adequate" rather than " substantial" may have been more appropriate. However even,tnis qualification deserves further attention in view of the following.
As a related matter, we Decame aware in October,1977 of h new information in the matter of the structur'ai response 't L of reactor internals to asymmetric loads. Our consultants at EG&G, Ioano, Inc., who assisted us on the North Anna evaluation, 5 f, i l
(3) With respect to the reactor coolant pump flywneel missile - (pp 57-59), the applicant states that: .
"The generic stuaies are being perfonned for the purpose p of subtantiating or modifying current mathematical models lj that predict the pump performance during a LOCA." 7 Tne staff concurs that the prooability of a LOCA occuring ) '
witn aeditional consequences caused by tne reactor coolant '~ l pump flywheel missiles is small; however, we do not completely discount this postulated accident. Rather, the staff has determined tnat the probability is low enough that this
, postulated accident does not result in an unacceptable condition. * -
Studies are being conducted by the Electric Power Research Institute (EPRI) to provide information regarding this matter. When these studies are completed, they will ultimately substantiate or refute the conclusions reached in the Westinghouse topical report WCAP-8163, " Reactor Coolant Pump Integrity in a LOCA." At tne present time, the staff has not evaluated tne conservatism in the calculations presented in WCAP-8163. i (4) With respect to the residual heat removal (RHR) system over-pressure protection (page 77), we agree with VEPCO's response, except for tne following: , a) The relief valves'are located on tne suction side l of the RHR pumps (not the discharge - page 78); and, i i o) Tne set pressure is 475 psig (not 600 psig - page 78). (S) Witn respect to LOCA Effects on Fuel and Pipe Break Spectrum (p'p. 74-77), the applicant states that:
"Using the assumptions and the version of the MULT FLEX -
code apprued by the NRC Staff, the analyses demonstrate 'c tnat substantial margin exists between the applied loads . and the load at which the fuel grids subject to maximum I f loading would sustain measuraole permanent deformation." F; (page 75). j ; b , V . u
.. Question 3: Is 1t tne Staff's conclusion tnat tne separation of cables and _
otner measures taken by the applicant to insure the independence. of recuncant safety caDies at North Anna provices a degree of protection to tne puolic- that is essentially tne same as would be ]t-y provioed if full compliance witn Regulatory Guide 1.75 and tne j applicable portion of 1.120 had been required? Q > Response: Regulatory Guides 1.120 and 1.75 do not apply strictly to North Anna 1 and 2 since they were issued after tne North Anna review h commenced. The staff.is re-evaluating the Nortn Anna Units 1 and d;e 2 Fire Protection Program against the guidelines of Appendix A to dranch Technical Position APCSB 9.5-1 "Guicelines for Fire Protection for Nuclear Power Plants Docketeo Prior to July 1,1976." wnen tne staff's re-evaluation is' complete, North Anna Units 1 , and 2 will have installed or be committed to install prior to plant operation following the.second fuel cycle any additional fire protection features that the staff concludes are necessary to provide additional defense in depth against fires. In the interim, wnile it is true that the plant may not conform - witn Regulatory Guides 1.75 and 1.120, the plant presently employs other provisions to provide an acceptaole level of fire protection in accordance with General Design Criterion 3. These include automatic fire detection and suppression systems in vital plant areas, such as uncerfloor area of diesel generator rooms; manual fire and rated five barriers which establish the defense in ceptn protection. These features in conjunction with the ongoing staff re-evaluation of fire protection, and the icw probability of a major fire during tnis period, make initial operation of the Nortn Anna Power Station, Units 1 and 2 acceptable. -i
)
Question 4: To what extent were tne departures from the General Design Criteria, discussed in the Affidavit of James W. Shapaker, reviewed with and approved by the ACRS. In particular, did the ACRS r~eview the use of check valves as isolation valves
- in lines covered by Criterion 57? -
il Response: Tnis matter was not discussed specifically with the ACRS for ? Nortn Anna. In the Affacavit of James W. Snapaxer we stated Q our' Dasis fo'r acceptance of tne valve arrangements. In our @ view, we aid not Delieve tnat discussion of tnis matter witn fi,- the ACRS was necessary. $4
. . e. e 5
.5-e are providing analyses in support of our generic review of calculational models for computing structural responses F_
being developed by reactor vendors. They found in a sensitivity study tnat the grid impact force is more
]G(
l sensitive to the core plate motion tnan was originally tnought. Grid impact force is the controlling factor on Q H
.l the fuel rod spacing. Since it is difficult to calculate tne core plate motion exactly tnere arises a question of T3 margin to deformation of the fuel grids at North Anna p and al] other pressurized water reactors.
