ML20073J659
| ML20073J659 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 04/15/1983 |
| From: | Wetterhahn M CONNER & WETTERHAHN, GULF STATES UTILITIES CO. |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8304190363 | |
| Download: ML20073J659 (87) | |
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UNITED STATES OF.AMERICAKO!'
NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and *d}c M Ng caru
-- - , .n t U-In the Matter of ) bbiyj CH
)
Gulf States Utilities Company, ) Docket Nos. 50-458 et al. ) 50-459
)
(River Bend Station, Units 1 )
and 2) )
APPLICANTS' ANSWER TO THE CONTENTIONS FILED BY PETITIONERS FOR LEAVE TO INTERVENE, STATE OF LOUISIANA THROUGH THE OFFICE OF THE ATTORNEY GENERAL, LOUISIANA CONSUMER'S LEAGUE, LOUISIANANS FOR SAFE ENERGY, INC. AND GRETCHEN REINEKE ROTHCHILD Introduccicn By a Memorandum and Order. da ud July 30, 1982, the presiding Atomic Safety and Licers:.ng Board (" Licensing Board" or " Board") held that petiticners for leave to intervene, namely, the State of Louisiana,1/ Louisiana Consumers League, Inc., Louisianans for Safe Energy, Inc.
and Gretchen Reineke Rothchild, had satisfied the interest 7crtion of the Nuclear Regulatory Commission's regulations relating to intervention. Accordingly, the Licensing Board
-1/ The State of Louisiana through the Office of the Attorney General had petitioned to participate under the provisions of both 10 C.F.R. SS2.714 and 2.715. By Memorandum and Order Ruling on Petitions to Intervene (February 10, 1982), the Board had admitted the State under the provisions of 10 C.F.R. S2.715(c).
8304190363 830415 O PDR ADOCK 05000458 0 PDR 1
, a-directed' petitioners to submit. proposed contentions by September 7, 1982. By an August 20, 1982 Order, the Board extended the - time for filing contentions . to December 13, 1982.- In a subsequent Order of December 21, 1982, the Board again extended the time for filing contentions to March 15, 1983.
On December 15, 1982, a pleading; entitled " Contentions By Joint Intervenors . [ Louisiana} Consumers' League, Inc.[,]
Louisianans for Safe Energy [,] Gretchen Reineke Rothchild" was submitted. I On March 15, 1983, the " Supplemental Petition of.the State of Louisiana" was filed, superseding an earlier submittal of contentions dated December 15, 1982.
As discussed below, Applicants, Gulf States Utilities, et al. ("GSU" or " Applicants"), submit that all of the 1 contentions filed are either inadequate under the Com-mission's rules or are premature. Thus, the Licensing Board should deny the pending petitions for leave to intervene.-
I. General Discussion of the Rules Governing Admissibility of Contentions It is well established that a contentien which merely alleges generally ths; the application f e r- an cperating
~2/ Petitioners have inaccurately referred to themselves as
" Joint Intervenors." Applicants note that the Board has not determined that the requirements of 10 C.F.R. 52.714 for filing contentions have been satisfied, and petitioners have therefore not achieved status as
.intervenors. Accordingly, Applicants will refer to them as " petitioners."
license is inadequate fails to meet the requirement of specificity contained in 10 C.F.R. S2.714 (b) . Licensing boards have refused to admit such contentions. This is particularly true where the Final Safety Analysis Report, Environmental Report-Operating License Stage, or other portion of the application discusses in detail the manner in which the Commission's requirements are being met.
For example, in Susquehanna 3/ the Licensing Board denied a contention which alleged that portions of the applicant's environmental' report understated certain effects of an accident. The Board held that the proposed contention failed to meet the Commission's- test for specificity and stated:
In order to evaluate whether a con-tention presents an issue in controver-sy, the regulations specify that their bases should'be set forth with reason-able specificity. Here.we are left to wander aimlessly in our speculation on the details of the allegations--a practice obviously unfair to proper procedure, to the parties and the Board.
The contention will not be admitted. 4,/
Likewise, in Offshore Power 1/ the Licensing Board 3/ Pennsylvania Power & Light Company (Susquehanna Steam Electric Station, Units 1 and 2), Docket Nos. 50-397 and 50-388, " Memorandum and Order on Pending Motions and Requests" (July 7, 1981).
4/ Id., slip op, at 4.
5_/ Offshore Power Systems (Manufacturing License for Floating Nuclear Power Plants), LBP-77-48, 6 NRC 249 (1977).
.. .. _ _ __ .. .~ . . _ ._ .
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4 denied admission.of'a contention which merely asserted that -
l the-Staff-'in the.. Final Environmental Statement inadequatcly considered.and improperly dismissed various alternatives to i-
- the . licensing of the proposed stations. The Board found this contention- Lto -be inadmissible because it was conclusionary and lacking in the necessary specificity and
, factual- . bases.6/ Further, the Board stated. that
"[cl ontentions ' which are -barren and- unfocused are of no assistance to us in the resolution of the issues to be decided." 7/ In the Seabrook proceeding, the Licensing j Board rejected a contention- regarding a ' general . deficiency-1 in the qualification of equipment as - too . broad to be lit-9 igated in an operating license proceeding because the intervenor had not specifically designatedfthe equipment or
-categories of equipment.to which the contention related.8/
-Similarly, in' the G.E. Morris 9/ proceeding, the
- Licensing Board rejected a contention that the applicant had-not taken into' account the close proximity of two 5/ II. . at 25C.
j 7/ Id. at 251.
! 8_/ Public Service Ccmpany of New Hampshire (Seabrook Station, Units 1 and 2), Docket Nos. 50-443 OL and 50-444 OL, Memorandum and Order" - (September 13, 1982)
(slip op. at 15-16).
9/ General Electric Comoany (GE Horris Operation Spent Fuel Storage Facility) , Docket Nc. 70-1303 OLA (Spent Fuel Pool) , " Order Ruling on Contentions of the Party" l (June 4, 1980).
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. a facilities, noting that "[n]o litigable issue is presented" by the mere recitation of this information in the peti-tion.10/ See also Commonwealth Edison Company (Dresden Nuclear Power Station, Unit No. 1) , Docket No. 50-10-OLA,
" Memorandum and Order" (July 12, 1982) (slip op. at 9);
Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2), Docket Nos. STN 50-556CP and 50-557CP, " Memorandum and Order" (January 11, 1982) (slip op, at 2-4); Long Island Lighting Company (Shoreham Nuclear Power Station, Unit 1),
LEP-31-18, 14 NRC 71, 75 (1981).
Thus, those contentions which merely allege inad-equacies or violations in the broadest of terms do not meet the standards of Section 2.714(b) for specificity and bases.
The proposed contentions in this proceeding should therefore-be denied.
Further, several of petitioners' proposed contentions are premature and must be denied or deferred pursuant to the Appeal Board's recent ruling in Catawba.11/ In that proceeding, the Appeal Board considered the circumstances under which the Licensing Board may allou thi ccnditional admission of a contention subject to later specification.
10/ Id., slip op. at 20.
M/ Duke Pcwer Company (Catawba Nuclear Station, Units 1 and 2), ALAB-687, 16 NRC _ (August 19, 1982).
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The Board held that such " conditional" contentions were not t'o be accepted and stated:
Nothing in the terms of Section 2.714 (b) explicitly vests a licensing board with the power to admit an unacceptably vague or imprecise contention . conditionally, subject to later revision upon receipt of additional information. Rather, as we read it, the Section conveys the clear- message that, in order to - be admitted, the contention must meet the
" requirements of this [Section] " ; i.e.,
it must set forth its bases "with reasonable specificity." Moreover, the administrative history of the Section precludes any suggestion that the Commission intended an implicit excep-tion to the - specificity requirements in circumstances where, because of a lack of available information, it is not possible for the petitioner to meet those requirements at the time its contentions are due. 12/
In applying this ruling to the contentions before it, the Licensing Board in Catawba held that contentions relat-ing to emergency planning were premature because the plans were not available.13/ The Licensing Board rejected those contentions which lacked the requisite specificity, but acknowledged that it would also be within the Board's discretion to defer such contentions. Under th'e rulings in Catawba, however, it is not within the discretion of the 12/ Id., slip op at 9.
M/ Duke Power Company, supra, Docket Mos. 50-413 and 50-414, " Memorandum and Order (Reflecting Decisions Made Following Second Prehearing Conference)"
(December 1, 1982) (slip op, at 5).
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Board to admit -such premature contentions. Petitioners' proposed contentions which are subject to this defect.should therefore be denied.
Utilizing these general principles, each of the proffered. contentions is discussed below. Because a number of ' contentions of the joint petitioners 'and the State overlap, they are discussed together. Where there . is ' a general allegation, the subject matter of which is discussed in detail in the application, this - f act is noted, not to refute the merits of the contention, but to indicate . the lack of specificity and the absence of any litigable issue.
II. Scecific Discussion of Proffered Contentions-Joint Petitioners' Contention 1 Financial and Technical Qualifications
.-Joint petitioners attempt to make a showing of "special circumstances" . pursuant to - 10 C . F . R. 52.758 (b) in order to overcome the Commission regulation which prohibits consid-eration of an applicant's financial qualifications. in licensing proceedings.14/ Petitioners have failed to cite any such special circumstances which would just.ify a waiver cf this regulatica in this proceeding anf, in fact, the arguments they offer were rejected by the Commission in its rulemaking proceeding for the adopted regulation. Nor have
_1_4_/ 10 C.F.R. S50.40 (b) .
petitioners complied with the . mandatory procedures for seeking a waiver of regulation.
As: noted, the sole ground for a challenge to any Commission rule or' regulation in an adjudicatory proceeding is-that "special-circumstances" with respect to the subject i matter of the particular proceeding are such that applica-tion of the rule or regulation would not serve the purpose for which it was adopted. Such a request must be accom-panied by an affidavit that identifies specific aspecta of the subject matter of the proceeding as to which application of the rule or regulation would not serve the purposes for which it was adopted. It must set forth with particularity the special circumstances alleged- to justify the waiver or--
exception requested. Unless a petitioner makes a crima facie showing that the challenged rule or regulation would not serve the purpose for which it was adopted, no evidence on the issue can be received and the presiding officer may not consider the matter further.15/
I No .af fidavit as required has-been submitted by peti-tioners. As "special circumstances," they derely allege that GSU's financial status has changed substantially for i
l the worse since the construction permit was granted,16/'
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15/ 10 C.F.R. 52.75 S (c) , (d).
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16/ Cententions By Joint Intervenors Consumers' League, Inc.[,] Louisianans for Safe Energy [,) Gretchen Reineke Rothchild (" Contentions by Joint Intervenors") at 2
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that there have been " numerous' inspection reports document-1 l ing where construction activities were not conducted in full -
compliance with regulatory requirements under circumstances f indicating that. cost-cutting measures were involved."17/
i j' The Commission has addressed and rejected virtually the same arguments in the rulemaking proceeding;18/ thus they j cannot constitute-the showing required by 10 C.F.R. S2.758.
. Underlying . their proposed . contention is petitioners' erroneous assumption that the general ability of a' utility to finance construction of new generating facilities' is inevitably linked to considerarions of the public health and I safety under the. Atomic Energy Act of 1954, a_s. amended, 42 U.S.C. 52011 et seg. .and implementing regulations. 'The Commission expressly rejected this view in eliminating the j financial qualifications rule "because of the lack of - any i ~
- demonstrable-link between public health and safety concerns i
, and a utility's ability to make -the requisite financial showing. 19/ -
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1 (December 13, 1982).
.17/ Id. r 18/ See ' generally 47 Faf. Reg. 13750 (March 31, 1932) i (codified in 10 C.F.R. S50.40(b)).
!.I' -19/ Id. at 137.51.
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In -a recent decision, the Licensing Board in Byron 20/ considered and rejected the admission of pre-cisely the issues sought to be raised here. The Board disposed of -the allegation that the ~ applicant- did 'not possess or have reasonable assurance of obtaining funds necessary to complete construction, operation and decommis-sioning of the facility safely- by noting that since
- "[nlothing was provided to show that a lacking of financial qualifications ipso facto translated itself into'a threat to public health and safety," a general allegation of insuffi-cient funds fails to establish the requisite special circum-stances.21/
In' order to qualify for a waiver of the regulation, the Licensing Board in Byron held that-a petitioner must affir-matively' demonstrate:
. . . an actual causal connection
.between a utility's financial qualifica-tions and it presenting a threat to public health- and safety . . . .
Without. the establishment of a demon-strable link between- financial quali-fications and it constituting a threat to public health and safety, there would be no showing that the' subject regu-lations do not sarve the purpose for which they were adopted. 22/
-20/-Commonwealth Edison Company (Byron Station, Units 1.and 2), Docket Nos. STN 50-454 OL and STN 50-455 OL,
" Memorandum and Order" (August 2, 1982).
21/ Id., slip op. at 4.
22/ Id., slip cp. at 5.
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Similarly, in this. proceeding, petitioners' bare assertions-
.that GSU's' financial status has changed fail'to provide.the
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requisite showing and therefore should be rejected.
Joint petitioners allege -that " numerous inspection reports" support their = position, -but cite only three.23/
Here, joint petitioners have failed to make any connection whatsoever between the - three cited inspection ' ' reports and GSU's financial condition. One report relates to storage procedures, the second to the reportability of certain matters and the third to' grades of reinforcing- steel. In the Byron proceeding, supra, the' Board' expressly addressed' the issue of similar, but' more numerous noncompliances and concluded that there was no ' showing that the . deficiencies were in . any -way related to applicant's financial condi-tion.24/ Similarly, the Commission in the rulemaking proceeding noted that violations detected by the NRC Office of Inspection & Enforcement cannot, as a general ~ matter, be
shown to arise from a licensee's alleged lack of financial qualifications."25/ No showing has been made that these are other than minor, isolated noncompliances w.-hich are nct unexpected on a project of this magnitude.
