ML20080D270
| ML20080D270 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/06/1984 |
| From: | Baxter T CAROLINA POWER & LIGHT CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| ISSUANCES-OL, NUDOCS 8402090067 | |
| Download: ML20080D270 (69) | |
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BEFORE-THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
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CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN -) 50-401 OL MUNICIPAL POWER AGENCY )
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'(Shearon Harris Nuclear Power )
Plant, Units 1 and 2) )
APPLICANTS' ANSWER TO WELLS EDDLEMAN'S MOTION FOR FURTHER DEFERRAL OF PARTS OF CONTENTION 107 AND TO WELLS EDDLEMAN'S NEW CONTENTIONS AND AMENDED DEFERRED CONTENTIONS-IN RESPONSE'TO STAFF SER Thomas A. Baxter, P.C.
Deborah B. Bauser SHAW, PITTMAN, POTTS & TROWBRIDGE Richard E. Jones Samantha Francis Flynn Hill Carrow CAROLINA POWER & LIGHT COMPANY Counsel for Applicants Dated: February 6, 1984 gg2Ogggg7nggygg o G
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'9 TABLE OF CONTENTS Page(s)
Introduction................................................. 1 The Eddleman Motion.......................................... 2 The SER Contentions.......................................... 6
' Contentions 107-X, Y and Z........................... 8-11 Contention 173...................................... 12-19 Contentions 174, 175, 176 and 177..'................. 19-24 Contentions 178 and 179............................. 24-29
. Contention 180...................................... 29-30 Contention 181...................................... 30-34 Conclusion.................................................. 34 Attachments A, B, C and D
6 February 6, 1984 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY'AND LICENSING BOARD In the Matter of )
)
l CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL
. MUNICIPAL POWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant, Units 1 and 2) )
APPLICANTS' ANSWER TO WELLS EDDLEMAN'S MOTION FOR FURTHER DEFERRAL OF PARTS OF CONTENTION 107 AND TO WELLS EDDLEMAN'S NEW CONTENTIONS AND AMENDED DEFERRED CONTENTIONS IN RESPONSE TO STAFF SER Introduction On January 17, 1984, intervenor Wells Eddleman filed two pleadings entitled " Wells Eddleman's New Contentions and Amended Deferred Contentions in Response to Staff SER" (SER Contentions) and " Motion for Further Deferral of Parts of Con-tention 107" (Eddleman Motion). In this Answer,~ Applicants re-spond to both-of these pleadings. As discussed in detail below, Applicants believe that the Board should not defer ruling on the admissibility of Contention 107 but rather, should reject it. Applicants also oppose the admission of Contentions 107-X, Y and Z, and 173 through 181.
k l The Eddleman Motion i.
The Eddleman Motion seeks to continue the deferral of Eddleman Contention 107-A through 107-L, deferred by the Board in its Memorandum and Order of September 22, 1982 pending issu-aace of the Shearon Harris Safety Evaluation Report (SER).
Memorandum and Order (Reflecting Decisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, 2106 (1982). Contention 107 is no more than a listing of unresolved safety questions allegedly applicable to the Shearon Harris fa-cility. See Eddleman Supplement to Petition to Intervene, May 14,.1982, at 213-15. The contention was proposed prior to is-suance of the SER. It contains no explanation of why the
' Staff's treatment of these issues in the SER is inadequate.
The rationale proferred by Mr. Eddleman for continuing to defer a Board ruling on the admissibility of proposed Contention 107 as originally drafted is Mr. Edd' leman's opinion that the SER fails to address generic unresolved safety issues in a manner
. that enables him to refine Contention 107. (Notwithstanding this claim, Mr. Eddleman also has formulated three contentions, 107-X, Y and Z, which challenge the adequacy of the SER's dis-cussion of generic unresolved safety questions. Contentions 107-X, Y and Z are discussed below.)
The Eddleman Motion is based on Mr. Eddleman's misunder-standing of the obligation of the NRC Staff to address in a
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L c - facility's SER unresolved generic safety issues applicable to
'.the particular plant. Undoubtedly, in the SER the Staff must provide an explanation of.why operation of a particular facili-ty can proceed even though'an overall solution to an applicable generic problem has not been found. Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2 ) , ALAB-491, 13 N. R.C. 245, 248 (1978); Gulf States Utilities 9
Company.(River Bend, Units 1 and 2), ALAB-444, 6 N.R.C. 760 (1977). The most common justifications for permitting the plant:to operate, notwithstanding the existence of significant applicable generic safety questions, are that a solution satis-factory.for the particular facility has been implemented; a re-striction on the level or nature of. operation adequate to elim-
~inate_the problem has been imposed; or the safety issue does L
not arise until the later years of plant operation. North Anna, supra, 8 N.R.C. at 248.
The SER must explain the Staff's position on significant
. generic safety questions applicable to the Shearon Harris fa-cility; the Staff cannot ignore these issues or state only that
'a. search for a generic solution is underway. North Anna, supra,.8 N.R.C. at 249. "The Board should ... be able to look.
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to [the SER] to ascertain the extent to which generic unresolved safety problems which have-been previously identi-fled in'(an NRC technical document] have been factorad into the
. staff's analysis for the particular reactor -- and with what
t result." River ~ Bend, supra, 6 N.R.C. at 775. However, in
. stating its reasons for allowing operation to go forward, the Staff ^certainly can rely on and refer to documents which pro-vide a much greater-substantive discussion of the generic safe-
.ty. issue. : North Anna, supra,.8 N.R.C. at 248 n. 7. If it were not permitted to do so, the SER would be an extraordinarily vo-
' luminous recounting of complex technical issues already articu-lated in detail in-other NRC technical documents.
Appendix C to the Shearon Harris SER is specifically in-cludedLin_the SER to respond to the Appeal Board's 1977 River Bend decision, ALAB-444,'and the Appeal Board's 1978 North Anna
^ decision,'ALAB-491, on the-SER treatment of unresolved safety questions. See SER,. Appendix C at C-2. In Appendix C, the Staff generally explains its policy on and treatment of l unresolved safety issues, and specifically addresses the 16 ge-neric safety ~ tasks applicable,to the.-Shearon Harris facility.
NUREG reports containing proposed. Staff resolutions of generic safety issues have been published on 7 of these 16 tasks and Jare appropriately. referenced by the Staff in Appendix C at~C-5.
'In addition,. applicable portions of the SER that address these
- 7. issues are referenced in Appendix C. See Appendix C at C-5 (Table-C.2). .With respect to the 9 ra.maining safety issues, Appendix C contains a detailed discussion of these tasks. This discussion appropriately includes references to related discus-
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sionsfin the SER, and to other relevant technical documents.
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See Appendix C, Section C.4 (pp. C-9 through C-21). Compare Louisiana Power and Light Company (Waterford Steam Electric Station, Unit 3), ALAB-732, 17 N.R.C. 1076, 1112 (1983) (refer-
'ence to' insufficient one page boilerplate discussion of unresolved safety issues in SER).
At the same time that the Staff is obligated to identify in the SER unresolved safety questions applicable to the Shearon Harris plant, and to discuss the Staff's basis for per-mitting the plant to operate notwithstanding these outstanding issues, Mr. Eddleman also has certain obligations as an inter-venor. Duke Power Company (Catawba Nuclear Station, Units 1 and-2), CLI-83-19, 17 N.R.C. 1041, 1048 (1983). " Parties in-terested in litigating unresolved safety issues must do some-thing more than simply offer a check list of unresolved issues; I
they must show that the issues have some specific safety sig-nificance for the reactor in question and that the application fails to resolve the matters satisfactorily." Metropolitan l Edison Co. et al. (Three Mile Island Nuclear Station, Unit No. 1), ALAB-729, 17 N.R.C. 814, 889 (1983), citing River Bend, l
l -supra, 6 N.R.C. at 772-73. Contention 107 as originally drafted utterly fails to satisfy this standard. It contains no discussion whatsoever'of the Staff's proposed treatment of ge-neric nresolved safety issues as they relate to operation of f the Shearon Harris facility.
In conclusion, there is no basis for continuing to defer proposed Contention 107. The issues identified by Mr. Eddleman in Contention 107 now have been addressed extensively by the Staff in Appendix C as well as in other sections of the Shearon Harris'SER. Contention 107 as originally proferred contains no supporting statement of basis for Mr. Eddleman's sweeping as-sertion that the Staff has inadequately addressed unresolved safety issues. Mr. Eddleman's failure to amend the original Contention 107 to address specifics of the Staff's treatment of the unresolved generic safety questions applicable to the Shearon Harris facility constitutes a default on these issues.
See Three Mile Island, supra; see also 10 C.F.R. 5 2.714(b)
.(contentions must state a basis with reasonable specificity).
Accordingly, the Eddleman Motion should be denied, and Conten-tion 107-A through 107-L should be rejected.
The SER Contentions The established standard for admitting new late-filed con-tentions in an operating license proceeding that are based on previously unavailable licensing-related documents is set forth in 10 C.F.R. S 2.714(a)(1). See Duke Power Company, et al.
.(Catawba Nuclear Station, Units 1 and 2), CLI-83-19, 17 N.R.C.
1041 (1983). Not only must a petitioner proposing a late-filed contention satisfy the " bases . . . with reasonable specif-icity" requirement of 10 C.F.R. S 2.714(b), but he also must t
establish that a balancing of the following five considerations favors admission of the late-filed contention: (i) good cause, if any, for failure to file on time; (ii) the unavailability of other means whereby the petitioner's interest will be pro- o tected; (iii) the extent to which the petitioner's participa-tion may reasonably be expected to assist in developing a sound record; (iv) the extent to which the petitioner's interest will be represented by existing parties; and (v) the extent to which the petitioner's participation will broaden the issues or delay the proceeding. 10 C.F.R. 5 2.714(a)(1).
The first factor in Section 2.714(a)(1), the so-called
" good cause" factor, is satisfied if an intervenor can estab-lish that the specific late-filed contention (i) is wholly de-pendent upon the content of a particular document; (ii) could not therefore be advanced with any degree of specificity (if at all) in advance of the public availability of that document; and (iii) is tendered with the requisite degree of promptness once the document comes into existence and is accessible for public examination. Catawba, supra, 17 N.R.C. at 1043-44. An intervenor is considered by the Commission to have accepted the obligation of uncovering publicly available information, not-withstanding its voluminous, abstruse or technical nature. Id.
at 1048. Thus, "the institutional unavailability of a licensing-related document," such as the Safety Evaluation Re-port, "does not establish good cause for filing a contention
late'if information was available early enough to' provide the basis for the timely filing of that contention."
Id.
Contentions 107-X, Y and Z. Contentions 107-X, Y and Z
- are premised on Mr. Eddleman's view that the SER's discussion of unreviewed safety issues (USIs) cannot rely at all on refer-ences to extrinsic technical documents. For example, in Con-tention 107-X,-Mr. Eddleman criticizes the SER treatment of USI _
A-40, Seismic Design Criteria -- Short-Term Program, because it
-references the Shearon Harris FSAR. SER Contentions at 2.
This absurd allegation of course is the basis for the Eddleman
. Motion seeking to defer'a Board ruling on Contention 107-A through 107-L. Contentions 107-X, Y and Z for the most part are.no more than a check list of unresolved safety issues.
(only three of the issues raised in the earlier proposed Con-tention 107(A) through (L) are excluded from the 107(X), (Y) and (Z) check list.) For-the reasons stated above in response to the.Eddleman Motion, the Staff SER satisfactorily states the
- position of the Staff on unresolved safety issues applicable to the Shearon Harris Facility. Accordingly, Contentions 107-X, Y and-Z should.be rejected.