Ed Tne consultant stated (Enclosure 1) that this new information is very preliminary. In addition, our preliminary understancing from Westinghouse is that tney have not observed this sensitivity "' ( to core plate motion in tneir analyses. The staff is continuing I to evaluate the nature of the sensitivity and the sensitivity i stucy itself continues at EG&G. This bears directly on the l discuss, ion on page 77 of the VEPC0 response with respect to the conclusion regarding "most severe potential consequences". Given the magnitude of the sensitivity, the preliminary nature , of its icentification, tne margin of failure in the grids, and tne generally acknowledged capability to cool the core even in the event of local grid deformation, we have determined that our conclusions stated in Section 4.2.4 of Supplement No. 7 to the Nortn Anna Power Station Units 1 and 2 Safety Evaluation Report regarding this matter remain unchangec. Question 2: To wnat extent did the Staff consider the question of " associated circuits" in its review of the routing and independence of safety cables in the Nortn Anna Units 1 and 2 installations? ( Response: The concept of " associated circuits" was first introduced in Regulatory Guide 1.75 (which endorsea IEEE Std 384) in 1974, several years after the start of construction of North Anna Units 1 and 2. Tne crux of this concern is the interaction of non-safety circuits on the independence of redundant safety systems and circuits. , 7-m Uuring the s'taff's review of tne North Anna application, it was estaolisned tnat the criterion for routing of non-safety if circuits whicn Decame ," associated with" safety circuits is that d tne non-safsty caoles were not permittea to be in close proximity Es with, or routed with more.than one division of safety cables. The ? applicant was required to implement tne plant caule routing to the f.$ above stated criterion and at a site visit tne staff has verified Q l tnat the plant caole routing was implemented in accordance with i l tnis criterion. The staff has performeo an audit of tnis design anc has found that it satisfies the above referred criterion. l
ENCU)SURE 1 , ; . [M.5/.'k3 lel:il"i lent. - F E.1. l illn , piret.L'm , h' I?. 'e I c r l'pri a t ions r,1:1 vigi anc. Divis len 1 lahn tirei .it lens Of f le.c . g;r,T , l<l.ibn 1 4 1 1 ' . Irla be H:1101 i rtir tillt A'.r,f f!Dt.-( litt.it! t!!rAl.1:15pf'll'I' /.ffAl}SI'. - S t.lg-?lf-77 - 1:r.f : (a) p. l.. Grntl.,It:R inel 't. sewl1) fle."hanical Response Analysis, - Idahe flat tenal Engince ioq 1.al uratniy , Pr-C-77-141, March 1977 (h) p. l.. I:cel b, .rvn 1901 A. trolly llechanical Response Analysis, , tn.cn.tmuni lie.1. idahu Of inoci Lnpincering Laboratory, Pr-l -7 7 14I.T. l' art.li. 1971 x .. (r; p. t . Orvi:b ani? II. F. Sofiell, Jr , f;on..l.incar lateral liechanical , I esi onse of prest.uri.'.cd I.'ctoi I!cas. lor l uci Assemblies, ASME I'rne 77.ltf /IN-ll:, l'ecer+her I?77 (al 11. Ilunn. I'. Ili tut a , an I l'. P.e'enna , Drnelopicent of Advanced .