23/ Contentions by Joint Intervenors at 2.
24/ Byron, supra, slip op at 7.
M/ 47 Fed. Reg. at 13751.
a ..
The -final item- cited by petitioners -relates 'to a request regarding .one-half inch tubing.26/ The request to allow alternate method:' to meet 'Conmission requirements. was made to the NRC publicly ' and. openly. NRC approval was therefore sought in the normal' course'of review.- The NRC memorandum to which the joint petitioners refer reflects a meeting it held with GSU on October-22, 1982. The memoran-dum states , . inter alia, that "the NRC Staff expressed a willingness to assist GSU- in . realizing its cost-effectiveness goals by responding in a reasonably short time-frame ~ to such proposals from GSU."27/ Inasmuch as GSU fully supported its proposed action.and provided the NRC with a clear and supportable basis -for accepting 'its pro-posal,28/ there is no basis or . support for petitioners' allegation ' that _ any adverse-effect upon the public ' health and-safety could result.
Having " failed .to establish' a- nexus"2'9/-
between Applicants' financial condition and any of the four items 26_/6 Contentions by Joint Intervenors_at 2.
-27/ Memorandum for A. Schwencer from J. Stefano at 2 (November 3, 1982). Rather than addressing the costs associated with the change, the quoted passage relates to obtaining an- early determination as to- the 2
acceptability of the change so the Applicants can know which' course to pursue.
28/ Id.
i 29/ Byron, supra, slip op at 7.
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l discussed above, joint petitioners have failed to make'.the requisite showing that tite rule on financial qualifications
.should be waived. This contention should therefore be
-denied.
Joint Petitioners' Contention 2=
Environmental Qualification
~ This contention alleges that "[t]he Applicants have not demonstrated that they will .be in compliance"30/ -
with interim NRC Staff positions related to environmental quali-fication of certain equipment for the River Bend Station.
This proposed contention raises no-litigable issue inasmuch as it merely engages in prohibited speculation regarding Applicants' compliance with.NRC~ requirements.
Section 3.11 of the Zinal Safety Analysis Report addresses with specificity: the d=: tails - of the - Applicants '
program to assure compliance wi environmental qualifica-tions requirements. In particular, page - -3.11-3 addresses compliance -with ITUREG-0588, .the document to which joint-petitioners refer. 1,/- Section 3.11 also refers to an additional document entitled " Environmental Qualification Occument,"32/ whica is a constituent p .rt cf the River Bend Station Final Safety Analysis Report. This document H/ Contentions by Joint Intervenors at 3.
31/ Id.
3J/ Final Safety Analysis Report at 3.11-1.
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addresses specific plant equipment and the program for its environmental l qualification. ,
Moreover, on January 21,.1983, the Nuclear Regulatory-Commission published in the Federal Register a final rule,
-entitled " Environmental-Qualification of Electric Equipment Important to . Safety for Nuclear Power Plants."33/ This amendment codifies in 10 C.F.R. 550.40 the environmental qualification methods and criteria- that meet the Com-mission's requirements in this area for nuclear power plants. Joint petitioners have pointed to no deficiency in the Final Safety Analysis . Report, including . the Environ-mental Qualification Document, nor any inability on the part of .the River Bend Station to conform to the . requirements ' of
'the Commission's rules regarding environmental qualifica -
tion. Thus, this speculative contention is -barren and unfocused and should be %nied.
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[ Joint Petitioners' Contention 3 Induced Seismic-Activity Joint petitioners allege that Applicants have failed to consider adequately the effects 'of seismic activity on the River Bend Station resulting frca e::ploratory and/or produc-( tion natural gas wells within the radius of the exclusion 1
zone and from subsidence due to withdrawal of water, oil 3_3/ 48 Fed. Reg. 2729 (January 21, 1983) (codified at 10 C.F.R. Part 50).
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and/or gas. This proposed contention is merely a generalized assertion, without any basis whatsoever, that Applicants have failed to consider induced seismic activity.
This contention ignores ' completely the information and analyses contained in the Final Safety -Analysis Report regarding the lack of induced seismic ~ activity. In particu-lar,. Sections 2.2.3.1.2, 2.5.1.1.6 (particularly at page 2.5-38), 2.5.'4.1.1, 2.5.1.2.8.6, and 2.5.l.2.3.4 treat these matters in detail.
Significantly, joint petitioners' proposed contention contains a serious misstatement of fact. They imply that there are exploratory or production natural : gas wells "within the pertinent radius of the exclusion ::one . "El To the contrary, as discussed in FSAR Section 2.1.2.1, GSU has ownership of mineral rights within~the exclusion area, and there is no such activity presently being conducted (nor contemplated) within the designated - exclusion area for the River Bend Station. Any statement or implication to the
- contrary in the proposed contention is entirely lacking in basis. If, at any time in the future such activity were tc be considered, ' Applicants would appropriately notify the IIRC . Mcwever, to assert that such activity might take place in the future is speculative and need not be considered by l
l-i M/ Contentions by Joint Intervenors at 3.
l 35/ Id.
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this Licensing Board.36/ Thus, no specific deficiency. in .
the application has been demonstrated and this c'ontention should therefore be-denied.
Joint Petitioners' Contention 4 Prematurity of Application 4
Joint petitioners allege that Applicants have failed to
. provide the information required by 10 C.F.R. Parts 50 and l 51 because they filed their application for an operating license "too early into its construction and planning process." 37 / Ho litigable issue is raised by this con-t tention. The contention is based upon an apparent misunder-standing of the regulatory process and the . interaction of 4
the NRC Staff and Licensing Board.
The. question of the acceptability for docketing of an application for an operating license, including the Final Safety Analysis Report and' Environmental Report, is not a matter within the purview of this Lice.13ing Scard. Ratheri Commission regulations assign this matter to the~NRC Staff, which in this instance- has accepted the application for i
docketing. Section 2.101 (a) ( 2) of the NRC 's, regulations
'tator that i .: crder to dete rin2 * - 2 t . 2 r such sppl1 cation i
is " complete and acceptable !cr docketing," an application 36/ Should there at any time in the future be-exploratory or production natural gas wells within such cene, tite Commission's regulations, notably 10 C.F.R. 32.005, prcvic.e adequin .aanns of relief.
H/ Contentions by Joint Intervenors at 3.
for an - operating license _ for .a production or utilization facility will be initially-treated as a tendered application after it is received.
Section 2.101(a) (3) states that this determination is to be made by the Director of Nuclea'r Reactor Regulation,- which necessarily means ,that this Licensing Board does not have ~ any authority over the docketing of the application. 8,/
In any event, petitioners have made no showing that the NRC Staff acted improperly in docketing the application. It is important to note that NRC procedures contemplate that not all information must be furnished at the time an appli-cation for an operating license is docketed. The NRC recognizes that, during the -course of the review, additional e
information will be submitted to complete the application and to - respond to Staff inquiries.39/ The Commission has 38/ See Offshore Power Systems (Floating IAiclear Power Plants), ALAB-489, 3 NRC 194, 202 n.25 (1973).
-39/ The technical specifications are noted as an example of information which has not been submitted. However, the final technical specifications which are issued by the NRC result from the review of the remainder of the application by NRC Staff and only come into focus near the end of this review process. Further, the NRC has promulgated-a set of standard technical specifications for the guidance of applicants. This further diminishes the need for submission of technical specifications by an- applicant at the outset of the review process.
recognized'that this practice.is not inconsistent with the rights of intervenors.40/
Thus, this contention raises no - f actual matters which may be litigated as a contention and it should be denied.
State of Louisiana's Contention 3 Joint Petitioners' Contention 5 Release-of Radioactive Material Through Liquid Pathways It is alleged that Applicants have failed to consider the effect of a release of radioactive material into surface and ground drinking water supplies.41/ The sole basis of these contentions is a general reference to the Mississippi River and an aquifer in the Baton Rouge area. Neither the bases nor the contentions states with specificity any inadequacy in the Final Safety Analysis Report with regard to the discharges from the River Bend Station. In fact, Sections 2.4.12 and 2.4.13 of the FSAR describe in detail the accident analyses- for the River Bend Station related to possible contamination of drinking water sources. Further-more, FSAR Section 15.7.3 contains an assessment of M/ See generally 10 C.F.R. 52.714 (a) (1) (filing of late contentions) ; Catawba, supra, ALAB-687, slip op. at i 14-18; Catawba, supra, " Memorandum and Order" (December 1, 1982) (slip op. at 2-7). In an Order dated December 23, 1982, the Commission indicated that it would review
! this aspect of the ruling in ALAB-687.
-41/ Contentions by Joint Intervenors at 3-4; Supplemental Petition of the State of Lcuisiana at 5-6
(" Supplemental Petition").
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accidental liquid releases into the Mississippi River as affecting surface water uses discussed in Invironmental Report-Operating License Stage ("EROLS")-Section 2.3.2.2 and groundwater uses. discussed in Section 2.3.2.1. Furthermore, Section 5.4 of.the EROLS discusses planned releases, includ-ing compliance with Appendix I to 10 C.F.R. Part 50.
In the Catawba case, the Licensing Board rejected a similar contention that " drinking water' of communities downstream . . .. will become contaminated by radioactive materials accidentally released from Catawba."S The Licensing Board noted that the proposed contention did not even reflect awareness of the discussion of the facility's liquid radwaste system in the FSAR, including its analyses of possible accidents and their effects. The Board there-fore concluded that the " vagueness of this contention provides no basis for arguments about the source or nature of the radioactive materials"SI or how downstream -drink-ing water would be affected, and the contention was accord-ingly denied.
In the Seabrook preceeding, the Licensing Board sini-larly excluded a liquid pathway contention as failing to state why special treatment of liquid pathways should be M/ Duke Power Company (Catawba Nuclear Station, Units 1 and 2), LBP-82-16, 15 NRC 566, 588 (1982).
43/ Id.
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required.E -Similarly, petitioners here have failed to identify any inadequacy in the treatment of accidental releases in the FSAR or EROLS for River Bend and have there-fore failed to specify any particular issue for litigation. 4 Furthermore, petitioners have postulated no mechanism for discharges into the Baton Roage regional aquifer.45/
Thus, this mat'ter lacks specificity and bases and should be denied. To the extent petitioners seek to include matters beyond the design bases accidents as discussed in the River Bend FSAR, the proposed contention is prohibited by the Com-mission's recent Policy Statement on Safety Goals.46/
This contention should be denied.
Joint Petitioners' Contention 6 Generic Safety Issues Joint petitioners assert that Applicants have failed to provide an adequate plan with respect to three unresolved safety issues which the NRC' identified as a result of 44/ Public Service Company of New Hampshire (Seabrook Station, Units 1 and 2), 50-443-OL and 50-414-OL,
" Memorandum and Crder" (September 13, 1952) (.5 lip op.
at 13-14).
M/ To the contrary, EROLS Section 2.3.1.2 discusses that no such mechanism exists.
-46/ See Safety Goal Development Program, 48 Fed. Reg. 10772 (March 14, 1983). See also Pennsylvania Power & Light Company (Susquehanna Steam Electric Station, Units 1 and 2), LBP-79-6, 9 NRC 291, 323 (1979) (denying contention postulating' that " Class 8" accidents involving pipe breaks are more likely to occur than as indicated in applicants' Environmental Report).
investigations of the Three Mile Island accident.47/
Petitioners err in stating that these matters are not addressed on the River Bend docket. In a letter from the NRC dated November 19, 1981, a copy of which is attached for the convenience of the Licensing Board, the NRC generic Issues Branch requested additional information on the status of unresolved safety issues pertaining to the River Bend Station. By letter dated June 3, 1982, Applicants submitted a response indicating in some detail the basis for proceed-ing with the licensing of the River Bend Station should these generic issues remain unresolved at the time an operating license is to be issued for the. Station. Included within such discussion were the three items, Tasks A-45, A-47 and A-48, raised by joint petitioners. Moreover, as encouraged by the NRC, GSU has indicated its intent to adopt the positions of Licensing Review Group II48/ as regards the resolution of these three generic safety issues.
As- the Board is well aware, the unresolved generic
. safety issues will be addressed by the Staff in its Safety Evaluation Report. Regardless of any cont'ention, the i
j Licensing Board is obliged to review the Staff's approach to l
l evary identified unresolved generic safety issue to 41/ Contentions by Joint Intervenors at 4.
.Q/ Licensing Review Group II is an ad hoc croup which includes utilities that own plants similar in design and licensing status to the River Bend Station.
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determine its plausibility and suf.ficiency.49/ Petition-ers have made no allegation that the Applicants' commitment with regard to any unresolved generic safety issue will be unsatisfactory, and any such allegation prior to the availability of the Stapf's Safety Evaluation Report would be premature at best.50/ There is no basis for the assertion that these three matters are not addressed in the application, nor is there any indication as to why these positions are in'any way inadequate. Thus, this contention should be denied.
Joint Petitioners' Contention 7 Cracking of Materials Joint petitioners allege that Applicants have not demonstrated that River Bend Units 1 and 2 meet the require-ments of 10 C.F.R. Part 50 Appendix A, General Design Criteria 4, 14, 30 and 31 "with regard to the adequacy of material selection and control and systems design."b Specifically, in subsection (A) petitioners allege that the use of appropriate materials and processes as specified by NUREG-0313, Rev. l b has not been fully followed. This
( 49/ See, e.g., Jersey Central Power and Light Company (Oyster Creek Nuclear Generating Station), ALAB-645, 13 NRC 1024 (1981).
50/ Seabrook, suora, slip op, at 26-27.
5_1_/ Contentions by Joint Intervenors at 4.
M/ Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping -
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contention is lackina in bases. In . response to Generic Letter 81-03, GSU performed an engineering assessment of the ASME Ccde Class l ', 2 and 3 stainless steel piping as outlined in NUREG-0313, Rev. 1. The.results of this review are contained in letters dated September 4, 1981 and Decem-ber 18, 1981. These two documents' demonstrate compliance with the referenced generic letter and the recommendation of NUREG-0313 (Rev. 1). Thus, Part A of the proffered ' con-tention is without basis.