There are, however, several more specific statements in Contentions 107-X and Y that merit an additional response. Mr.
-Eddleman complains in Contention 107-X that in its discussion
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fof USI A-17, Systems Interaction in Nuclear Power Plants, the SER. fails to relate a referenced Sandia study to the Harris l
' plant.- As, explained in the SER, the Sandia study is the gener-
-ic NRC Staff effort, currently in progress, directed at identi-
.fying potential. adverse systems interactions that may not have
.been considered by current review procedures. SER Appendix C
-at C-10. The Harris plant is one facility that potentially
-could be'affected'by the outcome of this study, which is why USI'A-17 and a discussion of the Gandia effort is included in ,
the Shearon Harris SER. . With respect to the one potential ad-verse interaction identified to date from the Sandia study, corrective measures have been implemented at Shearon Harris.
Id. at C-10 and C-11.
In Contention 107-Y, Mr. Eddleman has misconstrued the Staff's position.regarding USI A-44, Station Blackout, in stating, " staff review not to be complete until some time be-
-fore fuel load." SER Contentions at 3. While it is true that the Staff will perform a review of emergency procedures and
. training programs for station blackout events prior to fuel loading for conformance to Generic Letter 81-04, the SER makes clear.that-completion of-this review is not necessary in order for the Staff 'tx) provide sufficient justification for operation pending resolution of this issue. As the Staff notes, there is l reasonable assurance that a total. loss of all AC power will not prohibit adequate core cooling'because of, e.g., the ability of
'the emergency feedwater system to function without AC power, and the programs undertaken-to improve the reliability of the
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diesel generators. SER at C-13. Mr. Eddleman takes issue with neither of these substantive findings, which form the basis for the Staff's conclusion that Shearon Harris can be operated be-fore.the ultimate resolution of USI A-44. He therefore has failed to set forth a reasonably specific basis upon which to challenge the Staff's resolution of this issue.
In Contention 107-Y, Mr. Eddleman also focuses on the in-terim status of resolution of USI A-47, Safety Implications of Control Systems, apparently inferring that there is no basis for allowing operation prior to the final resolution of this issue. Again, however, Mr. Eddleman fails to address the tech-nical basis for the Staff's conclusion that, subject to the re-ceipt of appropriate information from Applicants on equipment qualification (see SER $ 3.11), and the completion of instru-mentation modifications committed to by Applicants (see SER S 7.7.2.2), there is reasonable assurance that the Harris Plant can be safely operated before the ultimate resolution of this generic issue. SER at C-18 and C-19. Mr. Eddleman has not specified any basis for challenging this Staff finding.
Applicants also oppose the admission of two of the safety issues which Mr. Eddleman seeks to raise under the aegis of Contentions 107-X and Z because they are the subject of two previously admitted contentions. USI A-1, "Waterhammer," ref-erenced in Contentions 107-X and discussed in Contention 107-Z, is encompassed by Eddleman Contention 45 dealing with the same subject. Both~ contentions concern the adequacy of the design of the main steam and feedwater systems to withstand water ham-merievents. Furthermore, the Maine Yankee water hammer event referred to in Contention 107-Z was caused by the top feed ring design of the Maine Yankee steam generator, which the Harris generator does not use. See SER, App. C at C-7 (reference to Harris steam generater " bottom-feed, preheat de~ sign"). Simi-larly, the concerns associated with USI A-3, " Steam Generator Tube Integrity," reference by Mr. Eddleman in Contention 107-X, already have been raised in Joint Contention VII, co-sponsored by Mr. Eddleman. :Because these two issues are clearly redun-dant, they should not be considered within the scope of pro-posed ~ Contention 107. See Memorandum and Order (Reflecting De-cisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, at 2090, 2095 (1982).
In summary, Applicants oppose the admission of proposed Contentions 107-X, Y and Z. These contentions are no more than unsupported " check. list" allegations that fail to meet both the general pleading requirements of 10 C.F.R. $ 2.714(b), and the specific pleading requirements for unresolved safety issue-con-tentions articulated by the Appeal Board in ALAB-444 (River
-Bend) and ALAB-729 (Three Mile Island). In addition, oortions of these contentions, specified above, are totally redundant of other admitted contentions.
Contention 173. The essence of proposed contention 173 is that ". . . the SER (Section 8.2-1, pp. 8-1/8-2) fails to ana-lyze common causes of failure of all power lines supplying Harris . . .". Applicants oppose admission of this contention on the ground that Mr. Eddleman has not satisfied the criteria of 10 C.F.R. $ 2.714(a)(1) for admission of this late-filed contention and on the further ground that he has failed to sat-isfy the requirement of 10 C.F.R. 5 2.714(b) of adequate basis with requisite specificity.
Contention 173 is clearly untimely and Mr. Eddleman has thus. failed to establish " good cause" for its late admission.
Catawba, supra, CLI-83-19, 17 N.R.C. 1041 (1983). Section 8.2 of Applicants' FSAR contains a detailed analysis of the offsite
-power system for the Shearon Harris plant. Virtually all of this analysis has been available since Amendment 2 was filed on March 31, 1982. The remainder of the relevant information was provided in Amendment 5 which was filed on April 3, 1983.
Thus, Mr. Eddleman has had the opportunity and sufficient in-
- formation for at least ten months to allege any deficiency in l the offsite power system, or the analysis thereof. The SER adds no new information with respect to this system or Appli-cants' analysis thereof other than the Staff's conclusions in sections 8.2.1 and 8.2.4 that the offsite power system for the Harris plant meets General Design Criteria 5, 17, and 18 and, is,-therefore, acceptable. As this Board has held, the l-
issuance of a Staff document containing Staff acceptance of Ap-plicants' analysis does not give Mr. Eddleman a "second shot" at a contention when information sufficient to enable him to frame a contention was previously available. See Memorandum and Order (Ruling on Wells Eddleman's Contentions on the Staff Draft Environmental Statement), August 18, 1983, at 15-16.
In addition to the SER, Mr. Eddleman refers in~his discus-sion of "WHAT'S NEW" to Applicant CP&L's " admission" on January 12, 1984 to the ACRS that CP&L has not ". . . analyzed common cause failure of power lines (other than tornado; probability
-4 estimated at 2 x 10 . . .). This so-called " admission" did not contain new information. CP&L merely reaffirmed what is apparent on the face of the FSAR -- that Applicants have not provided the-NRC with a probability analysis of common mode failures of the seven transmission lines that will serve the Harris plant. CP&L provided the ACRS with an estimate of the probability of a tornado capable of a common mode failure of all seven lines in order to respond to a query by Dr. Kerr dur-ing the ACRS subcommittee review of the Harris operating li-cense application on January 3-4, 1984. Dr. Kerr, while recognizing that such an analysis is not required under Commis-sion regulations, expressed his interest in having some infor-mation on that subject and his opinion of its importance.
)
CP&L, accordingly, provided the information to which Mr.
Eddleman refers during the full ACRS review on January 12, 1984.1/
In addition to the lack of good cause for the untimely filing of this contention, application of four of the remaining factors set forth in section 2.714(a)(1) argues against admis-cion of' Contention 173. First, as indicated above, the NRC Staff has examined the offsite power system and concluded that it meets. applicable URC regulations. Moreover, the ACRS spa-cifically inquired about.the Harris offsite power system, and in its favorable letter of January 16, 1984, the ACRS did not request any additional information on this subject. Thus, Mr.
Eddleman's' interest has been protected in another forum, i.e.,
through the Staff and ACRS review of the Harris operating li-cense application. Moreover, Mr. Eddleman has asserted no technical expertise with regard to this matter and it is un-likely, therefore, that he could add anything meaningful to the l evaluation already provided by the ACRS. Mr. Eddleman concedes that admission of this contention would broaden the issues in this proceeding and would result in at least some delay.2/
1/ A copy of the relevant portions of the transcripts of both the January 3-4, 1984 subcommittee meeting and the January 12, 1984 meeting of the full Committee is attached. See Attachment A.
l 2/ HMr . Eddleman argues here and elsewhere in support of his l SER contentions that discovery on safety contentions is just beginning. This is not true. Discovery was available as of (Continued Next Page)
-Upon balance, applicants believe that the criteria of section 2.714(a)(1) require rejection of Contention 173 at this late date.
Proposed Contention 173 is also inadmissible in that it lacks basis with requisite specificity. As stated above, the thrust of contention 173 is that neither Applicants nor the Staff has analyzed " common-cause failure" of all of the~trans-mission lines which will provide offsite power to the Harris plant.3/ There is absolutely no requirement in the Commission regulations, however, that an applicant perform such an analy-sis. This proposed contention evidences a complete lack of un-derstanding by Mr. Eddleman of the Commission's requirements with respect to, and the nature of, the offsite power system.
As stated in section 8.2.2.1 in the FSAR, the preferred power supply for the Harris plant is any two of the 230 kv lines serving the switchyard and the switchyard itself. The offsite power system is ot a safety related, Class lE system.
(Continued)
September 22, 1982. In February, 1983, the parties agreed to defer discovery on seven contentions (two Joint, one by Dr.
Wilson and four by Mr. Eddleman). Discovery on ten other safe-ty contentions proceeded. Moreover, discovery requests on safety issues are scheduled to end on March 15, 1984.
3/ Among the examples listed by Mr. Eddleman is included
- transformer! fires in the Harris switchyard." Applicants wish to point out that the transformers are not located in the switchyard.
i
IEEE Standard for Preferred Power Supply for Nuclear Power Generating Stations, Std. 765-1983 provides:
4.2 Safety Classification. 'Che preferred power supply is not a Class LE system. Re-quirements for redundancy, independence, separation, application of tae single fail-ure criterion, seismic, and equipment qual-ification, etc., which are associated with
' Class lE installations, do not apply.
The regulatory requirements governing electric power sys-tems, including the offsite power system, are General Design Criteria (GDC) 17 and 18. GDC 17 provides that the safety function for the offsite power system (assuming the on-site system is not functioning) is to provide sufficient capacity and capability to assure that plant systems important to safety perform as intended. GDC 17 then sets forth the means by which such assurance is to be provided. GDC 17 provides, in part:
Electric power from the transmission net-work to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and locat-ed so as to minimize to the extent practical the likelihood of their simulta-neous failure under operating and postu-
. lated accident and environmental condi-
,tions. A switchyard common to both circuits is acceptable. Each of these cir-cuits shall be designed to be available in sufficient time following a loss of all onsite alternating current r.ower supplies and the-other offsite electric power cir-cuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant acci-dent to assure that core cooling.
containment integrity, and other vital safety functions are maintained. (Emphasis added)
'Thus, GDC 17 requires two independent circuits designed to min-imize to the extent' practical the likelihood of their simulta-neous failure.
In section 8.2 of the FSAR, Applicants have provided a de-tailed explanation of the design of the offsite power system for Harris which, as the Staff has concluded, satisfies the re-quirements of GDC 17. Indecd, the Harris offsite system far
. exceeds those requirements. Section 8.2.1.1, states that the Harris plant will be served by a total of six 230 kv transmis-sion lines. In fact, as CP&L advised the ACRS, a total of seven 230 kv lines will ultimately be available to the plant.
Thus, the Harris.offsite power system significantly exceeds the requirements of GDC 17 in this respect.