!!.:ll.cd re, l ect Er i .n ic f.nalyt.I'. <lth International Conference en r ti nt.tui .il llachanics in Pc n.L'?r lethnolo'Jy, San Francisco,
- f. alii nt e:la . (TA , Aucus t , l'177 (c) I!. L . I:ivbl . l'ossal li i ts S linh f or l' entitling tlie Lateral PWR '
- vel As'enhly I!ichanical 1sdionso Analysis, Idaho National I nginerrmg Lal4rainry, pl -l-7 7-1(0, Pev.1, July,1977 tii ar f tr. Iilint:
" ~ A pai .urr s i ii. !.tur.ty t o ai.scss t he elii e i ol ariaticus in cerc plate motions - in. toni a ...bly .l.:irei grid rinr.hinn loa #, is current.ly in progress. A
<e.t..iy el. .iipt.ien of thir sterly iniInding pieliriinar y results has been -
- p. . . i .n . . ' zi ' the teiluest of the lincle.i pegulater) Conwission's Division ", ,
ei a... . t er '.3 f ety, ( cre rcrf un niance l'i an' li. I'c'ults of this study indicate - ' s.lia l a < ca l l i;r i t a l. inn in i.nic p101.c fretinent, ru,s have a significant elf <cI en
- as.or oriilv. rushing lea'Ir. A: tho st.udy is not complete, these tesult : '.he.uld tc rein .itlet cia in cliniis ,iy.
. A re.l.anisiin t.as p< ..lulatu! in Referenr.e (c) uhic h indit.ated that the in- ,.
rut core lilate action e onid significan'.ly af fcci spacer grid crushin y h;. il . . 1 F. Pr in.ar y ni.ici.I I'c cf f.he prer.e nt sintly v.as to deternr ne f
. yj t.hi' n ot.h.in i ,im e.nl:I l e slimen to ( r i .i . A s. rurolar y oh.iective is to k, e omnai e Iice pc anel, nonIincar analyt. is t ct.hnlilues, in sunwary then tha NU pm i n' n n' Ihis stus;t is tv.nfolel: fj
- tt (1) c t:, list ie.t.ll s detciniine t he cf f cc t of Lei c pla te frequency 'p and r:aanitut'e en the incl c.5,cu bl,s ninitrein spater grid L.s rinchinn 1 .J. , and 1
(*) *t.alistically cenipare lint.i anel nonlincar analysis metPods t i:r latcial turi essenhly ner.Lanical iesponse in an attempt tr ?.iuplif y tini nonlivr ar analy'i! . I
i 7-Tne Board also requested tne present status of tne staff's plan to provide a resicent NRC inspector at Ndrth Annh Units 1 and 2. Initial implementation ... of Ine resident program including manning of the Nortn Anna Units 1 and 2 , facility was discussea in a Commission Information Report, SECY 77138B , (Enclosure 2). Tne cate of actual impleinentation is cependent upon approval ;[ of Ine program via the budget process. It is currently expected th,at the , i idortn Anna site will De raanned by Spring,1978. r f - E' i
-. i
( .- l t l 5 m ( e
, t 1,
i, t
,- ~ --
;o
;- a:::c:a
. np s.;
R. C. Tiller ,
.froV - 4 181/. . .
S tirJ-316-77 Page 3 . t Re ferent.e (e) cauld actually be clict cd in thr* nonlincar analysis. The en:ch- - ani:.tn c;i n : s D.i .:<ist; ;'ieruby causf ug concern t. hat rim. ncnt dufomation i h[ h of r.p. cue f id : :y r ccur. . h Uten cc::plet.ities of thf 2 :tudy the conclusions presented in Reference (b) - wil 1 Le re.:sse:.a.eJ.
. Very truly yours. -
3 .. . m:cNA: 5 cst:0 ny R. R. Stir;cr. i'.anager Reactor Dr:h.Tvior Df vision BFS cij cc- V. Stelle. f;nC-00:t , P. S. C!ver.t.. !;::C-USS S. B. K 1-). L .-033 R. J. ! bit o:i ::::C-fL*.S . R. O. l'.'yi:r. t:P.C-3ss - D. F. h ' L. tiAC-USS R. *.J. Ki n:.rs. ECl4 Idaho . tcc: R. f. . Gru!: j R. 'J. l'.:ee.t C. A. Yavr:: - C. F. G./c dain@,r .c.' * - D. F. Sat ell 8/ W
~
.~, . - .