In Subsection B, it is asserted that the reccmmenda-4 tions contained in NUREG-0619,23/ relating to the instal-lation of a low-flow controller to be used to control the feedwater flow over a range of flows has not been adequately implemented. Contrary to the assertion, the River Bend design incorporates a low flow controller to be used to control the feedwater -flow over a range of flows.54/ The response to Question 410.1855/ addresses compliance with i
Resolution to Generic Technical Activity A-42 (October
! 1979).
-53/ BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking -
Resolution to Generic Technical Activity A-10 (November 1980).
-54/ See FSAR Figure 10.4-7 at coordinates F-2.(valve F002).
Subsequent revision of the text of the FSAR will describe the function of this valve in more detail.
$_5_/ FSAR.at Q&R 4.6-3.
24 -
r NUREG-0619 as to control rod drive nozzles.SI Thus, the River Bend Station is.in agreement with - NUREG-0619. Sub-section (B) of this contention is also without basis and should be denied.
State of Louisiana's Contention 4 Joint Petitioners'~ Contention 8 Old River Control Structure In these two similar contentions, it is alleged that Applicants have not adequately considered the effect of the failure of the Old River Control Structure on the . safe operation of the River ' Bend Station.H It is alleged that an " antiquated and structurally questionable barrier known as the Old River control structure"E could fail thus allowing diversion of Mississippi River water to the Atchafalaya River. The asserted consequence would cause a significant decrease in the amount of water ficwing through the natural course of the Mississippi River. It is alleged
~
56/ See also Section 4.6.1.1.2.4.2.4 of the FSAR. As the Appeal Board has stated, "an intervention petitioner has an ironclad obligation to examine the publicly available documentary material pertaining to the facility in question with sufficient care to enable it to uncover any information that could serve as the foundation for a specific contention." Catawba, supra, ALAB-687, slip op. at 13.
M/ Contentions by Joint Intervenors at 5; Supplemental Petition at 6. The joint petitioners' contention is even less specific than the State's.
58/ Supplemental Petition at 7.
that such a failurb could have consequences on the safe operation of River Bend Station, Unit 1.
Initially, it must be emphatically noted that as thoroughly discussed in the FSAR Sections 9.2.5 and 9.2.7, the design of the River Bend Station is such that it places no reliance upon Mississippi River water as the ultimate
-heat sink for the Station. The Station utilizes a Category I cooling tower which stores sufficient water in its basin to shut down the facility and maintain shutdown for an extended period of time. This system utilizes deep wells and not Mississippi River water for makeup. Thus, the Old River Control Structure could not possibly have any effect on the safe shutdown capability of the plant.
In any event, it is entirely speculative that the control structure would be permi:ted to deteriorate to such a state that it would collapca. Indeed, the Federal courts have rejected the argument that such conjecture is necessary in complying with the National Environmental Policy Act of 1969, 42 U.S.C. S4321, et seg. See, e.g., Warm Springs Dam Task Force v. Gribble, 621 F.2d 1017, 1026 (9th Cir. 1980)
(evaluation of possible dam failure not required). The structure has such a significant effect upon all downstream users, including the cities of Baton Rouge and New Orleans, that the effects on the River Bend Station of its failure would be de minimus by comparison.
There is absolutely no reason given why the NRC cannot license this facility under the present flow regime in the
- -~26 -
l l
Mississippi River, or must assume hypothetically the col- l lapse of the Old River Control Structure. Because safe operation of the River Bend Station would not be affected by the hypothetical loss of the Old River Control Structure, there is really no issue which this Licensing B o a r d -. c a n consider nor any appropriate relief which-it could grant.
The loss of the Old River Control Structure as a contingency on the normal operation of the River Bend Station could only be considered based upon the actual circumstances if that hypothetical event were to occur. Any potential impacts which operation of the River Bend Station might have upon river flows would have to be considered together with all other impacts by the cognizable officials at'that time. To single out River Bend at this time for hypothetical treatment of this speculative event would serve no useful purpose.
As to consideration of the particular impacts alleged, it is noted that matters related to thermal discharges are beyond the jurisdiction of the NRC N and would be re-solved by the appropriate NPDES permit issuing Nuthority for discharges occurring within the State of Louisiana which is at this time the United States Environmental Protection Agency. Finally, the two contentions are entirely lacking 59/ Public Service Company of New Hampshire (Seabrook Station, Units 1 and 2), CLI-78-1, 7 NRC 1, 26 (1978).
t.
tin specificity with regard to the allegation that the salt content of Mississippi River water would be substantially increased should the old River Control Structure fail.
Thus, Applicants submit that these contentions raise no litigable n:tter before the Licensing Board and that such contentions should be denied.
State of Louisiana's Contention 2 Joint Petitioners' Contention 9 Emergency Response Plan In these two contentions,b the State of Louisiana and the joint petitioners allege certain deficiencies in the emergency planning efforts that the State, parishes and Applicants are making with regard to the River Bend Station.
By way of background, it should be noted that the State of Louisiana has already published and placed in operation the State of Louisiana Peacetime Radiological Response Plan.
This plan has undergone review by both the Federal Emergency Management Agency (" FEMA") and the NRC in the context of two completed operating license reviews. It is this plan which is referenced in the present Final Safety Analysis Report for the River Bend Station. Furthermore, the State of Louisiana has extensive experience in the implementation of emergency planning, including large-scale evacuation, as a M/ Contentions by Joint Intervenors at 5-7; Supplemental Petition at 2-5.
result of a number of non-nuclear occurrences within the.
State.
Applicants have been working closely with the State of Louisiana and the involved parishes in order to complete the site and the plume exposure pathway Emergency' Planning Zone
(" plume EPZ") plans for the River Bend Station.61/ It is expected that such planning efforts will be completed by the end of 1983, culminating in the submission of the completed plans for the River Bend Station. These plans include a revised -State of Louisiana Peacetime Radiological Response Plan, which, in turn, will include a specific attachment for planning for the River Bend Station and the involved parish-es. Additionally, designation of the plume EPZ by the officials having jurisdiction is expected to be complete and documented by July, 1983.
Thus, considerable effort has been made in the area of emergency planning for the River Bend Station, and no obstacle to the completion of an emergency plan has been encountered. However, in accordance with the Appeal Board's
, decision in ALAB-687, Applicants submit that it'is premature to entertain general contentions related to emergency planning.
6J/ In addition, planning for the ingestion pathway EPZ is also underway.
_. . _ . _ . _ . _ -- _ . . _ - - . _ m- .__.
Nonetheless,.there are certain' parts of the' contentions related to emergency planning submitted by the ' state of Louisiana and the joint. petitioners'which are impermissible as contentions 'under the Commission's rules.. Applicants submit that-these. items should:be. denied now by.the.Licens- -
- ing Board so -that permissible contentions, if any, in
[- specific areas of emergency planning can be quickly brought
. into focus when the emergency. planning submittals are'made.
For example, joint petitioners allege in subsection (A) 3 that the proposed plant site is in close- proximity to.the Mississippi. River and a regional aquifer, and that a. reactor (
meltdown at River Bend will- endanger these drinking water' sources. This contention 1has no apparent relevance in the .
context of emergency planning and seems to be a reassertion
- of joint petitioners' prop'osed Contention 5. As discussed above, the Commission's Policy Statement on Safety Goals makes it clear that accidents beyond design basis are not to be considered in the . licensing process. This includes planning for emergencies under 10 C.F.R. 550.47 and Appendix E to 10.C.F.R. Part 50.62/ -Alternatively, pet'itioners may be' raising a siting issue which, as discussed previously, is inappropriate at the operating license stage. Therefore, i admission of this proposed contention would be improper
[ under the Commission's rules.
i 6_2] 48 Fed. Reg. at 10779.
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As' a basis for this contention, joint petitioners allege that "[t]he Capital Area Groundwater Conservation Commission ~has petitioned the U.S. Environmental Protection Agency to declare the Baton Rouge Aquifer a sole source 63 /
aquifer under the Federal Safe Drinking Water Act."
The_ legal authority for this petition apparently is Section 1424 (e) of the Safe Drinking Water Act.b A - reading of this section reveals that it is .related to the interim regulation of underground injections into a sole source aquifer. For River Bend, no such injections are contem-plated.S! Thus, this proposed contention is defective and should be denied.
Subsection (B) of joint petitioners' proposed con-tention alleges that the number, location, and capacity of local sheltering facilities and the ' degree of protection from radionuclides afforded thereby are inadequate.66/
Joint petitioners apparently do not understand the Com-mission's requirements with regard to " sheltering facil-ities." Relocation centers must be provided outside the plume exposure EPZ to house evacuees from evac'uated areas.
6_3_/ Contentions by Joint Intervenors at 6.
M/ 42 U.S.C. S300h-3(e).
-65/ 42 U.S.C. S300h (d) (1) defines the term " underground injection" as meaning "the subsurface emplacement of fluids by well injection."
66/ Contentions by Joint Intervenors at 6.
Such facilities are to be located at least five miles from the perimeter of the plume exposure EPZ.67/ Thus, it is clear under the Commission's regulations that relocation facilities would not be required to afford " sheltering" in the sense of protection from radionuclides released from the Station in the event of an accident.
Rather, the protective action of " sheltering" during an emergency would entail instructions by authorities to residents within the plume exposure EP:: or designated portions to stay in their homes. This protective action would be chosen over other alternatives if appropriate to minimize doses to the individual, taking into account actual emergency conditions including protective sheltering factors as they exist. Neither the Ccmmission's regulations nor guidance require the construction of special " fallout" type shelters within the plume exposure EPZ. Thus, this con-tention which alleges that additional protection is required lacks any basis and should be denied.
In subsection (C), joint petitioners attempt to raise the matter of "[t]he heightened sensitivity to~ radiation of children and pregnant women over that of the the average
-67/ NUREG-0654 (Rev. 1), Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants at 63 (November 1980).
healthy adult male."SI This proposed contention is an attack upon the Commission's regulations which adopt the EPA Protective Action Guides for initiation of protective actions.S/ Thus, as an unauthorized challenge to NRC' regulations, this contention should be denied.S Subsections (D) and (E) apparently relate to the designation of the plume EPZ. It is alleged that local meteorological conditions including temperature inversions must be considered. . Applicants submit that local meteorological conditions are not applicable factors dis-cussed in 10 C.F.R. 550.47 (c) (2) with regard to the desig-nation of the plume EPZ. The factors which the NRC has included in this determination are: demography, topography, land characteristics, access routes and jurisdictional boundaries.71/ Other licensing boards have ruled that local meteorological conditions do not play a part in the designation of the plume exposure EPZ inasmuch as M/ Contentions by Joint Intervenors at 6.
M/ See NUREG-0654, supra at 60, which incorporates by reference recomre.endations of the EPA Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (EPA 520/1-75-001).
p/ See generally Metropolitan Edison Comoany (Three Mile
- Island Nuclear Station, Unit No. 2), ALAB-456, 7 NRC l 63, 67 n.3 (1978); Potomac Electric Power Company 1 (Douglas Point Nuclear Generating Station, Units 1 ana l 2) , ALAB-218, 8 AEC 79, 89 (1974).
H/ 10 C.F.R. S50.47 (c) (2) .
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i meteorology was taken into account by the Commission in selecting a 10 mile radius for the generic plume EPZ.
Further, under the Commission's requirements for emergency planning, meteorological conditions existing at the time of an accident would be available to Applicants via the meteorological instrumentation readings at the Station.
Such information would be utilized in determining the protective actions, if any, which had to implemented.
Subsection (F) of joint petitioners' contention appar-ently argues that the Commission's emergency planning criteria utilized a 3200 megawatt thermal reactor as the model for determing the si::e of the plume EPZ and that the River Bend reactor is larger than the model. Without conceding that the ten mile EPZ is inappropriate for all sizes of reactors licensed by the NRC, it is-noted that the
-thermal megawatt rating of each of the River Bend reactors-is 2894. Therefore, each is actually smaller than the one utilized ~in devising the plume exposure EPZ size. If joint i
-72/ See NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants at Appendix I, I-20-26 (December 1978). The Licensing Board in The Cincinnati Gas & Electric Comnany (Wm. H. Zimmer Nuclear Power Station), Docket No. 50-358, excluded testimony on local meteorology as it allegedly related to selection of the size of the plume exposure EPZ, finding that the emergency planning regulations had taken this matter into account (Transcript of hearing at 5296-5301 (January 22, 1982)).
petitioners are asserting that the' thermal rating for both units must- be considered cumulatively, the contention -
represents a challenge to the Commission's regulations i.e., that simultaneous accidents in reactor units need not be considered.73/
While Applicants believe that the assertions of sub-section (G) are without merit and raise -questions that in effect challenge the Safety Goal rulemaking, these arguments are more appropriately considered when the plans for the parishes have been submitted. Thus, Applicants will not further respond to them at this time since they are prema-ture.
Applicants' objections to the State of Louisiana's Contention 2 are similar. It should be pointed out that agencies of the State of Louisiana and the parishes have been working towards the selection of a plume exposure EPZ specific to the site around the River Bend Station, which is not completely circular in configuration. Further, Appli-cants note that plans are being made to assure that adequate protective measures can be taken for the ' institutions i enumerated in the State of Louisiana's petition. It should be clear from the outset that such protective actions need not as a matter of necessity include evacuation depending l-H/ 10 C.F.R. Part 50, Appendix A, Criterion 5.
l t
upon the. type . of facilities and their locations. .Again, J
~
this is a matter to.be_ resolved at a later point.
Joint Petitioners' Contention 10 Construction State Joint petitioners ask that the record as to' contentions be kept open until 15 days before the regulations found at 10 C . F . R. . S2.714(b). Initially, this is not a factual contention that may be ' considered by the Licensing Board.