In section 8.2.1.1 of the FSAR, Applicants go on to de-scribe-the characteristics of the offsite system design which provide' assurance of adequate capacity and capability to permit proper functioning of all safety-related equipment at the plant. These include design to withstand loadings due to maxi-mum observed natural occurrences within the CP&L service area
' including lightning, ice storms and tornados. Moreover, Appli-
. cants have met the separation and independence requirements of GDC 17 by demonstrating in section 8.2.1.1 that a single trans-miss'i on structure failure could remove no more than one other 1
transmission line from service. In addition, section 8.2.2.1 states that the CP&L transmission system is a part of the East-ern United States Power Grid which provides a high degree of reliability and availability to the CP&L system. In accordance with GDC 17, FSAR section 8.2.2.1 provides an analysis to dem-onstrate the adequacy of the transmission of CP&L and neighbor-ing utilities to withstand:
a) Sudden loss of the entire generating capability at any plant.
b) Sudden loss of any large load or load center.
c) Sudden loss of all lines on a common right-of-way.
d) ~
The delayed clearing of a three-phase fault at any point on the system to breaker failure.
e) .The outage of the most critical transmission line caused by a three-phase fault during an outage of any other critical transmission line.
Based upon the information provided in the FSAR, the Staff has concluded in section 8.2 of the SER that the Harris offsite power system satisifes GDC 17 as well as GDC 5 and 18. Mr.
Eddleman has not contended that the offsite system does not comply with GDC 17 or any other NRC regulation. Mr. Eddleman does make'the broad unsupported statement that the SER's loss of Offsite Power Analysis at C-13 of the SER " depends on the accuracy of~ analysis in Section 8.2 of the SER" and that "that r-analysis is faulty." In view of the detailed analysis in the FSAR which is accepted in the SER, Mr. Eddleman was obliged to state with specificity what in that analysis is inadequate. He
points only to the purported " failure" to assess the probabilities of common mode failures which, of course, is not a requirement under the regulations. Mr. Eddleman has not made the requisite demonstration.
For the reasons set forth above, therefore, proposed Con-tention 173 must be rejected as untimely and as lacking basis with the requisite degree of specificity.
Contentions 174, 175, 176 and 177. Mr. Eddleman has pro-posed four new contentions on the adequacy of the Staff's Part 100 (seismology) analysis in the Shearon Harris SER. Appli-cants do not understand there to be any distinction among Con-tentions 174 through 177, each of which asserts that in view of a U.S. Geological Survey (USGS) letter of Novembar 18, 1982, referenced in Appendix F to the SER, the Staff is required to analyze the impact on the Harris facility of a Modified Mercalli X intensity earthquake occurring at the site or closer to the site than Charleston, South Carolina.
Applicants object to the admission of Contentions 174 through 177. These contentions are untimely and, accordingly, fail to satisfy the " good cause" prong of the five part stan-dard for admitting late-filed contentions. See 10 C.F.R.
$ 2.714(a)(1)(i). Applicants also oppose the admission of these seismology contentions because there is no evidence that Mr. Eddleman can make a valuable contribution to the record on this issue, which is being evaluated on a generic basis by the
Staff, and because these contentions will significantly expand the scope of the Harris plant operating license proceeding.
The Shearon Harris Draft SER, issued in January, 1983 and served on the. parties, alerted Mr. Eddleman to the following facts:
(1) The NRC Staff holds the position tl:at the rela-tively high seismic activity within the Coastai Plain Province in the vicinity of Charleston, S.C,, including the 1886 MM Intensity X earthquake, is related to unique tectonic structure. Inerefore, in the context of the tectonic approach, an MM. Intensity X earthquake should not be assumed to occur anywhere else. Draft SER at 2-36,
'2-41.
(2) Lacking definitive evidence, this NRC conclusion was based to-a. great extent on advice from the USGS. Id.
(3) The Charleston, S.C. region is presently under intensive seismological investigation by USGS. Id. at -
2-36, 2-40.
(4) As a result'of these studies, a great deal of in-formation has been obtained, but the source mechanism of the seismicity still is not known. Id. at 2-36.
(5) Many working hypotheses are described in the Virgil C. Summer SER (1981). Id.
(6) Some of these theories postulate that the Charleston earthquake of 1886 could recur in areas of the Piedmont and Altantic Coastal Plain in addition to the epicentral area. Id.
(7) The USGS clarified its position in a November 18, 1982 letter frorc James F. Devine to Robert E. Jackson, NRC. Id.
(8) The maximum earthquake "that shall be considered to occur near the Harris site" is not the Charleston 1886 earthquake, but an event with a maximum intensity of VII (MM) or a maximum magnitude of 5.3(mb). Id.
In sum, all of the facts which are the basis for Conten-tions 174 through 177 were stated very clearly in the Shearon Harris Draft SER one year ago. On the basis of the Draft SER, Mr. Eddleman was in a position to challenge the Staff's posi-tion not to use the Charleston earthquake in its tectonic prov-ince analysis of the maximum earthquake at the Harris site.
While the substance of the November 18, 1982 UGSC letter was not provided in the Draft SER, Mr. Eddleman clearly was on no-tice of the letter's existence and, more importantly, of the ongoing investigation of the significance of the Charleston earthquake in NRC's seismology analyses.
In addition to the notice provided to Mr. Eddleman of this issue in the Draft SER, there are several other public docu-ments which would have alerted Mr. Eddleman to the existence of this. issue significantly prior to issuance of the final SER. A public meeting to discuss the Charleston Earthquake in the
u t 4 ..
context of eastern seismicity was held by NRC and USGS at which interested members of the public were invited to ask questions
- and' raise concerna. This meeting was noticed in the Federal Register. See 47 Fed. Reg. 53538 (Nov. 26, 1982). At this
- public' meeting,-copies of the Nov. 18, 1982 USGS letter and of a Nov. 19, 1982 Memorandum from the Executive Director of Op-erations to the Commissioners on this subject were available.
See Attachment B (transcript pages from meeting) and Attachment C (Nov. 19, 1982 EDO letter with attached Nov. 18, 1982 USGS letter).
Furthermore, the Catawba SER, issued in February, 1983, includes in Section 2.5 a discussion and excerpt from the November 18, 1982 USGS letter. The Catawba SER also explains that the "USGS clarification represents not so much a new un-darstanding but rather a more explicit recognition of existing uncertainties with respect to.the causative structure and mech-anism of the 1886 Charleston earthquake." Catawba SER at 2-24.
The information in the Catawba SER is the same information, some of which is referred to by Mr. Eddleman, contained in Sec-tion 2.5 and Appendix F of the Shearon Harris SER. Thus there is nothing new in the Shearon Harris SER, nor has the Staff's position on the use of the Charleston earthquake in its seismic analyses changed since the issuance of the Draft SER.4/
4/ :While the Draft SER references the fact that the Staff
- will respond to the USGS' letter in the final SER, Mr. Eddleman (Continued Next Page)
O As a participant in an NRC licensing proceeding, Mr.
Eddleman has an obligation to familiarize himself with publicly available information, such as the information described above, and to promptly formulate contentions that are based on this information. Catawba, supra, 17 N.R.C. at 1048. As stated above, "the institutional unavailability of a Licensing-related document (such as the final SER] does not establish good cause for filing a contention late if information was available early enough to provide the basis for the timely filing of that con-tention'." Id. Mr. Eddleman has failed to meet his obligation to file timely contentions on the Charleston earthquake. Con-sequently, he has not established " good cause" for the admis-sion of Contentions 174 through 177. See 10 C.F.R. 5 2.714(a)(1)(1).
-Moreover, recognizing that there are other factors to be considered in_ evaluating the admissibility of late-filed con-tentions, Applicants believe three of the other four factors weigh heavily against admiesion of Contentions 174 through 177.
There is no reason to expect that Mr. Eddleman's participation (Continued) was obligated to formulate contentions which challenged the Staff position at the time of the Draft SER's issuance. In the event.the final SER changed a Staff position taken in the Draft SER, then Mr. Eddleman's claim of new information would be ap-propriate. However, as discussed above, this was not the case in this instance.
. will-assist in developing a sound record on this issue. See-10 C.F.R. l 2.714(a)(1)(iii). Mr. Eddleman is not a geologist,
- nor does he' contend that he will have any expertise available to him that will contribute to the extensive record on this ge-neric issue that has been and continues to be developed by the Staff. Moreover, the Staff's continuing in depth consideration of this issue, which has applicability to all Eastern seaboard ,
nuclear power plants,.ought to protect Mr. Eddleman's interest, albeit outside the forum of this hearing. See 10 C.F.R. 5 2.714(a)(1)(ii). See Shearon Harris SER at 2-30. In addition,
= as Mr. Eddleman admits,.the admission of Contentions 174 through 177 would broaden the issues now being litigated.
There'is not now a seismology contention admitted in'this pro-ceeding. 'Sem 10 C.F.R. I 2'714(a)(1)(v).
. Consequently, liti-gation of this complex issue could and probably would be very time consuming.
In summary, under the test set forth by the Commission in
.the Catawba decision, CLI-83-19, Contentions 174 through 177 should be rejected by the Board.
Contentions 178 and 179. Contentions 178 and 179 chal-lenge1the adequacy.of the Transamerica DeLaval, Inc. (TDI) die-sel generators that will provide backup emergency power at the
- Harris site. The basis cited by Mr. Eddleman for these conten-tions-is hiscreceipt on December 19, 1983, of a December 1, 1983-letter from NRC to TDI that was served on the Harris 4
, - . . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ __ ._ _ _ . _ _J
l proceeding participants. Mr. Eddleman claims that before his receipt of the December, 1983 letter, he "had no solid basis to relate the TDI failure pattern directly to Harris." SER Conten-tions at 8.
Applicants object to the admission of Contentions 178 and 179. These contentions are not timely filed; consequently, Mr.
Eddleman has not established " good cause" for their late admis-sion at this juncture. Catawba, supra, CLI-83-19, 17 N.R.C.
1041 (1983); 10 C.F.R. 5 2.714(a)(1)(1). In addition, Mr.
Eddleman has not established any expertise in this field, the issues are being fully addressed by the Staff, and Contentions 178 and 179 unquestionahiy would broaden the issues already being litigated in this proceeding. See 10 C.F.R. $ 2.714(a)
(1)(11)(111) and (v).
The fact that Shearon Harris will use TDI emergency diesel generators is reflected in the Shearon Harris FSAR at Table 3.10.1-2, which was part of the FSAR as originally submitted in June 26, 1980. Table 8.3.1-8 of the FSAR indicates that the Harris diesel generators are the same model as the diesels being used at Grand Gulf Station Unit 1, at which some of the significant problems in this equipment have been discovered over the past two years. This Table was part of Amendment 5 to the FSAR, dated April 12, 1983.