G. L. Thim.cs D - e T. R. II;o:.;;;.cn h~. IW P. 11. Vandce Hida .
?d L. J. Yharren.!o !3 C.,n t.ra l F 1 l e -
F 1 l e; LM l 3 t IC
)
-d
. l i
\
, e t
f'e, y a-R. T. Tiller 1:vv. _ l 7 7 S t l .j- 3 3 in-7 / P.v;c % w The structure! -cJe1 u!.ili.:ed to analyze the funi assec:bly r..echanical - d re. <,nse is 1:as t. elly dewr ribed in Cercrer:ces (4) thrautzh (c). Two J exce;stier s ir><. l uded i n ' t.e pic*,er.t study ore the use of fus-1 assccibly C.. ori e.15 r i.t a,1 tre. t.ti nr les ne .I ::o.:a s'i.if us or.d utili ta t. lutt of thu r:.ethod [
~
pr e-'.e nts: f in R :.c. unce (d) t er calculation of spar.cr orf l d crushing i are h 1 .m D. . T t.e 7,. s t n.s 1 f t,e t t o '$ pre :.. n te;d in Pr.fi ,i.. e U.)g. f une ..hilet!un.et.c.tv.cre variaplatetion ,det.e on t.fic eret ons frequency y and i.:.q,l i t us.!*. <,i s.c e e 3.l e t .. o nt:(,n*, are to 1:e cuoside red, unty the four - - - o.c t re: .e- cec:. cre e ri!rt:n,.d t e; thi s d r:,ct.s len. I nt: four ca'.er. are It 4 10. varf.. Lit.n t. t r i. .n . s. . . ,- s v. + ls.', variattun of the ar;:litude. . T . nu t e.! ti.3t ali t ne te e.,aere ic :1.r.e.a ir.ed in the core plate eutions ' f varie d tt:u sei e e .uant. !;calintar dyncr.ic analysis as dencribiad *
- 1. .h.ferent_es (a ) ths .a.qh (d) 15 ir. progre>> and pralininary restif ts cre prov(<:ed in ;chle 1 is 1Inear oneIy5is i5 also,beinej r.ucsued using the s.,et.l.nd: uullis,.'d in Eet er ence (c).
)
T/.Pt r 1 RATIO Of P 'AK f.T r*,CLE G;.ln Cflus!UtiG LOAC TO THE l l.r.it!:171 CbsM-It.G LO,*.[i Me r. in i.t.: Cru:,hing Load /*;cais.el Crushing Load Spanar Grid Elcyation .10. 'i fr.eu acncy
+10f Frcquency. .107 Anulitude 6101 #a.plItude
=-
~
1.76 0.801 0.979 1.04 Canter ( 1.45 0. M 4 0.812 1.23 ster-up 0.830 0.835 1.2b Center-dc4,n 7.11 1.31 0.945 0.771 1.34 Tn;. 1,56 O.8ed 0.863 1.26 . ent ter,
. . ~....
f* f". '. " ' '..,,. 3 .M d !s ? d- [.t*d hj.dc.*r e.cid (.r uf. hint) In/A GDtdingd j :. (f (.f'; i t t al * [ I I U. a I.P - L '. t * - I.Ur e I ' #**b IV ' . *8 .
. - - ..... t
'c l};
, ' r . % '. 1 - ? . 1 t e ce '. Appear t!. t. e variat ion in freque.ncy
'4 tw .' e i *.! . r s. . 'en.;c in ti e Sr.acer grid crushine, load,.
? . .:<t, e ,f:. iw.t
;;.:: .:..;:ter sicy Le in urifer (nr this L)pe of r.e.. :: rt.4
* .e < c , .i *. i u . tr. *nt.