In addition, the Board has already decided this matter. It has permitted several extensions of time,74/ totalling almost nine months, during which the joint petitioners could have submitted additional. and more specific contentions.
That time.has now. expired without any effort by petitioners to take advantage of-the period afforded them by the Licens-ing Board to revise and ' further specify their contentions.
The Commission's regulations contemplate that the prehearing conference would be held within 90 days after the notice of opportunity for a hearing.75/ Thus, the Commission has deemed that period of time as a reasonable one within which to formulate specific contentions. In this proceeding, it uns not unreasonable for the Board to have established a i
cut-off for contentions some 18 months after the initial I
-74/ See Memorandum and Order at 3 (July 30, 1982); order at L 2 (August 20, 1982); order at 2 (December 21, 1982).
75_/ 10 C.F.R. 52.751a.
i 7
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notice of hearing ' in . the Federal Register. The joint petitioners.have not demonstrated otherwise. No legitimate claim for further extensions has been made, and petitioners' request should be denied.
Joint Petitioners' . Contention 11 Potassium Iodide Tablets It is alleged that Applicants have not providad for the distribution and storage of potassium iodide in accordance with. accepted public health practice.E! Applicants submit that the decision as to whether to utilize potassium iodide for those present within the plume exposure EPZ is a matter that is left to the discretion of the states, and
'that the Applicants play no part in such decisionmaking.78/ The Commission has accepted in a number of cases the decisions of cognizable state officials either in favor of or against the provision of potassium iodide.79/ In fact, in a case where the EP" falls within the boundaries of two states, one state, constituting approximately one-half of the plume exposure EPZ, decided to utilize potassium iodide for residents and the other state H/ 46 Fed. Reg. 44539 (September 4, 1981).
H/ Contentions by Joint Intervenors at 8.
H/ Thus, NUREG-0654 only requires state and local agencies to identify their plans or intentions regarding the possible administration of pctassium iodide. See NUREG-0654, suora at 63.
H / [ Citation]
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in the plume exposure EPZ did not. Thus, Applicants submit there_is no litigable issue. The decision of the emergency planning. authorities.for the particular. state to provide or not to provide potassium iodide is binding upon this Licens-Ing Board.80/
Joint petitioners allege that monitoring should be expanded to include the human population residing within the ingestion pathway of Iodine-131. While not clear as to what is being requested, it should be noted that plans .for monitoring during an incident do call for monitoring Io-dine-131.81/
No litigable issue has therefore been raised.
Joint Petitioners' Contention 12 Funds for Premature or Early Decommissioning 82/ -
The matter of the financial qualifications of the Applicants has been completety addressed in the response to the joint petitioners' Contentica 1. This contention, which is similarly a prohibited c tack on the regulations, and-
~~80/ It was reported in 1982 that FEMA had decided not to create a national stockpile of potassium iodide despite a Congressional appropriation for that purpose. A FEMA representative reportedly stated that this decision was based upon widely divergent views by the states as to the necessity for its distribution. See 23 Nucleonics Week 1 (October 14,-1982).
81/ See NUREG-0654, supra at 18 (Table 3), 58.
82/ Joint petitioners -misnumbered this as a second Contention 11. Contentions by Joint Intervenors at 8.
For reference, Applicants have designated it Contention 12.
even more vague,'should likewise be denied. In the Seabrook I proceeding, a contention which raised " questions about the
' financial capability . . of applicant. to safely decommission" was rejected on the basis of 'the- new rule eliminating financial qualification of applicants as an issue in operat-ing license proceedings. b It should also.be noted that the matter of decommis-sioning is presently under review by the Commission in a rulemaking - proceeding 84 / and thus is not appropriate for consideration in individual adjudicatory proceedings..
Moreover, neither is the availability of repositories for the long-term storage of spent fuel an appropriate subject of a contention. In the Limerick proceeding, the Licensing Board ruled that the Commission had expressly prohibited such contentions in individual licensing proceedings in light - of its rulemaking.85/ Further, subsequent to the filing of this contention, Federal legislation entitled the Nuclear Waste Policy Act of 1982, Public Law No.97-425 (January 7, 1983) was enacted to deal with specific arrange-ments regarding the disposal of reactor fuel. This con-tention raises no litigable issue and should be denied.
83_/ Seabrook, supra, slip op. at 95.
84/ 47 Fed. Reg. at 13751.
8_5,/
5 L i m e r i c k , supra, 15 NRC at 1455. See also Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station), ALAB-655, 14 NRC 799, 816 (1981).
. 4 Joint Petitioners' Contention 13 86/
State of Louisiana Contention F Construction of Unit 2 During' Unit 1 Operation In these contentions, joint petitioners and the State of Louisiana allege that Applicants have failed to consider the effect of future construction activities for Unit 2 on the safe operation of Unit 1. The State cites 10 C.F.R.
S50. 34 (b) (6) (vii) as basis for its contention, and correctly asserts that Unit 2 is currently not scheduled and con-struction on that unit has been halted. Moreover, Appli-cants will make no decirion with regard to Unit 2 unt:il late 1985.
Under these circumstances, it is entirely reasonable for Applicants not to have submitted detailed plans with regard to the construction of Unit 2. Inasmuch as the timing, schedule and number of construction workers could not be determined with any specificity, no specific plan could be submitted at this point which would satisfy the cited regulation. As is obvious, the NRC Staff would need to review and approve such plans only prior to the resump-tion of Unit 2 construction. It is there fore premature to consider this contention, which should be denied. The Commission has adequate remedies both within and outside its adjudicatory hearing process for this matter to be raised 86/ This contention has been renumbered. See n.82, supra.
'should. Unit'2 construction.be recommenced. Applicants,.of.
course, 'will notify the NRC'of their intent to resume: con--
struction of Unit 2.87/
State of Louisiana Contention 1 Table S State of Louisiana ' Contention 1 alleges that
"[alpplicants have ' failed to allow for proper consideration of the uncertainties concerning the long-term - isolation of high-level and transuranic wastes, and ... failed [ed] to allow for proper. consideration of the health, socioeconomic and cumulative effects of fuel-cycle activities.'" 8/ The contention is based entirely upon a recent decision of the
-Court of Appeals for the District of Columbia Circuit which invalidated the Commission's rule for consideration of the fuel cycle in individual reactor licensing proceedings, the so-called Table S-3 to 10 C.F.R. Part 51.89/ Commission precedents as well as its specific instructions in this
~
-87/ Applicants would note that the Cornission has recently rejected a rulemaking petition filed by Wel.ls Eddelman,.
47 Fed. Reg. 46524 (October 19, 1982), whi'h c would have recuired ~ separata operating license hearings fcr multi-unit. facilities. Applicants submit that this action confirms the propriety of considering the issuance of operating licenses for.both Units 1 and 2, recognizing that the particular information regarding Unit 2 may be properly submitted at a later time.
8_8_/ Supplemental Petition at 2. I 89_/ Natural Resources Defense Council, Inc. v. Nuclear Regulatory Commission, 685 F.2d 459 (D.C. Cir. 1982),
cert, granted, 51 U.S.L.W. 3419 (November 29, 1982)
'(No.82-545) ("S-3 decision").
g instance are clear that, pending completion of Supreme Court review, this matter may not be litigated in individual licensing proceedings.
The State of Louisiana attempted to raise precisely this issue in Mississippi Power &_ Light Company (Grand Gulf Nuclear Station, Units 1 and D ALAB-704, 16 NRC (December 8, 1982). The Appeal Board affirmed the Licensing Board's denial of Louisiana's intervention petition which-was based, as here, on the Table S-3 decision. There, the Appeal Board concluded that " guidance of the Commission leaves no room for doubt that the question of safe waste disposal as reflected in the S-3 table of effluent releases is not a matter for case-by-case litigation in individual reactor licensing proceedings at this time."El The Appeal Board based its decision in large part upon a November 8, 1982 Commission policy statement with regard to the District of Columbia's S-3 decision.b! Therein, the Commission determined that Table S-3 issues could not be litigated on a case-by-case basis:
To move further toward case-by-ca'se litigation would reintroduce the signif-icant burdens the rule was intended to relieve. Use of the S-3 rule has served the important purpose of providing the 9_0/ Grand Gulf, supra, slip op. at 12.
-91/ " Statement of Policy, Licensing and Regulatory Policy and Procedures for Environmental Protection; Uranium Fuel Cycle Impacts," 47 Fed. Reg. 50591 (November 8, 1982) ("S-3 Policy Statement").
. e underlying basis for consideration of fuel cycle impacts, and the Commission believes that an attempt to proceed without the rule would probably prove unworkable . . . . The resulting delay and drain on staff resources would be substantial, and would not on7.y delay licensing of qualified facilities, but would also substantially disrupt the Commission's regulatory program, includ-ing its program -o develop safety standards for high-level waste disposal facilities. 92/
The Appeal Board has directed that Licensing Boards "act as if the District of Columbia Circuit's decision, which is now under review by the Supreme Court, is currently of no operative effect.93,/-
Thus, this contention must be denied.
State of Louisiana Contention 6 Asiatic Clams It is alleged that Applicants have failed to provide adequate assurance that River Bend Station components and systems relying on Mississippi River water for their opera-tion will be adequately protected against infestation of Asiatic clams (Corbicula leana). With regard to this contention, the State has attached a-letter of February 14, 1983 from the Applicants which addresses the question of f
! steps necessary to control any infestation of Asiatic clams including details on chlorination levels. Compared to the l
i l 92/ 47 Fed. Reg. at 50592 (footnote omitted).
M/ Grand Gulf, supra, slip op. at 12-13.
I l
i l
o
,,-- e Applicants' specific discussion of the matter in that letter and in Section 3.6.1.3.2 of the EROLS, the State. of Louisiana merely asserts without basis - of any kind that Applicants have failed to demonstrate that infestation can-and will be controlled. This contention is completely lacking in specificity and bases and should be rejected.
Conclusion As discussed-in detail above, the . contentions of both the State of Louisiana .which as intervened through the office of Attorney General and the joint petitioners are defective and should not be considered for litigation before this Licensing Board. Furthermore, a number of the matters raised are premature and under the Commission's precedents must be rejected at this time.
Respectfully submitted, CONNER & WETTERHAHN, P.C.
!]
Mark J. Wetterhahn Counsel for the Applicants April 15, 1983 -
ee
,(
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )
)
GULF STATES UTILITIES ) Docket Nos. 50-458 OL COMPANY, et al. ) 50-459 OL
)
(River Bend-Station, Unit 1 )
-and 2 )
SERVICE LIST I hereby certify that copies of " Applicant's Answer to-the Contentions Filed by Petitioners for Leave to Intervene, State of Louisiana through the Office of the Attorney General, Louisiana Consumer's League, Louisianans for Safe Energy, Inc. and Gretchen Reineke Rothchild" dated April 15, 1983, in the captioned matter, have been served upon the following by deposit in the United States mail this 15th day of April, 1983:
B. Paul Cotter, Jr., Esq. James W. Pierce, Jr., Esq.
Chairman, Atomic Safety and P. O. Box 23571' Licensing Board Baton Rouge, LA 70893 U.S. Nuclear Regulatory Commission Doris Falkenheiner, Esq.
Washington, D.C. 20555 Stephen M. Irving, Esq.
355 Napoleon Street Dr. Forrest J. Remick Baton Rouge, LA 70802 305 East Hamilton Avenue State College, PA 16801 William Guste, Jr., Esq.
Attorney General' Dr.-Richard F. Cole State of Ecuisitna Atomic Safety and Licensing 234 Loyola Avenue Board New Orleans, LA 70112 U.S. Nuclear Regulatory Commission Ian D. Lindsey, Esq.
Washington, D.C. 20555 Department of Zustice 7434 Perkins Road David.A. Repka, Esq. Suite C Counsel for MRC Staff Baton Rouge, LA 70803 Office of the Executive Legal Director Decketing & Service Section U.S. Nuclear Regulatory U.S. Nuclear Regulatcry Commission Commission Washington, D.C. 20555 Washington, D.C. 20555
0' Linda B. Watkins, Esq. Gulf States Utilties
- , 355 Napoleon Street Company Baton Rouge,-LA.70802- . Attn
- Zir._ James E.. Booker.
Manager - Engineering.
and Licensing
, P. O. Box 2951 Beaumont, Texas 77704
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Matr) J . Wetterhahn i
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'g q'r ,g NUCLEAR REGULATORY COMMISSICN -
2g.y .3 wAssmGTON. D. C. :C555 by ~
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Docket Nos: 50-458 and 50-459 -
'83 APR 18 A10:08
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Mr. William J. Cahill, Jr d iscissW Senior Vice President - River Bend Nucl, ear Grospi Gulf States Utilities Company Fost Office Box 2951 Beaumont, Texas 77704
Dear Mr. Cahill:
Subject:
Requests for Additional Iaformation Regarding the Status cf Unresolved Safety Issues The Generic Issues Branch has identified a need for additional information regarding the status of Unresolved Safety Issues pertaining to River Eend. This informational request is provided as enclosure (1). Your response to enclosure (1) should be provided no later than June 1,1982.
Enclosure (2) is the Generic Issues Branch SER contribution for a
- recent BWR plant, Grand Gulf. This enclosure is provided for your information and to assist you in your rescanses.
Sincerely, [
7 la .
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A. dchwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosure:
As stated j c:: See next page
--..------.J
Mr. William J. Cahill, Jr.
I Senior Vice President . j River Bend Nuclear Group l Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 cc: Troy B. Conner, Jr., Esquire Conner and Wetterhahn 1747 Pennsylvania Avenue, N. W.
Washington, D. C. 20006 Mr. J. E. Booker Manager -Technical Programs Gulf States Utilities Company Post Office Box 2951 Beaumont, Texas 77704 Stanley Plettman, Esquire Orgain, Bell and Tucker Beaumont Savings Building Beaumont, Texas 77701 Karin P. Sheldon, Esquire Sheldon, Harmon & Weiss _
1725 I Street, N. W.
Washington, D. C. 20006 William J. Guste, Jr., Esquire Attorney General State of Louisiana Post Office Box 44005 State Capitol Baton Rouge, Louisiana 70804
- Richard M. Troy, Jr. , Esquire Assistant Attorney General in Charge State of Louisiana Department of Justice 234 Loyola Avenue New Orleans, Louisiana 70112 -
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Enclosure 1 ...