More importantly, tne potential applicability of certain of the generic TDI diesel generator problems to the Shearon t
/
Harris TDI diesels is unequivocally clear from numerous public documents of which Mr. Eddleman should have been aware. The following documents are among thosa on the subject of TDI die-sel generators available in the Public Document Room (PDR):
1.* July 30, _1981 Part 21 notification from TDI to NRC (Inspection and Regulation) regarding potential valve spring problem in TDI diesel generators; explicit reference is made to-Harris diesels
- 2. December 9, 1981 Part 21 notification from TDI to NRC (Inspection & Regulation) regarding potential problem with TDI diesel gen-erator governor lube oil cooler assem-bly; explicit reference is made to Harris diesels
- 3. March 19, 1982 Part 21 notification from TDI to NRC (IE) regarding potential problem with TOI diesel generator sensing line be-tween the starting air storage tank and the starting air compressor; ex-plicit reference is made to Harris diesels 4.* May 19, 1982 Letter from CP&L to NRC (Region II) transmitting Harris Unit 1 Interim Re-port on potential defect in pressure sensing line of TDI diesel generators
- 5. June 23, 1982 Part 21. notification from TDI to NRC (IE) regarding potential problem with TDI. diesel generator governor drive coupling; explicit reference is made to Harris diesels 6.* . October 28, 1982 Part 21 notification from TDI to NRC (IE) regarding potential defect in TDI diesel generator engine piston skirt casting; explicit reference is made to Harris diesels 7.* July 14, 1983 Letter from CP&L to NRC (Region II) transmitting fourth Harris Unit 1 In-terim Report on TDI diesel generator valve spring problem
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _]
.. ' I
- 8. August 30, 1983 IE Information Notice No. 83-58 re-garding TDI diesel generator crank-shaft failure; explicit reference is made to Harris diesels
- 9. Septemb$r 21, 1983 Part 21 notification from TDI to NRC (IE) regarding potential problem with TDI diesel generator engine mounted fuel oil line; explicit reference is-made to Harris diesels
'10.* December 14, 1983 Letter from CP&L to NRC (Region II) transmitting second Harris Unit 1 In-terim Report on piston skirt problem in'TDI diesel generators
- ' Astericked documents:were available in the Shearon Harris
. docket, 50-400. The other documents were located in other fa-cility1 dockets or, in the case of-document #8, an NRC Informa-tion Notice, in a separate IE file. However, all of these doc-uments are readily accessible by computer using TDI as the
~ basis for such a document search.
The potential applicability.of the identified TDI problems Lias the. Harris site is unequivocally clear in all- of these docu-ments.. Thus,-while Mr. Eddleman personally may not have been
- aware of this information,. a " solid bas'is tx) relate the TDI failure pattern to Harris" has existed for a long time. Conse-
.quently, there is no " good cause" for submission of these con--
tentions at this late date.
In' Contentions 178'and 179, Mr. Eddleman also refers to
. quality ' assurance problems with the TDI diesel generators. The
'following seven inspection' reports of the TDI facility in
-Oakland,. California are publicly available documents:5/ 79-1
'jbf :These documents can be located in the PDR under the docket number;for QA vendor inspections of TDI.
iii. . . . . .. - .
s (March 20, 1979); 80-01 (Jan. 22, 1981); 81-01 (May 27, 1981);
81-02 (Sept. 18, 1981); 82-01 (April 15, 1982); 82-02 (Dec. 8, 1982) and 83-01 (Oct. 3, 1983). Had Mr. Eddleman taken an in-terest in the subject of TDI diesel generators at an earlier date, he readily would have discovered these reports, which discuss the QA problems associated with the TDI diesel genera-tors that are referred to by Mr. Eddleman in Contentions 178 and 179.
In addition to the absence of good cause for the late sub-mission of Contentions 178 and 179, three of the other four factors that must be balanced in deciding whether to admit late-filed contentions support rejection of these contentions.
See 10 C.F.R. $ 2.714(a)(1). The generic consideration of this issue by the Staff will protect Mr. Eddleman's interest. See 10 C.F.R. 5 2.714(a)(1)(ii). There is no reason to believe the admission of.these contentions in this proceeding will in any way augment the NRC's continuing oversight of this issue, which is also the subject of a TDI owners' group verification pro-gram. See 10 C.F.R. 5 2.714(a)(1)(iii). Certainly, the admis-sion of Contentions 178 and 179 will significantly broaden the issues in this proceeding, with a corresponding increase in proceeding time to resolve all outstanding issues. See 10 C.F.R. $ 2.714(a)(1)(v). Compare Mr. Eddleman's ridiculously circular assertion that the admission of Contention 179 "will not significantly broaden the issues since so many safety i
issues are already admitted." SER Contentions at 9. Thus, al-though the issue of TDI diesel generators is not being pursued by any other intervenor in the proceeding, see 10 C.F.R. 5 2.714(a)(1)(iv), a balancing of the five factors for de-termining whether to admit late-filed contentions clearly es-tablishes that late-filed Contentions 178 and 179 should not be admitted.
Contention 180 cites the SER as a basis for the allegation that the ability to isolate the steam generators within thirty minutes in the event of steam generator tube ruptures is not established. Joint Contention VII, which is sponsored by Mr.
Eddleman and which has been admitted by the Board, states in part:
Applicants have failed to demonstrate that the steam generators to be used in the Harris Plant are adequately designed and can be operated in a manner consistent with the public health and safety and ALARA ex-posure to maintenance personnel in light of
. . . (4) existing tube failure analyses.
Memorandum and Order (Reflecting Decisions Made Following Prehearing Conference), LBP-82-119A, 16 N.R.C. 2069, 2077 (1982). The allegation raised in Eddleman 180 goes to Appli-cants' steam generatar tube rupture analysis and therefore is encompassed by the above-quoted portion of Joint Contention VII.g/ Consequer.cly, Applicants submit that Eddleman 180 s/ Applicants' counsel (Baxter) discussed this matter with Mr. Eddleman on January 30, 1984. Mr. Eddleman authorized Ap-(Continued Next Page)
- should be rejected as redundant. See id. at 2090, 2095
~
(Eddleman 1, 2,~29 and 30 rejected in whole or in part as re-dundant of Joint Contentions).
Contention 181 contends that the SER reveals that the Staff's review of control room design requirements is not com-plete due to Applicants not providing a program plan showing-how.each detailed control room design review (DCRDR) activity
- was accomplished and failing to address several areas and items in violation of NUREG-0737, Supplement 1 requirements for docu-mentation.7/ SER Contentions at 10. Contention 181 as pro-posed by Mr. Eddleman relies on a number of erroneous conclu-sions drawn from the SER and is, therefore, lacking in technical basis and should be rejected by the Board.
First, the lack of detail in the SER is not due to any delay or default by the Applicants (there is none), but is sim-ply a result of the timing involved -- the DCRDR was completed ltoo close to the_ issuance of the SER to allow the DCRDR to be (Continued) plicants to report that he agrees that Eddleman 180 is covered by.and redundant of part (4) of Joint Contention VII.
7/ Mr. Eddleman previously proposed control room design re-view contentions'on January 8 and July 2, 1983. See Appli-
~ cants' Response to Proposed Contentions on the Detailed Control Room Design Review (DCRDR) Proferred by Intervenor Wells Eddleman, July 29, 1983. Eddleman 132(C)(II), on this subject, was admitted by the Board. Memorandum and' Order (Ruling on Wells Eddleman's Proposed Contentions Concerning Detailed Con-trol Room Design Review . . . ) at 8 (Oct. 6,. 1983).
U i. .
fully addressed therein. Advisory Committee on Reactor Safe-guards (ACRS) 285th meeting, January 12, 1984, Transcript Vol.
I at 186. The Staff plans to issue a supplement to the SER on this issue.- Id.; see also SER at 18-2.
Second, Contention 181~ states that the requirements of NUREG-0737,' Supplement 1, have not yet been met. In fact, Ap-plicants have addressed all relevant requirements, including the items referred to by Mr. Eddleman, and have continued to pursue those areas where questions may have arisen. Specifi-cally, Mr. Eddleman cites the lack of a program plan showing how each DCRDR activity was accomplished and failure to address three areas set forth at page 18-2 of the SER. Those items are as follows:
(1) a description of the process used to define control and display requirements and the ,
basis for determining how control and dis-play arrangements were made; also, a de-scription of how the applicant will verify that the selection and arrangement of con-trols and displays and other equipment re-quired during emergency operations will en-able operators to effectively execute plant emergency operating procedures.
(2) for items and areas in the control room not yet reviewed, an assessment and proposed corrective actions for HEDs identified at least 6 months before licensing (3) acceptable corrective actions and an imple-mentation schedule for HEDs identified by the NRC audit (these will be addressed in more detail in the audit report that will be forwarded to the applicant when it is completed.
Contrary to Mr. Eddleman's Contention, both the program plan for accomplishment of DCRDR activities and item (1) above are addressed in a response to information requested by the NRC Staff after its DCRDR audit in August 1983. The information was provided to the NRC by letter to Mr. H. R. Denton dated September 27, 1983, a copy of which was sent to Mr. Eddleman.
See Attachment D. If Mr. Eddleman found anything inadequate in this submittal, it was incumbent of him to say so. Catawba, supra. Not only did Mr. Eddleman not make a timely specific complaint, but his newly proposed Contention 181 utterly fails to address the specifics contained in the September 27 letter.
The remaining two items set forth above call for action which necessarily cannot be taken until some future period - " items and areas in the control room not yet reviewed," and corrective actions and implementation schedule for human engineering discrepancies (HEDs) identified by the NRC Staff Audit, which has not been published yet. See Memorandum and Order (Ruling on Wells Eddleman's Proposed Contentions Concerning Detailed Control Room Design Review . . .), October 6, 1983, at 12, 13 (the fact that the process is not complete yet provides insuf-ficient basis for a contention); Memorandum and Order (Ruling on Wells Eddleman's Proposed On-Site Emergency Planning Conten-
, tions), November 1, 1983, at 11 (Contention 143). Applicants have addressed every aspect of DCRDR requirements to date and the NRC Staff has sought some further information from
Applicants but has never challenged the adequacy of any of Ap-plicants' control room review. SER at 18-2.
It should be noted that CP&L actually conducted its DCRDR from April 1980-to January 1981. A summary report was issued January 23, 1981. SER at 18-1. Subsequently, since the SHNPP control room and Main Control Board (MCB) were not yet built,
' Applicants proceeded with changes to both control room and MCB layouts to improve human engineering aspects. All of this was prior to the issuance both of NUREG-0700 " Guidelines for Con-trol Room Design Reviews," dated September, 1981, which pro-vides guidance for the review, and NUREG-0737 Supplement 1 (December 1982). Id. CP&L submitted the summary report in December 1982, and provided additional information on the pro-gram plan to the NRC in June 1983. SER at 18-1. Finally, the NRC Staff conducted its DCRDR audit on August 15-19, 1983; how-ever, the circumstances of the Staff review were unusual in that Applicants had already conducted their review prior to is-suance of the NUREGs and then made changes thereto, id., and both the control room and MCB were not complete but were still under construction. Therefore, the Staff was not able to con-duct its review of a finished plant as the guidance of NUREG-0700 would direct. Again, however, it must be emphasized that no discrepancies were found on which the Applicants had not already initiated corrective action.
s Thus, Applicants have taken the initiative and conducted their human engineering review, including significant changes thereto, on their own with no official guidance or directives from the NRC. There has never been a challenge to the adequacy of Applicants' review -- only requests for further information, SER at 18-2, which is to be expected in a plant still under construction. Mr. Eddleman's assertion that Applicants fail "to assure adequate human factors design implementation at Harris" and indeed the entirety of proposed Contention 181 is totally without basis and should not be admitted in this pro-ceeding.
, Conclusion For the reasons stated above, Applicants urge the Board to deny the Eddleman Motion, and to reject all of the new SER Con-tentions.
Respectfully submitted, Thomas A. Baxter, P.C.
Deborah B. Bauser SHAW, PITTMAN, POTTS & TROWBRIDGE 1600 M Street, N.W.