1r.1: s o.* p .t. d...1d i.e p..it.t.:d cut that U:c s u.:c1 stunted rep- O9 tz t e r.ra t io.m
:1 t t; . , . t. , a <;. r.e .e '. r e:r.' i e;., r *. . r. .
11.e ;*arpC'.e of. tl i:, study s..n not a direct on.ly.is of 4 4 .. i f 1 (. platit Lut to deterr:ine if t.hu ::ct.hantsim pastulated in
T. ; , r..' ,,
. ~
The Commissioners - Additional information pertinent to the eight selected T~ facilities is presented in Table I. - p uk It should be noted that implementation o'f the Resident Inspection Dragram will be on a provisional basis pending i)8 allocatio of resources and approval of the program via i the budge; process. [3j 1 2. h/ - E st Volgenau ^ Director Offi'ce of Inspection .
- and Enforcement
Enclosure:
Table I , DISTRIBUTION Comissionets .
"ormaission Staff Offices
- Exec Dir for Operations Secretariat
}
e I = m 7~ w 1 z 3, m ep e
ENCLOSURE 2 : July 20, 1977 . . tAITto starts SECY-f7'-1383
~
NUCLEAR REGULATORY COMMISSION WA8HINGTQM. D. C. 2C855 IE BLE COPY For: INFORMATION ine ,Lca.m ssanErs REPORT - From: Ernst Volgenau, Director - Office of Inspecticn and Enforcemen r Thru: Executive Director for Operations # i
~
Subject:
RESIDENT INSPECTION PROGRAM
Purpose:
To infor=.the Cc=:::1ss!.or. of [hh selectica of eight sites fo.
. initial manning By Resident Inspectors. ..
( Discussion: The initial selection of sites for resident insp'ectors - has been made using the following criteria: - e Involvement of all Regions.. e Mix of facilities in the three phases--operations. ' preoperational testing, and construction.
' ~
. . e Mix of sites with high'.and low numbers of rea.1 or apparent problems.
e Available heavy workload to permit inspectors to build an experience base as rapidly as possible. e Mix of types of reactors.
.( The eight sites selected fer the beginning of the Resident
. Inspection Program are:
Millstone Point Salem Browns Ferry North Anna . - r-C0py SEMI '- g,3Cg g{gQM D. C. Cook 32 Midland P
/f
- Arkansas Nuclear 1 San Onofri .]
c, The chosen facilities represent a good cross sect' ion for b% ~ gaining diverse experience during the early stages of the % Resident Inspection Program. This experience wil-1 contribute d greatly to the final program development, and it was con-side ed vital that the selactions be nade representative of the spectrum of conditions that may be encountered later. CONTACT: . J. H. Sniezek 49-27451 -
TABLEI(Continued) NRC . SITE UTILITY REGION REACTOR PilASE REACTOR TYPE , REMARKS D. C. Cook Indiana & Michigan III Unit No.1 - Unit No.1 . Resident inspector Power Company- Operation Westinghouse, PWR presently assigned. s Unit No. 2 - Unit flo. 2 - Preoperational Testing Westinghouse, PWR Midland Consumers Power III lini t No. 1 - ' U' nit No. 1 - I Company Construction B&W, PWR Uni t flo. 2 - Unit No. 2 - Cons truction B&W, PWR i AND Arkansas Power IV Unitflo. 1 - ' Uni t No.1 - i & Light Co. Operation B&ll, PWR Unit No. 2 - Unit No. 2 - ' Preoperational Testing CE, PWR l " l San Onorra Southorn Cali- V . , linit tin. 1- linit No.1 -
- fornia Edison Co. Operation Westinghouse, PWR Unit No. 2 - Unit No. 2 -
Construction CE, PWR Unit No. 3 - Unit No. 3 - Construction CE, PilR l i
. .t
. i: -
t i
$b N
Tygsysiah1 [MWhiiMal '
- _ _ _ _ . _ _ _ _ _ _ .}}