REQUEST FCR INFCRMATICN ,
The Atemic Safety and Licensing Appeal Board in ALAB 24a determined that .
the Safety Evaluation Repcrt for each plant should contain an assessment of each significant unresolved generte safety question. It is the staff's view that the generic issues identified as " Unresolved Safety ~
Issues" (NUREG-0606) are the substantive safety issues referred to by the Appeal Board. Accordingly, we are requesting that ycu provide us with a summary description of your relevant investigative pregrams and the interim measures you have devised for dealing with these issues pending the comaletion of the investigation, and wnat alternative courses of action might be available shculd the program not produce the envisaged result.
There are currently a total of 26 Unresolved Safety Issues discussed in NUREG-C606. We do not require information frem you at this time for a number of the issues since a number of the issues do not acoly to your type of reactor, or Secause a generic resolutien has Seen issued.
Issues wnich have Seen resolved have been or are being incor: orated into the NRC licensing guidance and are addressed as a part cf the normal review arecess. Mcwever, we do request the information noted above fer each of the issues listed Selcw:
- 1. Waterham.er (A-1)
- 3. Reactor Vesset Materials Toughness (A-11; 4 Systems Interaction in Nuclear Pcwer Plants (A-17)
- 5. Safety Relief Yalve Pool Oynsnic Loads (A-39) -
- 6. Seismic Design Criteria (A 40)
- 7. Centainment Emergency Suma Reliability (A 43)
- 3. Station Blackcut {A 44)
- 9. Shutdown Decay Heat Removal Requirements (A 45)
- 10. Seismic Cualificatien of Ecuf pment in 0:erating Plants (A 46)
- 11. Safety fantications of Control Systems (A-17)
- 12. Hydregen Control Measures and Effects of Hydrogen Burns :n Safety Equipment (A 43) ,
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Enclosure 2 , , ,
APPENDIX C MJCLEAR RE3ULATORY COMMISSION (NRC)
UNRE3OLVED SMETY ISSUE 5 C.1 Unresolved Safety Issues The NRC staff ct.ntinuously evaluates the safety recuirements used in its reviews against new infornat2n as it becomes available; Infomation related to the safety of nuclear power plants comes frem a variety of sources including experience frem coerating reactors; research results; NRC staff and Advisory Connittee on Reactor Safeguards (ACRS) safety .
reviews; and vendor, architect / engineer and utility design rev Dws.
Each time a new concern or safety issue is identified fecm one or more of these sources, the need for imediate action to assure safe cceration is assessed. This assessment includes consideratien of the generic imolications of the issue. .
In some cases, imediate action is taken to assure safety', e.g. , the derating of boiling water reactors as a result of the channel box wear problems in 1975. In other cases, interim measures, such as modifications to ocerating procedures, may be sufficient to allow further study of the issue criar to making licensing decisions. In most cases, however, the initial assessment indicates that immediate licensing actions or chances in licensing criteria are not necessary. In any event, further study may be deemed acar:griate to make judgments as to wnetter existing NRC staff recuirements shculd be mcdified to address the issue for new plants or if backfitting is acorcoriate for the long term cperation cf _
plants already under construction er in coeration.
These issues are scmetimes called " generic safety issues" because they are related to a particular class or type of nuclear facility rather than a specific plant. Certain of these issues have been designated as
" unresolved safety issues" (NUREG-CA10, "NRC Program for the Resolution 1 of Generic Issues Related to Nuclear Power Plants," dated January 1, 1978). Mcwever, as discussed above, such issues are cons.idered on a ceneric basis only after the staff has made an initial determination
- f. hat tne safety s'ignificance of the issue coes not :renibit continued operation or recuire licensing actions wnile the longer-term generic review is unde nay.
C.2 ALAB aLa Recuirements These longer-term generic studies were the subject of a Cecision by the Atomic SafetyThe Commission. and Licensing "ecision was Acceal Board issued en of the 23,1977 Novemcer Nuclear (ALAB Reculatory aaa ) in cennection with the Ac;eal Boarc's consideration of tre Gulf States Utility Cemaany acolication for :he River 3end Station, Unit Nos. I and 2.
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_ - - - - - - 4., _ - . . ._ _. . . . _ _ _ ___ . _ __..__ _ _ . ._ __ _
. , g "In short, the board (and the public as well) should be in a position
- to ascertain frem the SER itself--without the need to resort to extrinsic documents-tne staff's perception of the nature and '~
extent of the relationship between each significant unresolved generic safety question and the eventual operation of the reactor under scrutiny. Once again, .this assessment might well have a direct bearing upon the ability of the licensing board to make the safety findings required of it on the construction oermit level even though the generic answer ta the cuestion remains in the offing. Among other things, the furnished information would likely shed light on such alternatively imoortant considerations as wnether:
(1) the problem has already been resolved for the reactor under study; (2) there is a. reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is put in operation; -
or (3) the problem would have no safety imolications until after several years of reactor coeration and, should it not be resolved by then, alternative means will be availacle to insure.that continued coeration (if permittad at all) would not pose an undue risk to the -
public."
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This accendix is soecifically included to rescend to the decision of the Atomic Safety and Licensing Acceal Scard as enunciated in ALAB , anc -. .
as applied to an ccerating license proceeding Vircinia Electric and l Power Comnanv (North Anna Nuclear Power Statien, uni: Nos 1 anc 2), _ ~
XDil~491, NRC 245 (1978). ~
C.3 " Unresolved Safety Issues" In a related matter, as a result of Congressional action en the Nuclear Regulatory Ccmission bud Act of 1974 was amended PL (get for on 95-209) Fiscal Year 1978, Decemoer 13, 1977 thetoEnergy include,Reorganization among other thines, a new Section 210 as folicws:
"UNRESCLVED SAFETY ISSUES PLAN" "SEC. 210. The Comission shall develop a plan providing for -
specification and analysis of unresolved safety issues relating to nuclear reactors and shall take such action as =ay be necessary to implement corrective measures with resoect to such issues. Such plan shall be sucmitted to the Congress en or before January 1, 1978 and progress recorts shall be included in the annual report of
- the Cc missicn thereafter.* _
The Joint Explanatory Statement of the House-Senate Ccnference Cemittee for the Fiscal Year 1978 Acorocriations Bill (Bill 5.1131) provided the following additional infomation regarding the Cemittee's deliberations on this portion of the bill:
"SECTION 3 -UNRESCLVED SAFETY ISSUES" "The House amendment recuired develcpment of a plan to resolve generic safety issues. The conferees agreed to a recuirement that ,
the plan be submitted to the Congress en or before January 1,1978.
The conferees also expressed the intent that this plan should identify and describe those safety issues, relating to nuclear pcwer reactors, which are unresolved on the date of enactment. It should set forth: (1) Comissien actions taken directly er indirectly to develop and implement conective measures; (2) further actions olanned cencarning such measures; and (3) timetables and cost estimates of such actions. The Ccmission shculd indicate the priority it has assigned to each issue, and the basis en wnich priorities have been assigned." ,
l In resconse to the reporting requirsents of the new Section 210, the
! NRC staff submitted to Ccngress en January 1,1978, a report, NURE3- .
j C410, entitled "NRC eregram for the Resolutien of Generic Issues Related to Nuclear Power Plants," describing the NRC generic issues pr gram.
The NRC program was already in place when PL 95-209 was enacted and is l of censiderably broader scoce than the " Unresolved Safety Issues Plan" l required by section 210. In the letter transmitting NUREG-0410 to the '
l Congress on Cecemcer 30, 1977, the Ccmission indicated that "-he progress l reports, wnich are required by Section 210 to be. included in future NRC l annual recorts, may be more useful to Congress if they fccus en *ne specific Section 210 safety items."
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It is the NRC's view that the intent of Section 210 was to assure that plans were develoced and i=clemented on issues with :otentially significant '"
public safety implications. In 1978, the NRC undertcok a review of over 130 generic issues adoressed in the NRC program to determine wnica '
issues fit this description ano cualify as " Unresolved Safety Issues" for reporting to the Congress. The NRC review included the develocment of proposals by the NRC Staff and review and final aaproval by the NRC 4
Commissioners.
This review is described in a recort NUREG-C510, " Identification of Unresolved Safety Issues Relating :o Nuclear Power Plants - A Report to Congress," dated January 1979. The report provides the -following definition of an " Unresolved Safety Issue:"
"An Unresolved Safety Issue is a matter affecting a number of .
nuclear oower clants that cases imocrtant cuestions concerning the adequacy of existing safety requirements for which a final resolution '
has not yet been developed and that involves c:nditions not likely to be accapable over the lifetime of the plants it affects.*
Further the report indicates that in apol,ying this definition, matters that pose "important questions cencarning the adequacy of existing safety requirements" were judged to be those for wnich resolution is necessary to (1) compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety. Cuite simoly, an " Unresolved Safety Issue" is potentially significant from a public safety stanc:oint anc its resolution is likely to result in NRC action on the affected plants. ,
All of the issues adcressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510. As a result, seventeen " Unresolved Safety Issues" addressed by twenty-too tasks in the NRC program were identified. The issues are listad beicw.
progress on these issues was first discussed in Se 1978 NRC Annual Recort. The number (sl of the generic task (sl (e.g., A-1) in the NRC program addressing each issue is indicated in parentheses f.ollowing the titl e.
"UNRESCLVED SAFETY ISSUES" (APPLICA3LE TASK NOS.)
- 1. Waterhammer-(A-1)
- 2. Asynnet-ic Blcwdown Loads on the Reactr- Ccolant System' - (A-2)'
- 3. Pressuri:sd Water Reactor Staan Generater Tube Integrity - (A-3, A-4,A-5)
- 5. Anticipated Transients Without Scram - (A-3)
- 6. 3WR No::le Cracking - ( A-10) .
- 7. Reactor Vessel Materials Toughness - (A-11)
- 8. Fracture Tougnness of Steam Generator and Reactor C:olant Pu=p Succorts - (A-12)
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- 9. Systems Interaction in Nuclear Power Plants - (A-17)
- 10. Environmentai Oualification of Safety-Related Electrical Ecuiement - --
(A-24)
- 11. Reactor Vessel Pressure Transient Protection - (A-25) -
- 12. Residual Heat Removal Requirements - (A-31)
- 13. Control of Heavy Loads Near Scent Fuel - (A-26) 14 Seismic Design Criteria - (A-40)
- 15. Pipe Cracks at Boiling Water Reactors - (A 42)
- 16. Containment Emergency Sump Reliability - (A 43)
- 17. Station Blackout - (A 44)
In the view of the staff, the ' Unresolved Safety Issues'" listad above are the substantive safety issues referred to by the Anpeal Board in ALA3-444 wnen it socke of "... those generic problems under continuing study which have.... potantially significant public safety implications." -
Six of the twenty-two tasks . identified with the " Unresolved Safety Issues" are not acclicable to Grand Gulf because they acoly to pressurized watar reactors only. These tasks are A-2, A-3, A 4, A-5, A-12, and A-
- 25. Also, tasks A-6, A-7, and A-8 only apoly to Mark I or Mark II boiling water reactor containments. With regard to the remaining 13 -
tasks that are acplicable to Grand Gulf the NRC staff has issued NUREG recorts providing its resolution of five of the issues. The table below lists those issues.
Task Number NUREG Recort and Title SER/SER Sucol . Section(s)*
A-10 NUREG-0619, " SWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" A-24 NUREG-0588, Revision 1,
' Interim Staff Position on Environmental Qualification of Safety-Related Electrical Ecuipment" A-31 SRP 5.4.7 and 3TP 5-1 "Residuai Heat Removal Systems" incor; orate ,
recuirements of US! A-31.
A-36 NL' REG-C612, dControl of Heavy .
Loads at Nuclear Power Plants" A 42 NUREG-0313, Revision 1.
"Tecnnical Recort on Material Selection and Precassing Guide-lines for SWR Coolant Pressure Soundary Piping"
'Not avaliacie at :his time. To be provided by the Project Manager.
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The remaining issues applicable to Grand Gulf are listed in the following --
tabl e.
GENERIC TASKS ACORESSING -e "UNRE50LtED SAFETv ISSU ES" .
THAT ARE APPLICABLE FG GRAND GULF UNIT 5 i AND 2
- 1. A-1 Water Hamer
- 2. A-9 ATWS
- 3. A-11 Reactor Vessel Materials Toughness
- 4. A-17 Systems Interaction in Nuclear Pcwer Plants
- 5. A-39 Safety Relief Valve Pool Dynamic Loa'ds S. A-40 Seismic Cesign Criteria
- 7. A-43 Containment Emergency Sumo Reliability -
- 8. A-44 Station Slackout With the excection of Tasks A-9, A 43, and A-44, Task Action Plans for the generic tasks above are included in NUREG-C649, " Task Action Plans for Unresolved Safety Issues Reiated to Nuclear Power Plants.* A tachnical .
resolution for Task A-9 has been procosed by the NRC staff in Volt =e 4 cf NUREG-0460, issued for comment. This served as a basis fcr the staff's precosal for rulemaking on this issue. The Task Action Plan for Task A-43 was issued in January 1981, and the Task Action Plan for A 44 was issued in July 1980. The information provided in NUREG-0649 meets most of the infornational recui'cments of ALAB "'. Each Task Action Plan provides a description of 2.e problem; the staff's acproaches to its resolution; a general discussicn of the bases ucon which centinued plant licensing or cceration can r. ceed pending ecmpletion of the task; the technical organizations invo'.ved in the task and estimates of the mancower recuired; a description of the interactions with other NRC -
officas, the Advisory Ccmmittee on Reactor Safeguards and outside crganizations; estimates of funding required for contractor-sucplied tecnnical assistance; prospective estes for ccmoleting the tasks; and a description of potential problems tnat could alter the planned apcrcach or scnedule.