Washington, D.C. 20036 (202) 822-1000
Richard E. Jones
' 'Samantha Francis Flynn Hill Carrow CAROLINA POWER & LIGHT COMPANY P.O. Box 1551 Raleigh, North Carolina 27602 >
Counsel for Applicants Dated: February 6, 1984 Q
\
e UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
CAROLINA POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL MUNICIPAL POWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant, Units 1 and 2) )
CERTIFICATE OF SERVICE I hereby certify that copies of " Applicants' Answer to
, Wells Eddleman's Motion for Further Deferral of Parts of Contention 107 and to Wells Eddleman's New Contentions and Amended Deferred Contentions in Response to Staff SER" with attachments were served this 6th day of February, 1984, by deposit in the U.S. mail, first class, postage prepaid, to the parties on the attached Service List.
Thomas A. Baxter, P'C.
--.----_...---.J
,j .
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
CAROLINA. POWER & LIGHT COMPANY ) Docket Nos. 50-400 OL and NORTH CAROLINA EASTERN ) 50-401 OL MUNICIPAL. POWER AGENCY )
)
(Shearon Harris Nuclear Power )
Plant, Units 1 and 2) )
SERVICE LIST James L. Kelley, Esquire John D. Runkle, Esquire Atmic Safety and Licensing Boarri Conservation Council of North Carolina U.S. Nm lear Regulatory Camission 307 Granville Road Washington, D.C. 20555 Chapel Hill, North Carolina 27514 Mr. Glenn O. Bright M. Travis Payne, Esquire Atmic Safety and Licensing Board Bielstein atxi Payne U.S. Nuclear Regulatory Ca mission P.O. Box 12607 Washington, D.C. 20555 Raleigh, North Carolina 27605 Dr. Janes H. Carpenter Dr. Richard D. Wilson Atm.ic Safety and Licensing Board 729 Hunter Stru t U.S. Nuclear Regulatory h iasion Apex, North Carolina 27502 Washington, D.C. 20555 Mr. Wells Eddleman Charles A. Barth, Emq" ire 718-A IrMa11 Street Janice E. Moore, Esquire Durban, North Carolina 27705 Office of Executive Iagal Director U.S. Nuclear Regulatory Oannission Richard E. Jones, Esquire Washington, D.C. 20555 Vice President and Senior Counsel Carolina Power & Light Canpany Docketing and Service Section P.O. Box 1551 Office of the Shu.etary Raleigh, North Carolina 27602 U.S. Nuclear Regulatory C = iasion Washington, D.C. 20555 Dr. Phyllis Intchin 108 Bridle Run Mr. Daniel F. Read, President @ 5=1 Hill, North Carolina 27514 CHANGE /EIP 5707 Waycross Street Dr. Linda W. Little Raleigh, North Carolina 27606 Governor's Waste Managenent Board 513 AL % rle nnilrli ng 325 North Salisbury Street Raleigh, North Carolina 27611
'F -
Bradley W. Jones, Esquire U.S. Nuclear Regulatory Ccmnission Region II 101 Marrietta Street l
Atlanta, Georgia 30303 Ruthanne G. Miller, Esquire Atcmic Safety and Licensing Board Panel U.S. Nuclear Regulatory Cm=4 anion Washington, D.C. 20555 Mr. Robert P. Gruber W *ive Di m. h Public Staff - NCUC P.O. Box 991 Raleigh, North Carolina 27602 9
t .
140
,*51b { Attachment A
(
1 MR. KERR: Does the committee want to hear a i
2 presentation on this, or do you want to ask questions? I'm j l
3 -
going to suggest, Mr. Prunty, that we ask for questions and 4 l not ask for your formal presentation, if that is okay with l 6
you. Are there questions on AC/DC System Reliability?
6 ffe did explore this on the Subcommittee. l l
7 MR. EBERSOLE: What was that extremely low number 8
- heard, about the AC power?
9 MR. PRUNTY: That was a number that considered l 10 I independent failure of all seven lines, coming into the !
11 switchyard. It did not consider the common event, such i t
as a major tornado or something. That has since be 13 reevaluated and a probability number of something on the i i.
14 order of 2 times 10~4 which is one occurrence approximately 15 I l every 5,000 years, has been given as a ecmmon mode failure l
16 [
i i
type mechanism in the switchyard. i l 17 ' i l
l, i
The other one, though, was all the sequential gg i
seven lines failing for independent causes, one after the j oth er .
.Y f
MR. KERR: Any further questions? i 21 I (No response.) l l
M ,
l I Thank you, Mr. Prunty. t
! 2 I I
I MR. OKRENT: Just a minute. Is 2 times 10"4 34 l
{ consistent with experience in the region for a less of all u l l
offsite power?
l
- 141
'51b2 1
1 l MR. KZRR: I haven't look at the region as such, 3 but the number sounds low to me. I.
3 MR. ZIMMERMAN : Clarification, that's the 1
4 tornado probability occurrence?
i 8 +
MR. PRUNTY: That's correct. That was an event e .
that caused damage in the switchyard. It would completely !
1 destroy the shipysrd.
a MR. KERR: But the total loss is perhaps somewhat ,
9 larger than that.
10 MR. OKRENT: It would have been helpful to get the 11 full answer, I guess, rather than from only one of the common 12 cause contributors.
18 I MR. KERR: Further questions? '
I 14 (No response.)
IO Thank you, Mr. Prunty.
N l That takes c'are of AC/DC. Emergency Planning l l -
1 I is the next item. Since tha Applicant has emergency plans l I
18 for three other reactors, I'm going to suggest, if you like, ,
that we handle this with questions, but a presentation is 20 available.
MR. CKRENT: I have a question on DC. If they 1
21 1 '
lose all AC, how long do they expect their oc to survive?
23 MR. ZIMMERMAN: Mr. Prunty will address that.
24 MR. PRU:'TY : :# we lose AC, how long with the .7C 28 l system survive? ,
I i
142
.ilb) .
p 1 MR. OKRONT:
Survive and perform vital functions.
1 1
MR. PRUNTY: 1 Our evaluation of this considered l 3 i design margin in the battery and the use of the design basis 1 4
event loads, as per the FSAR.
i "he number that was arrived
& f f at was on the order of six hours before battery failure, 1 6
l The actual load, during a blackout, would be less than those 1 7
required to mitigate the effects of a design basis event, so 8
we are semething in excess of six hours.
9 MR. KERR: Further questions.
10
, (No response.)
11 Thank you, Mr. Prunty.
12 i
I don't see any requests for a full :,cale presen- !
13 l i
nation.
! Are there questions about emergency planning? f I4 l Mr. Moeller? 1 II l !
MR. M0ELLER: l 16 ! or Ohls i s more a question or a ccament {
suggestion to the Staff.
In terms of emergency planning 17 i l
and the impacts of major accidents cne can read the final '
IO {
i l environmental statement, which has the best information, j
! I' on page 5-31, in the final environmental statement, I wanc 20 to offer a suggestion.
i It says if a certain accident occurs,j
- 1 there will be so many cancer deaths a ! in t he surrounding r.c.eulaf tien. '
And my suggestion is that the Staff, who or which, I
23 I- f
}'
whichever word you use, does a good job on this -- that they i
24 consider adding the word excess cancer deaths or additiena'. i i
as
{
\ cancer deaths because the number you are giving would be ne !
\
<t
MM8,syl 6 .
267 l 1
MP.. KERR: I'ay we continue?
2 MR. BERLY:
'I'm Bcb Berly, Manager of T mnsmission 3
and Communications Planning Section.
I will talk to you 4
about the off-site AC power source, starting with the big 5
picture and moving to the Harris site. Twenty-six transmissio n
8 ties give the Fastern CP&L 7
service aret a grid strength of most of North America and part of Canadu. In thecompany area, 8
thousands of miles of well-planned, strongly designed 230 Kv 9
transmission lines operate with high reliability.
10 There has never been a loss of availability of the 11 total'230 Kv grid, nor has there ever been a total loss of 12 availability of 230 Kv power to any nuclear or coal-fired 13 plant served by this grid.
14 tslice,j 15 Studies show complete adequacy with five lines
! 16 serving Harris, but plans call for construction Of ::cven lines, 17-as shown on this slide. I 18 :
MR. KCRR: Excuse me, 18 how long has the 230 Kv grid
{
been in existence?
i I'm trying to get an idea of how long.
20 1
MR. BERLY: We built the first one 18 years ago.
21 -
w,*ve got over 2500 miles of 230 now.
So we're talking about 18, years of 23' history over w.wh there has been no loss of power to any
- ~ 24 station?
28 M;. , 3 ,,7 r
' There's been no loss of total 230 KV m
f sy2 k
268 i to any station served by 230, and that's about 10 plants.
2 MR. KERR: Thank you.
3 MR. BERLY:
Any single line is capable of supplying 4
all emergency safety equipment.or startup load requirements.
5 As you see, we have seven lines.
Seven lines radiate from 6
Harris in all directions. Parallelism is kept to a minimum.
7 Cape Fear steam electric plant is located approximately 7 miles a from Harris.
It's a little, hazy over here. but you see on 9
the lef thand side of the drawing Cape Fear plant to and it can furnish suffi.-ient power for Harris safety-related equipment.
11 In addition to fcssile-fired units at Cape Fear, 12 there are four IC turbines, 13 two of which can be black started.
.t5 tide.)
14 An inverted breaker and a half, or double breaker ,
15 termination --
16 MP. KERR:
Excuse me, those IC turbines at Cape i 17 Fear, --
18
.IP . m
~Y:
Yes, sir, four turbines at Cape Fear. 1 19 Two can be black started, two can be started in 20 minutes. !
20 l An .
2rt ed breaker and a half, a double breaker 21 termina tion , .s
. .i e d for the generator, and all line termina-tions in the 22 2:
.i switchyard.
n r- .
of two 230 Kv main buses is provided.
24 Breakers ar-
.ent pole tripping, and their dcsi !.
..,. 25 permits full.
.e open/ closed operations without operatio n
syj s 369 1
of the air compressor. Redundancy is provided in protective 2
relay and control circuits, and in battery power capability.
3 This concludes my formal presentation.
4 MR. KERR: I believe I read that there were two 5
separate batteries available for switchgear operation in your 6 switchyard.
7 MR. BERLY: There are two batteries available for 8
each control circuit in those ones there.
8 MR. KERR: Are those batteries seismically qualified?
10 MR. BERLY: I can' t answer that.
11 MR. ZIMMERMAN: We can find that out.
12 MR. KERR: Are there other questions?
13 l (No response.)
14 Incidentally, in your decision on emergency power, 15 did you attempt to calculate or estimate the probcbility of 16 loss of off-site -- all off-site power at Harris?
4 17 MR. BERLY: With respect to transmission, yes, sir.
18 MR. KERR: What is the probability number at which 19 you arriv'd? e
' E l
MR. PERLY: For all seven 230 KV lines, it's 1. - 3 21 times 10 -33 ,
22 MR. KERR: Wait a minute. 1.53 times 10 --
23 only three significant figures?
- 24 (Laughter.)
- 10
-33 per year?
MR. BERLY: This is related to loss because
- Cy4 i
270 1
independent causes.
2 MR. KERR:
3 I'm interested in the probability cf loss.
4 I just wondered if you tried to calculate it, if any of your decisions were based on an attempt to estimate that 5
number.
6 MR. BERLY:
7 The best we considered to be such a high measure of reliability, so it really couldn't determine our design. '
9 MR. KERR:
10 I recognize that the 10~33 has nothing in the realm of possibility. I'm trying to find out whether 11 you a ttempted 12 to estimate a real number and used it in your decision-making process at all.
13 l' t MR. BERLY: No, sir.