In additicn to the Task Acticn Plans, the staff issues the "Acua 3cok" (NUREG-0606) on a quarterly basis. This bcck entitled, 40ffice cf Nuclear Reac r Regulation Unresolved Safety Issues Summary, Acua Scok,*
crovices current schedule information for eacn of the " Unresolved Safety Issues." It aisc includes information relative to the implementation status of each " Unresolved Safety Issue" for which technical resolution is ccmoleta.
We have reviewed the eight " Unresolved Safety Issues" listed above and the four new USIs discussed in Section C.4 as they relate to Grand .
Gulf Units 1 and 2. Discussion of each of these issues including references to related discussions in the Safety Evaluatien Recor is crovided belew in Secticn C.5. We have satisfactorily concluced cur review for all but .
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the A-39, " Mark III Safety Relief Valve Fool Dynamic Loads" issue.
That issue is currently incemolete. We will discuss resolution of
- this issue in a supplement to the Safety Evaluation Recort. Based on our '
review of these items, we have cencluded, for the reasons set fortn in '
Section C-5, that with the exception of A-39 there is reasonable assurance
- that the Grand Gulf Unit Nos.1 and 2 can be coerated prior to the ultimate resolution of these generic issues without endangering the health and safety of the public.
C.1 New "Unrasolved Safety Issues" ,
An in-depth and systematic review of generic safety concerns identified .
since January 1979 has been perfomed by the staff to determine if any of these issues should be designated as new " Unresolved Safety Issues." -
The candidate issues originated from concerns identified in NUREG-0660, "NRC Action Plan as a Result of the TMI-2 Accident," ACRS rec:mendations, abnormal occurrence reports, and other'coerating experience. The staff's proposed list was reviewed and cemented on by the ACRS, the Office of Analysis and Evaluation of Ooerational Data (AECD) and the Of' ice of -
Felicy Evaluation. The ACRS and AE00 also orososed that several additional
" Unresolved Safety Issues" be considered by the Comission. The Comission considered tne above information and aaproved the following foun new
" Unresolved Safety Issues:"
A-4S Shutdown Cecay Heat Removal Reouirements
.A 46 Seismic Gualification of Ecuicment in Ocerating Plants A 47 Safety Imolication of Control Systems A 48 Hydrogen Cont ci Measures and Effects of Hydrogen Burns on Safety Etuipment A description of the above process together with a list of the issues considered is cresentad in NUREG-0705, " Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Scecial Report to Congress,"
dated March 1981. An excanded discussion of each of the new " Unresolved Safety Issues' is also c:ntained in NUREG-070s.
l The acclicability and bates for ifcensing crior to ultimate resolution of the- four new USIs for Grand Gulf Units 1 and 2 are discussed in Section C.5.
! C. 5 Discussion of Tasks as Thev Relate to Grand Gulf
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This section provides the NRC staff's evaluation of the Grand Gulf facilities for each of the acclicable Nnresolved Safety Issues." This includes our bases for licensing prior :: ultimate resolutien of these issues. Our I
conclusions are based in part on info m atien provided by the acclicint
- i in their letter of August 7,1981 frem L. F. Cale, tiississicci Power and Light Cemeany t: Robert L. Tedesco, NRC.
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s A-1 Waterhamer .
Waterhamer events are intense pressure pulses in fluid systems caused ~
by any one of a number of mechanisms and system conditions such as raoid condensation of steam pockets, steam-driven slugs of water, pumo startuo '-
with partially emoty lines, and rapid valve motion. Since 1971 over 200 incidents involving waterhamer in pressurized and boiling water reactors l have been reported. The waterhammers (or steam hamers) have involved steam generator feedrings and piping, the residual heat removal systems, emergency core cooling systems, and ccntainment spray, service water, feedwater and steam line.
Most of the damage reoorted has been relatively minor, ' involving oice hangers and restraints; however, several waterhamer incidents have resultad in piping and valve damage. The most serious waterhamer events have occurred in the steam generator feedrings of pressuri:ed -
water reactors. In no case has any waterhamer incident resulted in the release of radioactive material.
Under generic Task A-1, the potential for waterhamer in iarious systems is being evaluated and aopreoria*e recuirements and systematic review -
procedures are being develcped to ensure that waterhamer is given appreoriate consideration in all areas of ifcensing review. A technical report, NUREG-0582, "'4ater-hamer in Nuclear Pcwer plants" (July 1979),
providing the results of an NRC staff review of waterhamer events in nuclear power plants and stating staff licensing positions, c:moletes a major subtask of Generic Task A-1.
Although waterhamer can occur in any light water reactor and over 100 actual and probable events have been reported in boiling water reacters, none have caused major pipe failures in a boiling water reacter such as Grand Gulf and none have resulted in the offsite release of radicactivity. -
As noted above, the most severe waterhammers coserved to date have been in steam generators. Since the boiling water reactor does not utili:e a steam generator, these worst cases are eliminated. Furthermore, any waterhamar which may occur in feedwater or main steam piping will not imoair the emergency core c: cling system since all ECCS water enters the reactor vessel via five separate reacter vessel noz:les indecendent of the feedwater and main steam piping.
Grand Suif has install 2d a systam : preclude satarht=er #-m cccurring in emergency core cooling system lines. This system consists of jockey cumos to keen the emergency core ecoling systam lines water-filled so that the emergency core cooling system pumos will not star: cumaing into voided lines and steam will not collect in the emergency core cooling system piping. To ensure that the emergency core cooling system lines remain water-fiiTed, vents have been installed and a Technical Scecificatien recuirement to periedically vent air frem the lines has been imocsed.
Further assurance for filled discharge piping is provided by cressure instrumentaticn at the pioing high points. An alarn sounds in :he main control recm if the cressure falls belcw a preceterninec se :oint indicating -
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difficulty maintaining a filled discharge line. Should this oc:ur, or --
if an instrument becomes inocerable, the recuired action is identified in the Technical Specifications. --
With regard to aeditional protection against potential waterhammer -
events currently provided in plants, piping design codes require consideration of impact loads. Aaproaches used at the design stage include: (1) increasing valve closure times, (2) ciping layout to preclude water slugs in steam lines and vapor formation in water lines,- (3) use of snubbers and pipe hangers, and (4) use of vents and drains.
In addition, we recuire that the apolicant conduct a preccerational vibration dynamic effects test prcgram in accordance with Section III of the American Society of Mechanical Engineers Code for all Class 1 and Class 2 piping systems and pioing restraints during startuo and initial
- operation. These tests will provide adequate assurance that the oiping and piping restraints have been designed to withstand dynamic effects due to valve closures, pump trips, and other ocerating medes associated with the design coerational transients.
Nonetheless, in the unlikely event that a large pice break did result from a severe waterhammer event, c:re cooling is assured by the amergency core cooling systems and protection against the dynamic effects of such pipe breaks inside and outside of containment is provided.
In the event that Task A-1 ldentifies potentially significant waterhammer scenarios which have not explicitly been accounted for in the design and operation of Grand Gulf, corrective measures will be required at that time. The task has not identified the need for measures beycnd those already imolemented.
Based on the foregoing, we conclude that Grand Gulf can be ocerated Orior to ultimate resolution of the A-1 generic issue without undue risk to the health and safety of the public.
A-9 Anticicated Transient Without Scram
^
Nuclear plants have safety and control systems to limit the consecuences of tercorary abnormal ocerating conditions or "anticioated transients."
Some deviations from normal acerating c:nditions may be minor; others, occurring less frecuently, may it:osa significant cemands en plant ecuicment. In some anticicated transients, rapidly shutting dcwn the nuclear reaction (initiating a " scram), and thus racidly reducing the generation of heat irt the reactor care, is an imoortant safety measure.
If there were a potentially severe ' anticipated transient" and the reactor shutccwn systems did not "scW as desired, then an " anticipated transient without scras, cr ATWS, wcuid have occur-ed.
Grand Gulf has been recuired to orovice a recirculation cumo trio in tne event of a reactor tric and to provide additicnal coerat:r raining for '
recovery frem anticicated transient witnout scram events. In acdition, nrand Gulf has imolemented energen.cy procacures and ocerator training to coce with potential anticipated transient without scram events.
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o Operator training and action.as described, in cenjunction with the automatic recirculation pumo trip, significantly improves the capability -
of the facility to withstand a range of anticipated transient without scram events, such that oceration of this facility presents no undue -
risk to the health and safety of the public while this matter is under review. Grand Gulf will have ATWS operator procedures and APT in place upon initial criticality.
The anticipated transient without scram issue is currently scheduled for rulemaking in sumer 1981. The applicant will be required *a comply with any further requirements on anticipated transient without scram which may be imposed as a result of the rulemaking. -
Based on our review, we conclude that there is reasonable assurance that Grand Gulf can be operated prior to ultimate resolution of this generic .
issue without endangering the health and safety of the public.
A-11 Reactor Vessel Materials Touchness Resistance to brittle fracture is described cuantitatively by a material .
croperty generally denoted as " fracture tougnness." Fracture toughness has different values and characteristics depending ucen the material .
being considered. For steels used in a nuclear reacter cressure vessel, three considerations are important. First, fracture toughness increases with increasing temoerature; second, fracture toughness decreases with increasing load rates; and third, fracture toughness decreases with neutron irradiation.
In recognition of these considerations, power reactors are operated within restrictions imposed by the Technical Scecifications on the pressure during heatuo and cooldcwn coerations. These restrictions .
assure that the reactor vessel will not be subjected to a comoination of cressure and temcerature that could cause brittle fracture of the vessel
'i f there were significant flaws in the vessel material. The effect of neut on radiation on the fracture toughness of the vessel material over the life of the plant is acccunted for in Technical Scecification limitations.
The principal objective of Task A-11 is to deveicp safety criteria to allcw a more crecise assessment of safety margins during nannal cceration, transiente and accident conditiens in older reactor vessels witn marginal fracture tcughness.
Sased on cur evaluation of this facility's reactor vessels materials tougnness, we have concluded that these units will have adequate safety margins against brittle failure during ocerating, test-ing, maintenance and anticiotted transient cenditions over the life of tne units. Since Task A-11 is crojected to Se comcleted well in advance of this facility's reacter vessel reaching a fluence level which would notacly reduce fracture resistance, accectable vessel intagrity for the postulatad accident conditions will be assured at least until tne reacter vessel is .
reevaluated for icng-tenn acceptability, as will be recuired as cur implementation requirement for Task A-11.
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In addition, the surveillance pregram required by 10 CFR 50, Apoendix H will afford an opportunity to reevaluate the fracture toughness periccically _
during the first half of design life.
Therefore, based ucon the foregoing, we have concluded that Grand Gulf can be operated prior to resolution of this generic issue without undue risk to the health and safety of the public..
A-17 Systems Interaction in Nuclear Power Plants Currently licensing requirements are fcunded on the principle of defense-in-depth. Adherence to this principle results in recuirements such as physical separation and indeoendence of redundant safety systems, and protection against hazards such as high energy line ruptures, missiles, -
high winds, flooding, seismic events, fires, human factors, and sabotage.
These design provisions are subject to review against the Standard Review Plan (NUREG-75/087) which requires interdisciolinary reviews and addresses many different types of potential systems interactions. The cuality assurance program which is followed during the design, construction, -
and operational phases for each plant is expected to provide added assurance against the potential for adverse systems interactions. Thus, -
the current licensing requirements and procedures provide for a degree of plant safety with respect to such interactions. ,
In November 1974, the Advisory Cc:nnittee on Reactor Safeguards requested that the NRC staff give attention to the need to increase safety by separately evaluating the plant frem a multidi~sciplinary point of view, in order to identify potentially undesirable interactions between plant systems. The concern arises because' the design, analysis and installation of systems is frecuently the responsibility of teams of engineers with ^
functional soecialties--such as civil, electrical, mechanical, or nuclear. Experience at operating plants led the ACRS to cuestion wnether the work of these functional specialists is sufficiently integrated to enable them to minimi:e adverse interactions among systems. Such adversa events have occurred because the teams did not assure by adecuate coordination that the. required independence of safety systems was providec under all l
conditions of coeration.
In mid-1977, Task A-17 was initiatad to assure that presens review procacures and safety critoria orovide an acceotaole lavel of redundancy and indecendence for safety functions. The task proceeded by evaluating the potential for undesiracle interactions between systems at a sacole.
plant.
The NRC staff's current procedures assign primary responsibility for t
review of various technical areas to soecific organizational units and assign secondary resconsibility to other units where there is a functional interface. Designers folicw somewnat similar procedures and provide the analyses of systams and interface reviews. Task A-17 orovided an indecencent .
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study of methods that could identify important systems intaractions tna; adversely imoact safety, and which were not considered by current review procedures. The first phase of this study began in May 1978 and was ccmoleted in February 1980 by Sandia Laboratories under contract to the
NRC staff. ,
The Phase I investigation was structured to identify areas where -inter-actions are cossible between systens and have the potential of negating or seriously degrading the performance of safety functions. The study concentrated on commonly caused or linked failures among systems that could violate a safety function. The investigatien was to then identify where NRC review procedures may not have properly accounted for these interactions.
The Sandia Laboratories used' fault-tree methods to identify comoonent failure combinations (cut-sets) that could result in loss of a safety function. The cut-sets were further reduc'ed by incorporating six common or linking systems failures inta. the analysis. The results of the Phase I effort indicate that, within the secpe of the study, only a few areas of the staff's review procedures need improvement regarding systems interaction. However, the level of detail needed to icentify -
all examples of cotantial system intaraction candidates observed in some ocerating olants were not within the Phase.I scope of the Sandia study.