14 MR. KERR: i 15 I'm a little puzzled that you did not, since I think --
16 MR. BERLY:
17 I was going to say we have information relative to outage records on the lines, which we did use in i
.I 18 designing the system, and that indicates less than one hour 19 interruption per hundred niles per year in the 230 .
20 MR. KERR: I raised the question because 1 think 21 it's important tha t you have AC available for many of the 22 things that you might need to do in an emerge.ncy. I don't 23 think that's news to you or anybody else.
{ It therefore seems to me some understanding of what b ~
the reliability of the off-site power is is needed, in oder
a2' , '
271 I 7 that you make a decision as to whether the on-site power,
/ 2 emergency power reliability, is appropriate.
3 MR. BERLY, 4
We consider it highly reliable in that we do radiate in different directions.
4 MR. KERR:
What would you be willing to accept 6{
as a failure probability for all off-site power? Once in 10 years, once in 100 years, once in 1000 years?
- . 8
- y MR. BERLY
} I guess we're talking maybe in the 8
range of once in 100 years or 200 even. It's very remote.
10 MR. KERR: Okay.
11 t
If you, for example, are willing to talk about once in 100 years being off for, say, an hour, 12
-that means that you need a fairly reliable on-site source.
I8 Did you u.e numbers like that
!, to decide on the appropriate t
reliability of your.on-site source?
MR. BERLY: Let me back up and ask -- are we 16 saying the probability of all seven lines being out?
17 MR. KERR:
18 I don't care how you get the AC power ,
4 to your plant.
tg >
i MR. BERLYs 21 All we use is one line.
4 MR. KERR: I'm saying have you estimated the I
probability that you wauld lose all off-site sources of AC, 22
'tso that you would have to use your emergency on-site source?
MR. BERLY: No, sir.
24
-MR. KERR:
2 l It would seem to me that a utility would want to do this,
.( given the importance of having AC available.
I e
)
sy6 L r 272 1
.- But if you haven't, you haven't. It certainly is not a 2
requirement of any of the regulations. I am looking toward 3
the day when utilities don't find themselves restricted by 4
the regulations but make decisions basad on what they think 8 ought l to be done, and I'm sure you do that.
l 6 But it seems to me this is a pretty crucial area.
7 MR. BERLY:
We considered it so remote, I guess we 8
didn't consider it as a possibility.
- 8 MR. KERR:
A remoteness of once in 100 years is 10 fairly high compared to, for example, the calculations that 11 EPRI have made and the numbers that they have used. Have 12 you looked at their numbers?
13
) '
MR. BERLY: ' No , sir.
1 14 i MR. KERR:
Well, I would urge that you do.
15
.. MR. BERLY: Shall I go on?
16 MR. KERR: Yes , sir.
cnd 8 17 l
18 19 20 21 22 23 24 25
I' 1
C19,syl
\ 273 1
MR. IIKiERMAN: Dr.
Kerr, you asked about the 2
seismic qualification of the batteries. They're not seismically 3 qual $.fied.
4 MR. KERR: Thank you.
5 MR. PRUNTY:
Good morning, my name is Bob Prunty, 6
I'm principal electrical engineer in the Harris Engineering 7
Section out at the Harris site.
8 (Slide.)
9 I'm going to address the on-site AC and DC distri-10 bution systems and follow that up with a discussion of station 11 blackout.
12 Shearon Harris on-site power system consists of 13 multiple independent power sources, an arrangement of inter-14 connecting busses and redundant independent safety trains 15 that insure max:. mum reliability and diversity.
16 The safety-related portion includes two diesel 17 generators, two 6.9 KV safety busus, several .280 volt buses, 18 two 125 volt '. C
- .a tteries and their associa ted charges which 19 supply two DC buses and four 120 volt AC uninterruptable buses.
20 The - . .
6.9 Kv emergency buses supply all the 21 safety-relate t i.is .
Norinal source of power is the main 22 generator, an. -
. nit's auxiliary transformer. The unit 23 ,
auxiliary tra er is shown here supplying power to the 24 auxiliary bus nence, to the 6.9 KV safety buses located!
26 here (indica- 'ne A-SA, one B-SB.
11
I
,k s ~
) E.s.. L .. '.: . .;.
~ Attachment B OFECN....fNSC 11PT PROCEZDINGS BI? ORE NUCLEAR REGULATORY COMMISSION U. S. GEOLOGICAL SURVEY DKT/ CASE NO.
TITLE MEETING TO DISCUSS THE NRC CHARLESTON EARTH OUnxE ru 2HE COu2 ext Or E2S ERN SEzSurC::r PLACE RESTOn,~v1RGraza DATE m vE= ER ao, 1982 PAGES 1 - 22o J
p-,- ,
r;- -
ll \
..__; . . _ . . ; -. d ACEREN i-d?'CbIdd (202) 628-9300 440 FIRST STREET. N.W.
WASHINGTON D.C. 20001
.m Y 3o v
^
1 3R. JACKSONa I think I'll give it back to 2 Jin, because you pointed the question back to him. But
.. 3 let me maka an attempt first.
4 In teras of introduction before your question, 5 ve have issued, for those here who are not aware, upon 6 ceceipt of the USGS latter we wrote what is called or 7 termed a Commission paper, and this is a paper which 8 advises the Commissioners themselves of significant 9 actions that have taken place. This var issued under 10 the Executive Director of Operations' signature late in 11 the afternoon or early the following morning of receipt 12 of the letter. .
13 A Commission meeting was then held that
'. 14 afternoon, which was Fridar afternoon -- I can't recall -
15 the date right now - a week ago last Friday. There are 16 copies of this document out on the table if they're not 17 already all gone. And attached to that also is a copy 18 of the USGS letter and our proposed plan of let's say an 19 outline for NRC approach to resolving this issue.
20 Now, I think that one of the problems we have 21 is that Charleston earthquake occurring at Charleston is 22 more or less an artifact of the licensing process. I
~
23- don't think it has ever been and I've never heard any 24 scientist consider the fact that we restrict Charleston 25 to Charleston being a scien tific truth. It is really
%d 1
AA.DERSoN REPolmMG CoMPANif. INC.
M0 MRST ST., N.W., WA34tNGToN. D.C. 31001 (aHI GM
' Attachment C
- 5' DISTRIBUTION
,; Dircks Roe Rehm
. Denton Case November 19, 1982 Vollmer p"Cunningham Atir:ogue DeYoung Davis '
Kerr, SP E00 R/F FOR: The Commissioners FROM: Executive Director for Operations
SUBJECT:
CLARIFICATION OF U. S. GE0 LOGICAL SURVE7 POSITION RELATING TO SEISMIC DESIGN EARTHQUAKES IN THE EASTERN SEAB0ARD OF THE UNITED STATES PURPOSE: To provide the Commissioners with infcrmation relating to the clarification of the U. S. Geological Survey Position with respect to the 1886 Charleston, S.C.
Earthquake reoccurrence DISCUSSION For the purpose of licensing of facilities in the Southeastern U. S., the NRC has taken a position, based primarily on the advice of the U.S. Geological Survey
('jSGS), that any reoccurrence of the 1886 Charleston, S.C.
earthquake (Modified Mercalli Intensity (MMI) X, estimated liagnitude about 7) would be confinei to the Charleston area. That is, the Charleston earthquake is assumed to.be associated with a geologic structure in the Charleston i
area. Nuclear power plants in the region east of the Appalachian Mountains are, therefore, usually controlled in their seismic design, according to Appendix A to 10 CFR Part 100, by the maximum historical earthquake not associated with a geologic structure. This controlling earthquake is typically an MMI VII or VIII. Since 1974, the NRC has funded an extensive research project in the Charleston arca to gain further information on the causative mechanism of this event.
On January 28 and 29, 1982 the Extreme External Phenomenona
. Subcommittee of the ACRS convened a meeting of expert professionals in the geosciences to obtain an overview of the state of knowledge and future NRC research needs in this area. During that meeting, we were informed by the USGS that it had formed a working group to reassess the validity of its position on the Charleston earthquake.
Contact:
R. Vollmer, NRR
- 492-7207 , ,
- l. F v
This information was conveyed to the Cc=nissioners in a Commission Information Paper (SECY-82-53) on February 5, 1982. In that paper we indicated that any major modification of the former USGS position could have significant impact. on many Eastern U.S. nuclear plant sites. -
After many months of deliberation, the USGS has clarified its previous position relating to the 1886 Charleston, S.C.
earthquake. The attached letter, James F. Devine, USGS, to Robert E. Jackson, NRC, November 18, 1982 provides the position and indicates that:
"Because the geologic and tectonic features of the Charleston region are similar to those in other regions of the eastern seaboard, we conclude that although there is no recent or historical evidence that other regions have experienced strong earthquakes, the historical record is not, of itself, sufficient grounds for ruling out the occurrence in these other regions of strong seismic ground motions similar to those experienced near Charleston in 1886. Although the probability of strong ground motion due to an
- earthquake in any given year at a particular location in the eastern seaboard may be very low, deterministic and probabilistic evaluations of the seismic hazard should be made for individual sit e~s in the eastern seaboard to establish the seismic engineering parameters for critical facilities."
Based on our discussions with USGS senior personnel, this clarification is not intended to recommend that we categorically consider a Charleston-type event in the seismic design of all nuclear plants in the eastern seaboard of U.S. The USGS does believe, however, that an earthquake of this size should not be categorically ruled out at locations away from Charleston based solely on the statement in the December 30, 1980 USGS letter which states, " Consequently, earthquakes similar to the 1886 event should be considered as having the potential to occur in the vicinity of Charleston and seismic engineering parameters should be determined on that basis." Instead, this clarification provides guidance that indicates that such a conclusion should be. reached only after deterministic and probabilistic evaluations of the seismic hazard for individual sites have been made.
~
e
- O 1
~
- r. 5 Our evaluation of the significance of this clarification has only just begun. Currently, a two day review reeting between FIRC (ORES and OriRR) and the USGS is planned for t!ovembe. 3D, 1982 and Decerber 1, 1982 to discuss both the status of ceoscience knowledge in the Charlesten region and future research efforts. The first day will be an open public reeting (noticed in the Federal Register) which will allow for comments and questions frcm interested parties and r.enbers of the public.
He have also attached our preliminary views on a plan to address this clarified USGS position. This plan includes eler.ents which relate to both ongoing research and licensing efforts and possible requirenents for new efforts (split approxinately 75% and 25% respectively).
This plen will be nodified and corpleted after several meetings with the USGS take place in order that a core complete understanding of its clarified position can be obtained.
(;5pCY.TiilisI.Tld William J. Dircks Executive Director for Operations Attachtents:
As stated I. p + DE:GSBJN SBrocoum 1 /1){82; nTvea , g 7 gg ome,> [ ... .
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. .; i United States Department of the Interior
~
GEOLOGICAL SURVEY
. ,* RESTON, VA. 22092 In Reply Refer To:
Mail Stop 905
~
~..
l'O'l 18 1982 3..f._..[IDr.Rober'tE.'
Jackson -
~.
Chief, Geosciences Branch .
= Division of Engineering
- U.S. Nuclear Regulatory Commission
- Washington, D.C. 20555 .
Dear Bob:
The purpose of this letter is to clarify our position on the seismic potential of certain regions of the Eastern United States. In our letter of December 30, 1980, on the same subject we expressed the view that ". . . the likelihood of a Charlesto sized event in other parts of the Coastal Plain and Piedmont is very low." ,
As you are aware, after several years of intensive study in the Charleston region, no geologic structure or feature can be identified unequivocally as the source of the 1886 Charleston earthquake. However, as studies in the Charleston region and elsewhere.along the Atlantic margin have progressed, it has become evident that the
' ~ ~ ~
general geologic structure of the Charleston region can be found at other locales within the eastern seaboard (Appalachian Piedmont, Atlantic Coastal Plain, and Atlantic Continental Shelf). .