The "NRC Action Plan Developed as a Result of the TMI-2 Accident,"
NUREG-0660, provides for a systems interaction follow-on study,Section II.C.3, " Systems Interactions." Since April 1980, tne Office of Nuclear Reactor Regulation has intensified the effort both by broadening the study of methods to identify potential systems interactions and by performing audit reviews of two plants for selected systems interactions.
Our recent excerience provides a basis from which we are develcping an improved systematic review process for potantial systems intaractions. '
The process will provide for a resolu:fon of USI A-17, assimilate ccerating reactor experience, and rank identified systems intaractions by their relative importance to safety.
In addition to the staff's interdisciclinary review, the Grand Gulf project acninistrative precedures (Project Procedures Manual and the Project Engineering Prcesoures Manual) provide the recuired guidance for interface cetween MPSL, GE, Sechtel and vendors.
In addition, the interfaca bet.veen BechteT , General Electric, and Mississicci
?cwer and Light is trackec by the Grand Gulf project control iog.
i To assure that all discipline interactions have identified all potantial ha:ards to safety related ecuitment, the Grand Gulf crojec has fomed the Engineering Review Team (E27). This team will review the as-cuilt candition of the plant for potential adversa effec:s :s safety related C-12 3
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- equipment. The team is made un of members of all disciplines and all -"
reports are c:ordinated with the resconsible disciplines.
The following safety issues are incluced in the review by the Grand un lf Engineering Review Team:
Non-Seismic Category I Cver Seismic Category I -
High Energy Line Break Flooding Jet Imoingement ,
Therefore, we conclude that there is reasonable assurance that Gran'd Gulf can be operated prior to the final resolution of this generic issue without endangering the health and safety of the public. .
A-39 Safetv/ Relief Valve Hydrodynamic toads All SWR plants are equipced with a numcer of SRVs to centrol prima ~ry system pressure transients. The SRVs are mountad en the main staam .
Ifnes inside the drywell with discharge lines routed througn the crywell into the suppression pool . When an SRV is actuated the steam released frem the primary system is discharged into the superessien pool wnere it is c:ndensed.
Actuation of an SRV can be either aut:matic, at a preset pressure, or manual by means of an external signal. A preselected numcer of SRVs are used for the Automatic Depressuritation System (ADS) wnich is designed to reduce the reactor cressure and permit operation of the icw pressure emergency core coolant systams. The ACS performs this function by aut matic actuation of the specified SRVs follcwing receict of specific -
signals frem the reactor protection system.
U:cn actuation of an SRV, the air column within the partially submerged s discharge line is ccmcressed by the high cressure steam and, in turn, acceleratas the water leg into the suooression pcol. The watar jets thus formed create cressure and velocity transients which are manifestac as drag or jet impingement loads on submerged structures.
Felicwing water ciaaring, tha c moressed air i: aisc 10:alerstad into the suoeression pcol forming high pressure air buboles. These bucoles executa a number of oscillatory ex:ansions and contractions before rising to the sucaressicn pcol surface. The associated transients again create drag loads on submerged structures as well as cressure loads en
- ne submerged bcundaries. These loads are referrec to as SRV air clearing loads. Containment structures, ecutienent and piping shall be designed j
to ac: mmodata these leads.
- n July 1976, the staff issued accactanca critaria fer SRV leads for :te Mark III containments. These critaria were estaolished en the basis of C-13 0
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our evaluation of the methcdology for predicting the SRV loads which was proposed by the General Electric Comoany. In late 1980, however, GE _.
proposed a revised method, wnich will result in substantial reduction of SRV loacs. This imoroved method was based on tne Caorso* inplant SRV __ ~
tests wnich were cerformed in January 1979 in Italy. In addition, Grand ,
Gulf has stated that they plan to perform in-plant confirmatory tests of _
their SRV ouencher discnarge. Grand Gulf has also used the revised SRY loads proposed by GE.
We are currently reviewing this new methodology for predicting the SRV loads. The results of our generic evaluation will be presented in a NUREG recort which is currently scheduled to be issued in the fourth ouarter of 1981. Our evaluation of the plant-specific acolication of this method for Grand Gulf will be reported in a Supplement to this SER. .
A 40 Seismic Desien Criteria - Short-Term procram .
NRC regulations require that nuclear power plant structures, systems and ccmoonents imcortant to safety 5e designed to withstand the effects of natural phenomena such as earthcuakes. Detailed requirements and guidance regarcing the seismic cesign of nuclear plants are provided in the NRC -
regulaticns and in regulatory guides issued by the Ccmmission. Mcwever, there are a number of plants with construction permits and ocerating licanses issued before the NRC's current regulations and regulatory guidance were in place. For this reason, rereviews of the seismic casign of various plants are being undertaken to assure that these plants do not present an undue risk to the public. . Task A-40 is, in effect, a c mpendium of short-term efferts to sucport such reevaluation efforts of the NRC staff, especially those related to older operating pl ants. In addition, some revisions to sections of the Standard Review Dlan and regulatcry guides to bring them more in line with the stata-of-the-art will result. ,
The saismic design Sasis.and seismic design of Grand Guif has been evaluated at the coerating license stage using curren: licensing critaria and recuirement. The staff's review of Grand Gulf to those critaria is discussed in Section of this Safety Evaluaticn Report. Should 1 the resolutien of Task A 40 indicate a change is needed in these licansing recuirements, all cperating reactors including Grand Gulf will be re-evaluatad on a case-cy-case basis. Accordingly, we have concluded that Grand Gulf can Se coeratad prior to ultimate resolution of this generic issue withcut andangering the health and caf2ty of the public.
A 43 Cen ainment ~:ercency Sume Reliacili v
~olicwing a postulated less-of-coolant accident, i.e., a break in the reactor c:alant systam pioing, the water ficwing frem the break would be collected in the suppression ccol. This water would be recirculated "Caorso 1s a f..R/? ark II plant locatad in Caorso, ?iancen:a in Italy. .
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througn the reactor system by the emergency core cooling pum:s to maintain core cooling. This wa:er may also be circulated thrcugn the -e containment scray system to remove heat and fission creducts from the drywell and wetwell atmosonere. Loss of the ability to draw water fr:m -
the suceression : col could disable the emergency c oling and containment soray systems.
One costulated means of losing the ability to draw water frem the sucaressice.
pool could be blockage by debris. A prir.cipal source of such debris could be the thermal insulation on the reactor coolant system oisino. In the event of a piping break, the subsecuent violent release to the hich cressure water in the reactor ecolant system could rio off the insulation in the area cf the break. This debris could then be carried over into the sucaression pool, cotantially causing blockage.
A second costulated means of losing the ability to draw water from the sucpressicn pool could be abnormal c nditiens at the cuma inle such as air entrainment or vortices. These conditions c:uld result in numn cavi-tation, reduced flew and cossible damace to the cumas. Due to the relatively Icw sucmergence for ECCS suction lines for Mark III containments (i.e. , .
4 ft. minimum sucmergence), the staff recuires that the acclicant cerform in-clant preccerational tests at minimum suction submergence for each of the ECCS systems to demonstrate that circulation through the pool can be readily ac: mplished without significant vortex formation. We will candition the coeratinc license for Grand Gulf that these tests be c:moleted by the fuel load date.
With regard to catential blockage of the intake Ifnes, the likelihood of any insulation being drawn into an emergency core c oling systam cuma suction line is very small. The potential debris in the drywell c uld only be sweet into the suporession pool via the hori: ental vents. Any pieces reaching the pool would tend to settle on the bottcm and would not be drawn into the cumo suction since the suction center line is 10.5 l
feet above the pool bottom. In aedition, boiling water reactor desiens emoloy strainers on the suction si:ed with flow areas 200t larger than the suction piping.
Accordingly, we cenclude that Grand Gulf can be operated prior to ultimate resolution of this generic issue withcut endangering the. health and safety of the public.
A la Statica Blackout Electrical pcwer for safety systems at nuclear power plants must be sucoliec by, at least, :No recundant anc indecendent divisions. The systens ased to rencve decay heat to c:ol the reactor core following a reactor shutdown are included among the safety systems that must meet these recuirements. Each electrical division for safety systems includes an offsite alternating current power connection, a standby emergency diasal generat:r altarnatinc current gewer sucaly, and direct current .
scurC3s.
Task 2 21 involves a study of whether or not nuclear power plants shculd be designed to ace;.a..cdate a ccmolete loss of all alternating current C-15 em e -q g oy,g ___ _
- cwer (i.e., a loss of both offsite and the emergency diesel generator -.
altarnating current cower sucolies). This issue arose because of coerating experience regarding the reliability of altarnating current ocwer sucolies. . . _
A numcer ot oceratina clants have excerienced a total less of offsite electrical ;cwer, anc mere occurrencas are excected in the futu-a Curing each of these ioss-of-offsita power events, tte ensita emergency alternating current :cwer sucolies were available to supoly the tower needed bv vital safety ecuipment. However, in seme instancas, one of .
the redundant energency power succlies has been unavailable. In addition, there have been numerous recorts of emergency diesel-generators failing to start and run in acerating clants during ceriodic surveillance tests.
A less of all altarnating current power was not a design basis event for the Grand Gulf facility. Nonetheless, a ccmbination of design, ocerating, and testing recuirements that have been imcosed on the acolicant -
will assure that these units will have substantial resistance to a loss of all alternating curren and that, even if a loss of all alternating current should cccur, there is reasonable assurance that the core will be ccoled. These are discussed below.
If offsita alternating current pcwer (three incecendent lines) is lost, -
three diesel-generators and their associated distribution systems will deliver emergency pcwer to safety-related equipment. Our review of the cesign, tasting, surveillance, and maintenance previsions for the ensite emergency diesels is described in Section of this SER. The recuirements include precoerational tasting to assure the reliability of the installed diesel-generators in accordance with our recuirements discussed in this recort.. In addition, Grand Gulf has imolemented a crogram for enhancament of diesel-generater reliability to bettar assure the icng-term reliacility of the diesel-generators.
~
If both offsite and onsite alternating current pcwer are icst, boiling water reactors may use a ccmtination of safety / relief valves and the reactor core isolation ecoling systam to remcve core decay heat witacut reliance en alternating current ccwer. These systems assure that adecuata cooling can be maintained for at least two hours, which alicws time for restcration of alternating current power frem eitner Offsita or ensite sources.
The issue of station blackout was considered by the Atcmi'c Safety and Licansing Apcaal Scard (ALAB-603) for the St. Lucie Unit Mc. 2 facility.
In adcitien, in view of the comoletion schedule for Task A 44 (Octocer 1982), the Aopeal Board recommended that the Commissicn take exceciticus action to ensure that other olants and their coerators are ecuipped cc acccmmcdata a station blackout event. The Ccmmission has reviewed this reccmmendatien and determined that some intarim measures jhculd be taken at all faciiities including Grand Gulf while Task A 4A is being concucted.
Consecuently, inter dm smergency procedures and coerator training for safe aceration of the #acility and restoration of alternating current cower will be recuired. The staff notified the acclicant of these recuirements in a letter ' rem D. Eisennut, MRC, to the acclicant dated .
We will condition the acerating license for Granc Gulf that :nese crocedures and tnis training Se comoleted by fuel lead data.
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Based en the ahove, we have c:ncluded that there is reasonacle assurance -
that Grand Gulf can ce oceratad prior to the ultimate resolution of this generic issue without encangering tne health and safety of the public. -,
A 45 Shutdewn Decav Meat Removal Recuirements Following a reactor shutdown, tne radioactive decay of fission prcducts continues to produce heat (decay heatl which must be removed frem the crimary system. _ The princical means for removing this heat in a boiling water reactor wnile at high pressure is via the steam lines to the turbine c:ndenser. The condensate is normally returned to the reactor vessel by the feedwater systam, however, the steam turbine-driven reactor core isolation c oling system is provided to maintain crimary system inventory, if alternating current power is not available. When the system is at low pressure, the decay heat is removed by the residual ,
heat removal systems. This " Unresolved Safety Issue" will evaluata the benefit of provicing alternate means of decay heat removal whien could substantially increase the plants' cacability to handle a broader scectrum of transients and accidents. The study will consist of a generic system evaluation and will result in recommencations regarding the desiracility ,
of and :ossible design recuirements for imcrevements in existing systams cc an alternative decay heat removal method if the imorov'ements or alternative can significantly reduce the overall risk to the public.
The Grand Gulf reactors have various metnods for the remcval of decay heat. As discussed above, the decay heat is normally rejected to the turbine condenser and returned to tne vessel by either the feecwater systam or the reactor core isolation cooling system (frem the condensata storage tank). If the condenser is not availacie (e.g. , loss of offsite ecwer), heat can be removed via the safety / relief valves to the sucpression pcol. Also, the high pressure core scray system is provided if the .
reactor care isolation cooling system is not available. Both of these systems can succly fluid to the vessel frem either the ccndensate storage tank or the suceression pcci. If the reactor care isolaticn c: cling and high pressure core soray are unavailabie, the reacter systam cressure can be reduced by the aut:matic decressuri:ation systam so that c:aling by the residual heat removal system can be initiated. When the condenser is not used, the heat rejected to the sucpression peal is subsequently removed by the residual heat removal systam.
The reactor core isciatica c: cling and hign pressure c:re scray systams it Grand Gulf have imorovements over c:mcaraole systems at older boiling watar reactors. The reactor core isolation c oling systen has teen L upgraded to safety-grade cuality (new recuired for ali boiling watar react:rsl, and the high pr?ssure c=re spray is powered by its own dedicated ciesel sc it can acerate with an assumed loss of all other sources of alternating current power. Also, the residual neat removal system centains three pumos; the flow capacey of any single ;ums (A or 3) is sufficient to easily remove the decay haat, d
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Folicwing the TMI accident, the industry per#crned and documented extensive analyses of feedwater transients and small-break loss-of-ccolant accidents "l to succort the acceotability of current designs. In addition, GE has. -
defined plant mecifications to increase the reliability of the decay -
heat removal systam, and is currently working to imolement those mdifications.
Based on the acove, we have concluded that Grand Gulf can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of the public.