Because the geologic and tectonic features of the Charleston region are similar to those in other regions of the eastern seaboard, we conclude that although there is no recent or historical evidence that other regions have experienced strong earthquakes, the historical record is not, of itself, sufficient grounds'for ruling out the occurrence in these other regions of strong seismic ground motions similar to those experienced near Charleston in 1886. Although the probability of strong ground motion due to an earthquake in any given year at a particular location in th eastern seaboard may be very low, deterministic and probabilistic evaluations of th seismic hazard should be made for individual sites in the eastern seaboard to establish the seismic engineering parameters for critical facilities.
As stated in our letter of_ December 30, 1980, earthquakes similar to the 1886
. Charleston, South Carolina, event shotid be considered, as having the potential to
.. Occur in the vicinity of Charleston and seismic engineering parameters of critical facilities in that area should be determined on that basis.
', '. Sincerely yours,
~ -
g n_ $ [0s Q famesF.Devine Assistant Director for
. Engineering Geology e
t
, v .
Outline for Recommended Plan Eastern U. S. Eartnauakes Introduction .
Based' on our preliminary assessment of the U. S. Geological Survey's (USGS) clarification of position relating to a Charleston-type earthquake, we do not see a need for any immediate action for specific
' sites at this time. Instead, we foresee that this clarification can be
' ~
addressed predominantly through existing ongoing programs at NRC with the possibility of additional requirements for work by the Utilities.
' he T USGS clarification indicates that deterministic and probabilistic evaluations should be made. Generally, for nost existing sites, extensive deterministic studies have been undertaken and used in developing the existing seismic design basis. We therefore believe that this element'of the clarification continue to be addressed through our long range research plan. Specific modifications to that plan can be made in order to address specific tectonic structures. If necessary, a few specific applicants or licensees may be required to investigate tectonic structures which may not have been previously ' identified during the licensing proccdure.
As many of the current working deterministic hypotheses are not directly amenable to investigation in the short term, we believe that the clarification issue should be pursued in the short term principally through a probabilistic assessment of plants in the eastern seaboard.
This probabilistic program can be coupled to the current ongoing NRC efforts in this area already underway. We also believe that utility-sponsored studies should be undertaken, preferably as a
, consolidated group, to assess the seismic hazard in the eastern seaboard. ,
Further specifics on this program will be provided after more extensive discussions with the USGS.
PROBABILISTIC EVALUATION:
In our view, the USGS clarification represents not so much a new
, understanding but rather a more explicit recognition of existing uncertainties with respect to the causative structure and mechanism of the 1886 Charleston earthquake. Many hypotheses have been proposed as ta the locale in the eastern seaboard of future Charleston-size earthquakes. Some of these could be very restrictive in location while others wnuld allow this earthquake to reoccur over very large areas.
. Presently, none of these hypotheses are definitive and all contain a -
strong, element of speculation.
- Traditional deterministic approaches are not generally designed to deal
- with this situation. Probabilistic methods which allow for the consideration of many hypotheses, their associated credibilities, and
- the explicit-incorporation of uncertainty are much better equipped to provide rational frameworks for decision making. We believe that the 6
~
- up- r N' I probabilistic approach described below, which takes into account the uncertainties, should be used to determine differences (if any) between seismic hazard levels associated with seismic design values in the eastern seaboard (i.e. as affected by the USGS clarified position on the Charleston Earthquake) and seismic hazard levels associated with seismic design values elsewhere in the central and eastern U. S.
Probabilistic Plan :
- 1. Continue development of Lawrence Livermore National Laboratory (LLNL) study on seismic hazard (probability of exceedance) for nuclear power plants east of the Rocky Mtns. This study (Seismic Hazard Characterization of the Eastern United States) is presently underway.
- 2. Compare of LLNL study with existing probabilistic studies (for example USGS Open File Report 82-1033) and other ongoing NRC Research into probabilistic seismic estimation.
- 3. Sponsorship by the industry as 'a whole of a probabilistic estimation of hazard for all nuclear plants on the eastern seaboard, along with existing studies for individual plants.
- 4. Make comparisons between plants in the eastern seaboard and other parts of U.S. using the LLNL and other studies to determine significant differences (if any) in seismic hazard associated with seismic design.
- 5. Integration of above into Systematic Evaluation Program-type s g 3 rj
. valuation e for possible engineering reanalysis.
- W y.[S +
DETEPJilNIST.IC EVALUATION:
.u -
f35YTE Deterministic studies in response to the USGS clarification should EY continue to be oriented toward determining the causal mechanisms of the dd'0% earthquake under NRC's existing research program. These studies should l[1
.Thi involve systematic testing of the se'veral hypotheses of the causative structure of the Charleston earthquake and investigations in areas of
(?g {' high seismicity and designated areas of potential seismicity for
[]ft54/ additional evidence of the cause. The type of studies most likely to
' ,' ~
lead to a better understanding of the causes of seismicity in the eastern seaboard of the United States are neotectonic investigations (recent crustal motions and seismicity) coupled with examination of crustal structure:
. These deterministic studies are basically four types:
- 1. The continuation of seismological research through the operation of the existing micro-earthquake networks and the development of a strong motion data base.
e mm . l i i i
' ' '- - - - - - aiu ' ika, d' ,
d *
- 2. The determination of the geometry of structure and tectonics of the earth's crust at depths where earthquakes are occurring (5-20 km) in the eastern seaboard using such techniques as seismic reflection profiling.
- 3. The continuation of subsurface neotectonic investigations of ,
earthquake source areas to determine if uplift, subsidence or differential movement is occurring. Such studies may include among others:
A. Tectonic Geomorphology B. Geodetic Measurements
. C. Geologic Mapping D.. Remote Sensing e
b e
ly I b %-
Attachment D q
SERIAL: LAP-83-426 September 27, 1983 i
Mr. Harold R. Denton, Director office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PIANT ,
UNIT NOS. 1 AND 2 DOCKET NOS. 50-400 AND 50-401 SUPPLEMENE 1 TO NUREG-0737 - DETAILED CONTROL ROOM DESIGN REVIEW
Dear Mr. Denton:
On August 15-19, 1983, the NRC staff conducted a Detailed Control Room Design Review audit of Carolina Power & Light Company's (CP&L) Shearon Harris Nuclear Power Plant (SHFPP) Units 1 and 2 control rocas. During the audit, your staff requested the following additional information to completa their review on Section 5 of Supplement I to NUREG-0737:
- 1. A description of the system's functional analysis performed on the SENPP Unit 1 Main Control Board during its redesign (Attachment 1),
- 2. A description of the method and a general target date for completion of the Task Analysis of the plant specific Emergency Operating Procedures, (Attachment 2) and;
- 3. A description of the Development, Verification and Validation process f or emergency operating procedures (EOPs) and a general target date for completion (Attachment 3).
We trust this submittal provides the information your staff needs.
Should you require clarification of the information provided, please contact g staff.
Yours verge g y, ORIGINAL SIG M. A. M-cuFFl3 , ,
M. A. McDuffie Senior Vice President Nuclear Generation MSG /tda (7896 MSG)
At tachsents ec: Mr. B. C. Buckley (NRC) Mr. Wells Eddleasa Mr. G. F. Maxwell (NRC-SBNPP) Dr. Phyllis Iotchin Mr. J. P. O'Reilly (NRC-RII) Mr. John D. Runkle Mr. Travis Payne (KUDZU) Dr. Richard D. Wilson Mr. Daniel F. Raad (CHANGE /ELP) Mr. G. O. Bright (ASLB)
-Mr. R. P. Gruber (NCUC) Dr. J. H. Carpenter (ASLB)
Chapel Hill Public Library Mr. J. L. Kalley (ASLB)
Wake County Public Library
. . . . 3
.-- n . -- a : .. .. . -
1
ATTACHMENT 1 CAROLINA POWER AND LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT (SHNPP) - UNIT NO.1 SYSTEMS FUNCTIONAL ANALYSIS PERFORMED ON THE MAIN CONTROL BOARD DURING REDESIGN Introduction Carolina Power & Light Company's (CP&L) Shearon Harris Nuclear Power Plant (SHNPP) Unit 1 Main Control Board (MCB) was redesigned during January through March of 1981, based upon a Human Factors Review performed by Essex Corporation of Alexandria, Virginia, using Draf t NUREG-CR/1580 criteria. The ikiman Factors (HF) Review identified 134 Human Engineering Discrepancies (HEDs); 49 of the HEDs were considered significant. These 49 HEDs concerned grouping and sequencing of controls, displays, light boxes, and annunciators. All 49 of these discrepancies were resolved with the redesign of the MCB.
The redesign and functional analysis were performed by CP&L engineering, operations, training, industrial engineering, and Essex Corporation personnel. Assistance f roa both the Architect-Engineer (Ebasco) and the Nuclear. Steam System supplier (Westinghouse) was provided as needed.
Functional Analysis Carolina Power & Light Company's design philosophy for the MCBs dictates the placing of controls on the bench section, indicators with light boxes (such as Status Light Boxes (SLBs) and Monitor Light Boxes (MLBs)) on the vertical section, and annunciators on the top tilted section of the MCB.
Experience has proven that this philosophy is both practical and effective.
.This design philosophy allows the Control Operator to perform administrative duties at his desk while still maintaining the ability to scan the MCB to determine plant status. It also allows the Shif t Foreman and other_ plant personnel to determine plant status f rom a distance, outside of the primary operating area, without interfering with the Control Operator. During the redesign process, CP&L established conventions for the MCB redesign effort.
The conventions chosen are as follows:
- 1. . Bottom to top layout based upon the physical layout (f rom Piping and Instrumentation Diagrams (P& ids)).
- 2. Layouts that must be horizontal will be lef t to right with appropriate demarcation lines or arrows to clearly indicate the system flow.
- 3. Series flow will be indicated by placing controls (and displays if possible) directly above each other from bottom to top, OR arrows or lines and arrows will be utilized to denote system flow. ~
- 4. Parallel flow will be indicated by placing controls (and displays if possible) side by side with "A" or "1" (if applicable) on the lef t and "B" or "2" (if applicable)'on the right.
' a 4
- 5. Common suction or discharge (header) will be denoted with a solid bar, as necessary, to clarify the arrangement.
- 6. Demarcation lines will be used to separate control display groups.
. 7. Demarcation lines or lines and arrows will be used where system flow is not obvious by components arrangement i.e. , memic or partial memic.
- 8. Summary labels and brackets will be utilized to clarify the arrangement.
- 9. Indicators will be placed in the order they physically appear in the system. If this is not practicable, indicators for level, pressure, flow, and temperature will be placed in this order (pref erably f rom lef t
& to right (first choice) or bottom to top (second choice)).
- 10. Recorders on the MCB will be placed on the vertical section at a level (height) where they can be easily read and maintained.
During the HF Review process, we determined which unnecessary components could be removed from the MCB. This review and subsequent determination was based upon design philosophy, operational need, operating experience, staffing, operating philosophy, and our decision to provide an advanced computer system. Our review culminated in the removal of approximately 200 controls and displays and approximately 250 annunciators f rom the MCB. In addition, coacurrence f or removal of components f rom the MCB was sought and obtained from both Ebasco and Westinghouse.
A revicw of MCB systems locations of the current design (i.e. prior to January 1981), revealed extensive thought and logic had been applied to systems locations. Generally, the systems location on the MCB remained unchanged during the redesign. A detailed review of each system, the method of which will be described later, revealed some components not properly located within their respective system and indicated arrangement of components within each system could be improved to f acilitate operation. Different methods of arranging components such as f requency of use, like components ,
(pumps, valves) grouping, modes of operations, and arrangement of physical layout (from P& ids) were evaluated. The physical layout method was chosen because: 1) it was more practicable f rom an operations standpoint, 2) this is how our operators learn plant systems, and; 3) systems have many different operating modes where sequences of operation vary.
The redesign began by constructing a quarter scale, single plane mock-up on cardboard, utilizing the same dimensions and MCB shape as a standard "D" sized engineering front panel view drawing of the MCB. The cardboard mock-up with the MCB panel outline was then covered with clear plastic. A set of the current design (i.e. prior to January 1981) f ront panel drawings was then utilized by cutting out each component, pasting it to a piece of the same type cardboard and applying " stick-um" which would allow the
_ component to be removed for rearrangement. Each component was then attached to the mock-up to reflect the current design. String was stretched across the mock-up to indicate reimary and secondary viewing heights.
4 The redesign team was provided an oral description of each system, prior to the system restrangement, by operations personnel which included system function (and how that function interrelated with the overall operation of the plant), the ph7 aical layout, and the instrumentation included in the design to accomplish the function. Each control, display, and other indications l'o cated on the MCB was then marked on the system P&ID. The controls and displays were then arranged on the mock-up, utilizing the (P&ID) and the conventions previously outlined. Demarcation lines were then added to the mock-up and the necessary labeling was determined. Exact labeling was cross-referenced to the mock-up with a numbering system because labeling would not fit on the mock-up.
Annunciators were rearranged into groups according to their applicable system by cutting and pasting drawings. Where possible, the annunciators were arranged bottom to top in relation to their respective components or sensor input in the physical layout. In cases where this annunciator arrangement w:s not practicable, the annunciators were logically grouped by function.
Next, all Status Light Boxes, Monitor Light Boxes, Trip Status Light Boxes, the Bypass Light Box, and the Engineering Safety Features (ESF) Light Boxes were reviewed f or logical groupings by either system or f unction. These light boxes were then rearranged into logical groupings by cutting and pasting the drawings.
As panel sections of the MCB redesign were completed and translated to engineering drawings, Ebasco and Westinghouse (which included the appropriate disciplines), the review team and other CP&L personnel held review meetings where the redesign was evaluated. Comments were incorporated and the redesign was implemented.
Adequacy of instrumentation was continually evaluated by the team throughout the redesign process. In several cases, additional instrumentation was needed to accomplish systems functions. The additional instrumentation was added to the drawings during the redesign process.
As a result of our redesign effort and continual review process, CP&L believes the SENPP Unit 1 main control board is a well designed, operationally functional, and Human Factored Control Board.
l l
1
-i ATTACHMENT 2 3 CAROLINA POWER AND LIGHT COMPANY q SHEARON HARRIS NUCLEAR POWER PLANT (SHNPP) UNIT NOS.1 AND 2 ,
TASK ANALYSIS OF THE UPGRADED 1 EMERGENCY OPERATING PROCEDURES (EOPs) 4 Introduction Carolina Power & Light Company's (CP&L) April 15, 1983 response to Supplement 1 of NUREG-0737 for the Shearon Harris Nuclear Power Plant (SHNPP)
Unit Nos. I and 2 stated our Emergency Operating Procedures (EOP's) Procedures Generation Package (PGP) would be submitted to the NRC nine mouths prior to f uel load. The PGP will provide plant-specific technical guidelines, a ;J Writer's Guide, and a description of our verification and validation 5 program. Operator training will be accomplished prior to SHNPP Unit ifuel ,
load which is currently targeted for June 1985. Carolina Power & Light f Company anticipates completing the E0P Task Analysis concurrent to submittal ;
of the PGP. 1 Task Analysis Met. hod 9
A Task Analysis has been performed on the High-Pressure (HP) Basic j version of the Westinghouse Owners' Group (WOG) Emergency Response Guidelines i (ERGS) by a working group under the purview of the WOG Procedures ;
Subcommittee. The primary outputs of this generic Task Analysis are tables j listing all Instruments and Controls utilized in performing the ERGS. The i detail of the generic Task Analysis is consistent with the detail provided in 2l the generic ERGS.
[ The generic Task Analysis utilized a top-down approach that i identifies the guidelines (i.e., event sequences), plant systems utilized in j responding to event sequences, operator f unctions and operator tasks performed 3 in responding to event sequences, and detailed elements that comprise the 3
operator tasks. Figure 1 illustrates this approach.
=
As a minimum, CP&L intends to identify the deviations f rom the j generic ERGS f or the SHNPP-Unit 1 EOPs, task analyze those differences and i generate plant specific lists of Instruments and Controls necessary to perform the E0Ps in the SHNPP-Unit I control room. Figure 2 describes this ;
approach. These Instruments and Controla listings will then be compared to control room instruments and controls to identify missing components or needed '
a components not included in the design.
Additionally, CP&L will review the generic Task Analysis along with the ERGS deviations analysis, thus insuring review of each step of the ;
SHNPP-Unit 1 E0Ps. Discrepancies identified during the review and analysis j will be judged applicable to the E0Ps, Control Room or both and will be 1 resolved and corrected by us. We believe that no major discrepancies will be identified because of our extensive functional analysis performed during the j SHNPP-Unit 1 MCB redesign process and because of the task analysis perf ormed j on the event-based procedures during the Control Room Design Review.
4
]i
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' 5 ATTACHMENT 3 $
CAROLINA POWER AND LIGHT COMPANY d
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SHEARON HARRIS NUCLEAR POWER PLANT (SENPP) UNIT NOS.1 AND 2 DEVELOPMENT, VERIFICATION AND 5 VALIDATION OF EMERGENCY OPERATING i PROCEDURES g A
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Development: $
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Carolina Power and Light Company's (CP&L) Shearon Harris Nuclear h Power Plant (SHNPP) Emergency Operating Procedures (EOPs) are currently being ;
developed utilizing the Westinghouse Owners' Group (WOG) Emergency Response i
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Guidelines (ERGS) Eigh Pressure (HP) version, Revision 1, as Technical i Guidelines. The E0Ps will follow the ERGS as closely as possible and any J deviations will be documented, exAained, and/or justified. Documentation i will also be generated f or the basis of the plant-specific calculations called j for in the generic guidelines. d ih The basic version of the WOG ERGS have undergone one week of ll simulator verification and validation testing. A program for a week of simulator verification and validation testing of Revision 1 of the ERGS is now
+ ..- being assembled by Westinghouse and will be performed during the week of a October 31 - November 4,1983.
Carolina Power & Light Company has been deeply involved in the 3 ; ., development of the ERGS since their inception through participation in both j
'Q. h.e the full Owners' Group and the Procedures Subcommittee, i fp j en ,
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%ct Verification and Validation Methods ih: E j V.M Tabletop Evaluations
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[N.) Tabletop evaluations will be performed on all E0Ps and will consist q of a talk-through of the procedures by qualified operatiois personnel and
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- D4 members of the team responsible f or developing the EOPs. Scenarios will not j
- Jg ' be utilized during the tabletop evaluation. The evaluation will be documented g Il Th...! as to time and date of performance. personnel involved, procedures utilized, i Tfj problems or suggestad improvenants noted and, later, the solution of those j
^ problems or suggested improvements. The evaluation criteria utilized during I j
[' the tabletop exercises are:
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- 1. EOFs are technically correct. j si
- 2. E0Ps are understandable as written.
- 3. EOPs are written in conformance with the Writer's Guide.
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- 4. Level of detail in the EOPs is consistent with the qualifications, i training, and experience of the operating staff. ?
Tabletop evaluations will be held in the spring of 1984. j i
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f Control Room Walk-Through he Control Room Walk-Through will consist of walking and talking through each EOP in the Control Room with a full operations staff coglement. Scenarios will not be utilized in the walk-throughs. A member of tha team responsible for developing the E0Ps will lead the walk-through. The walk-through will be documented as to time and date of performance, personnel involved, procedures utilized, problems or suggested improvements noted, and later, the resolution of those problems or suggested improvements. In addition to the criteria utilized in the tabletop evaluations, three additional criterion will be utilized in the walk-through. These criterion ares r
- 1. Control room staff size is adequate to carry out the actions in the E0Ps.
- 2. Instruments and controls necessary to carry out the EOPs actions are available.
- 3. Operators can carry out the E0Ps actions without physical interference.
he walk-throughs cannot be carried out at SHNPP until the Control Room is functional (where functional is defined as):
- 1. Structurally completed (ceiling, lighting, and HVAC installed; panels correctly and permanently placed; etc.)
'2. All instruments and controls installed but not necessarily operable.
- 3. Manned with a full operations shift complement.
We expect the SHNPP Unit 1 Control Room to be f unctional in late 1984 or early 1985. The walk-throughs are being planned for this time frame.
_ Simulator Evaluations The simulator evaluation will consist of utilizing the SHNPP simulator to dynamically test the E0Ps with accident scenarios. Testing of two eight-hour shif ts, where preselected scenarios will be imposed on a f ull e complement operating crew using the EOPs, will be performed at the SHNPP simulator. This testing is judged as adequate because:
- 1. SHNPP EOPs are very similar to H. B. Robinson Unit 2 (HBR) EC?s and the results of the HBR testing (which will be completed first) will be input to the SHNPP E0P's development.
- 2. The HBR E0Ps have undergone 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> (as of September 1983) of dynamic testing at the SENPP simulator.
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- 3. Many of the same personnel involved in the development and testing of the HBR E0Ps will be involved in the writing and testing of the SENPP E0Ps.
l l The simulator evaluations will be documented as to the time and date of performance, personnel involved, procedures utilized, scenarios selected, expected path through the EOPs, actual path through the E0Ps with deviations explained, operators' debriefing critiques, observers' critiques, problems or suggested improvements noted, and later, resolution of those problems or suggested improvements.
The simulator exercises will be oriente i toward the practical performance of the E0Ps and is expected to be performed in the mid- to late-1984 time f rame.
(7896MSGeda)
SYSTEM REVIEW AND TASK ANALYSIS DEVELOPMENT APPROACH '
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- o OUIDELINES j
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SYSTEMS l l a
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OPERATOR FUNCTIONS l m
G OPERATOR TASKS R
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- g if OPERATOR TASK REQUIREMENTS
- Function
- Task / Subtask
- Objective
- Decision (Criteria) Requirements
- Knowledge Requirements
- Ins trumenta tion (Criteria) Requirements
- Action (Criteria) Requirements
- Conlrof Capability (Criteria) Requirements
- Consequences of Error / Omission L_
7 SYSTEM REVIEW AND TASK ANAL.YSIS '
DEVELOPMENT APPROACH l l n
5 OUlDELINES I 8 8 DE T S Em h3 SYSTEMS C n 1 3,
OPERATOR FUNCTIONS a
y OPERATOR TASKS R
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- g 1 f OPERATOR TASK REQUIREMENTS
- Funcilon
- Task / Sublask
- Objective
-Decision (Criteria) Requirements
-Knowledge Requirements
- Ins trumonta tion (Calloria) Requirements
- Action (Criteria) Requirements
- Controf Capaisility (Criteria) Regulrements
-Consequences of Error / Omission 1 \ -
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