A 46 Seismic Qualification of Ecuioment in Oceratino Plants The design criteria and methods for the seismic qualification of mechanical and electrical ecuipment in nuclear power plants have undergene significant .
change during the course of the comercial nuclear pcwer crogram.
Ccnsecuently, the margins of safety provided in existing ecuipment to resist seismically induced Icads and perform the intended safety functions may vary considerably. The seismic qualification of the ecuipment in operating plants must, therefore, be reassessed to ensure the acility to .
bring the clant to a safe shutdcwn condition when subject. to a seismic event. The objective of this " Unresolved Safety Issue d is to estaolisn an exclicit set of guidelines that could be used to judge the adecuacy of the seismic oualific:, tion of mechanical and electrical equipment at all operating plants in lieu of attemoting to backfit current design criteria for new plants. ??i guidance will concern equipment required to safely shut dcwn the plant, as well as' ecuipment whose function is not retuired for safe shutdown, :ut whose failure could result in adverse i conditions which might imeair shundcwn functicns.
. Grand Gulf was reviewed again:: current seismic criteria and accreved by -
the Ccmissicn staff in accorcanca with cur ent design criteria and l
methods for seismic qualification. The staff's review is discussed in Section of this Safew Evaluation Recert. Therefore, we conclude :nat Grand Gulf can be c: crated prior to resolution of this generic issue without undue risk to the health and safety of tne puclic.
A-47 Safety Imc11 cations of Centrol Systems This issue concerns the potential for transients er accicents being mace more severe as a result of cent ci systam "ailures er nifunctions.
These failures or malfunctions may occur indepercently or as a. result of the accident er transient under consideration. One concern is the potential for a single failure such as a loss of a power sucply, shcrt circuit, caen cir:uit, or sensor failure to cause simultaneous malfunction of several cen:rci features. Such an occurrence would conceivaoly result in a :nnsient more severs than those transients analy:ed as
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anticipated coerational occurrences. A sec:nd concern is for a postulated accident to cause control system failures which would make the accident __-
more severe than analy:ed. Accidents could conceivably cause control ,
system failurs by creating a harsn environment in the area of the _
control equipment or by pnysically damaging the control equipment.
Although it is generally believed that such control system failures would not lead to sericus events or result in conditions that safety systems cannot safely handle, in-depth studies have not been rigorously performed to verify this belief. The potential for an accident that would affect a particular control system, and effects of the c:ntrol system failures, may differ from plant to plant. .Therefore, it is not possible to develop generic answers to these concerns, but rather clant-soecific reviews are required. The purpose of this " Unresolved Safety Issue" is to define generic criteria that will be used for plant- -
specific reviews.
The Grand Gulf control and safety systems have been designed with the goal cf ensuring that centrol system failures (either single or multiole failures) will not prevent autcmatic or manual initiation-and operation -
of any safety system equipment recuired to trip the plant er to maintain the plant in a safe shutdown c:nditien folicwing any " anticipated Opera-ticnal occurrence" or "ac:ident." This has been acccmplished by either previding inde:endence between safety and non-safety systems or providing isolating devices between safety and ncn-safety systems. These devices preclude the propagation of non-safety system equipment faults such that operation of the safety system ecuipment.is not i=caired.
A widd range of bounding transients and accidents is presently analy:ed to assure that the postulated events would be adequately mitigated by the safety systams. In addition, systamatic reviews of safety systems - '
have 5een performed with the scal of ensuring that the control system -
failures (single er multiple) will not defeat safety system action.
Scecifically, these reviews have included:
(1) IE Bulletin 79-27 i A series of tables has been develeced wnich lists GGNS pcwer sources dcwn to the fuse level, to include alarm indications, instruments and centrol devices en these ;cwer scurces. Ccmpletion of the taoles with primary and secondary effects frem less of the pcwer scurces is in progress. Design modifications will be made as necessary wnen the determined effects have an adverse impact on plant safety.
(2) NRC Letter Catad Aoril 15, 1981, "Centrol System Failures" ,
i To address item (1) of this letter (identification of centrol l systems failures whicn could imtact plant safety), chencmena C-19
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- which could occur to initiate or worsen a transient / accident were determined. An exhaustive study was then made to cetermine all
control systems failures which could result in the phencmena. -
Identificarica of the cower panel, MCC, LCC, bus, transformer, battery and/or inverter, as applicable for each control syatem identified in item (1) was made. A rearrangement of this information showed control systems with cennon power sources and the effects of cascading cower losses.
A determination of control systems identified in item (1) that ,
receive input signals from common sensors was c mpleted.
An evaluation of the effects of the loss of a common sensor or power source on the analyses presented in FSAR Chapter 15 is now .
being conducted.
(3) NRC Letter Dated April 16,1981, "High Energy Line Breaks and Consecuential Control Systems Failures,' IE Notice 79-22 A matrix is being develoced which shows the effects,i ' f any, of high energy line breaks in control systems. If interacticn is discovered, the imcact of failure of the applicable system ucon the GGNS safety analyses will be evaluated.
A saecific subtask of this " Unresolved Safety Issue" will be to study the reactor overfill transient in boiling water reactors to determine the need for preventative and/or mitigating design measures to creclude or minimize the consecuences of this transient. Several early bofitng water reactors have excerienced reactor vessel overfill transients with subsecuent two-phase or liquid flow through the safety / relief valves. -
Following these early events, c:mmercial-grade high-level tries (level
- 3) have been installed at most boiling water reactors (including Grand Gulf) to terminate ficw from the appreoriate systems. These hign-level trips are single failure prcof and periodic surveillance is recuired by the Technical Scecifications. No overfilling events have occurred since the level 8 trics were installed.
Based on the above, we have concluded that there is reasonable assurance that Gr?.nd Gulf c:n be Operated prior to the ultimate resciutien of tnis generic issue without encancering the health and safety of tne cublic.
A 28 Hydrocen Control Measures and Effects of Hveracen Burns on Saferv Ecuicment co llowing a loss-of-coolant accident in a light water reactor plant, c:mbustible cases, principally hydrogen, may accumulate inside the primary reactor containment as a result of: (1) metal water reaction involving the fuel element cladding; (2) the radiolytic dec:mecsitien of the water in the reactor core and tne containment sumo; (3) -he corrosion C-20
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of certain construction meterials by the soray solution; and (2) any -~
synergistic chemical, therwal and radiolytic effects of post-accicent envicrnmental conditions on containment protective coating systems and --
electric cable insulation.
Becase of tne potential for significant hydrogen generation as the result of an accident,10 CFR 50.44, " Standards for Combustible 4as Control System in !.ight Water Cooled Power Reactors," and Criterion 41 of the General Design Criteria, " Containment Atmosphere Cleanup," in Aopendix A to 10 CFR Part 50, requires that systems be provided to control hydrogen concentrations in the containment atmosonere following a postulated accident to ensure that centainment integrity is maintained.
. The regulation,10 CFR Section 50.44, requires that the combustible gas control system provided be capable of handling the hydrogen generated as a result of degradation of the emergency core ccoling system such that .
the hydrogen release is five times the amount calculated in demonstrating compliance with 10 CFR Section 50.46 cr the amcunt corresponding to reaction of the cladding to a depth of 0.00023 inch, whichever amount is ~
greater.
The accident at TMI-2 en March 23, 1979 resulted in hycrogen generation well in excess of the amounts specified in 10 CFR Secticn 50.a4 As a result of this knowledge it became apparent to NRC that speci'fic cesign measures are needed for handling larger hycrogen releases, particularly for smaller, icw-pressure c:ntainments. As a result, the Ccmmission determined that a rulemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need to be taken into act unt in plant design. An advance notice of this rulemaking preceeding en degraded core issues was published in the Federal Recister en Oct:ber 2, 1980.
Rec:gni:ing that a nunber of years may be recuired to comolete this ruiemaking proceeding, a set of short-tarm or interim actions relative to hydrogen control requirenents was develeced and i=clemented. These interim measures were described in a sec:nd October 2, 1980 Federal Recister notice.
For plants with Mark III c:ntainments such as Grand Gulf, the preposed interim rule specified that either it must be demonstrated .: bat the c:ntainment can withstand hycregen burns or exolosions or a detailed evaluation of~ possible. hydrogen c:n rci measures must be performed anc the selected measures installed.
Grand Gulf was recuested to c moly with these interim measures prior to fuel lead. In subnittals made to :ne NRC on April 9 anc June 19, 1981, the acclicant's evaluation cf aiternate hydrogen control measures was cravided. A Hydmgen Ignition Systam (HIS) was selectec and detailed evaluations of centainment pressure and tarcarature rescense were cerformed.
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The HIS consists of gicw plug igniters distributed throughout the centainment and drywell. The HIS is designed to ignite hycrogen at low --
concentrations, thereby maintaining the concentration of hydrogen below its detonable limit and preventing c:ntainment overpressure failure. --
Containment response to the burning of hyoregen has been analyted using the CLASIX-3 c moutar code develcped by Offshore- Power Systams. An -
analysis of the ability of essantial equipment to survive the hydrogen burn environment is undenvay- the anticicatec comoletion data is Cecamcer 1981.- The MIS will be installed and fully operable by the Decemoer 31, 1981 Unit i fuel load data.
Significant additional work is underway to demonstrate that the containment pressure and temperature resconse calculations are adequata, that potential detonations do not constitute a threat to safety, and that essential ecuicment will survive hydrogen burns resulting from operation of the HIS. -
In addition, Mark III owners have formed an cwners group to evaluata hydrogen control measures for Mark II! centainments, and the apolicant is actively involved in the ongoing evaluations of that cwners group.
The staff has reviewed and aporoved (1) the Grand Gulf Hydrogen Ignition System, and (2) the applicant's analysis of the ability of essential equicment to survive the hydrogen burn environment. This evaluation is provided in Sections and of this Safety Svaluation Report.
Based on the above, we cenclude that Grand Gulf can be operated prior to resolution of the " Unresolved Safety Issue dand the precosed rulemaking without undue risk to the health and safety of the public.
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' June 3,'1932 REG- 12. 75 8_. ;
File No. G9.5 Mr. A. Schsenc r, C'hief Licensing Branch No. 2 Division of Licensing Office of Nuclear Reactor Regulations -
U. S. Nuclear Regulatory Ccemission
- Washington, D. C. 20555
Dear Mr. Schwencer:
s River Bend Station Uni:s 1 & 2
, Docket Nos. 50-453 & 50-459 4
5 This letter is provided in response to your November 19, 1981, request for additional informati:n regarding the status of unresolved '
safety issues at Gulf States Utilities (GSU). GSU participated in the Licensing Revieu Group II-(LRG-II) activities pertaining to the
- generic resolution of unresolved safety issues. Position paper 1-GI3 was included in the submittal by the LRG-II via a letter dated January 25, 1982 to Howard J. Faulknuer (NRC) from D. L. Holt:cher, Chairman of the LRG-II Working Group. GSU intends to formall'/ endorse this position on our docket in the near future, together with the other applicable LRG-II positions. A copy of 1-GIS is attached for your convenience. ~
Sincerely,
/>
h ' $Wh J. E. Ecoher Manager-Engineering & Licensing River Bend Nuclear Group
,.4 M.
J3/L.C/k:
Attachment
_ - - . _ _ . _ - _ _ - . . . ~ , _ _- . , _ - _ - . . _ _ _ , , __ _ . _ _ _ _ . _ _ . - - ,
1/25/82 1-GIB - -s.w 4.fJrff # l 0 $ - - -
'~.
INTERIM LICENSING BASES PENDING RESOLUTION 0F UNRESOLVED SAFETY ISSUES ISSUE LRG-II plants will develop unified bases and justification for licensing and operation while the identified generic safety issues remain unresolved and provide a summary description of relevant investigative programs and interim measures pending resolution of the unresolved safety issues.
LRG-II RESPONSE LRG-II participants have reviewed the generic issues identified in NUREG-0606, " Unresolved Safety Issues." The following information is
, provided for each of the applicable " Unresolved Safety Issues" as a bases for licensing prior to ultimate resolution of these issues.
=~
Nuclear nocument C IO U ?! b Q, MJA: rf: csc/118A7
1/25/82 o .
1-GIB (Page 2)
A-1 Waterhammer Waterhammer events a're intense pressure pulses in fluid systems caused by any one of a number of mechanisms and system conditions such as rapid i condensation of steam pockets, steam-driven slugs of water, pump startup with partially empty lines, and rapid valve motion. Since 1971, over 200 incidents involving waterhammers in pressurized and boiling water reactors have been reported. The waterhammers (or steam hammers) have involved steam generator feedrings and piping, and residual heat removal systems, emergency core cooling systems, and containment spray, service water, feedwater and steam lines.
Most of the damage reported has been relatively minor, involving pipe .
hangers and restraints; however, several waterhammer incidents have resulted in piping and valve damage. The most serious waterhammer events have occurred in the steam generator feedrings of pressurized water reactors. In no case has any waterhammer incident resulted in the release of radioactive material.
Under Generic Task A-1, the potential for waterhammer in various systems is being evaluated and appropriate requirements and systematic review procedures are being developed to ensure that waterhammer is given appropriate consideration in all areas of licensing review. A technical report, NUREG-0582, "Waterhammer in Nuclear Power Plants" (July, 1979),
providing the results of an NRC staff review of waterhammer events in nuclear power plants and stating current staff licensing positions, completes a major subtask of Generic Task A-1.
i Although waterhammer can occur in any light water reactor and over 100
, actual and probable events have been reported in boiling water reactors, none have caused major pipe failures in a boiling water reactor such as the LRG II plants and none have resulted in the offsite release of radioactivity. As noted above, the most severe waterhammers observed to date have been in steam generators. Since the boiling water reactor does not utilize a steam generator, these worst cases are eliminated.
Furthermore, any waterhammer which may occur in feedwater or main steam piping will not impair the emergency core cooling system since all ECCS water enters the reactor vessel via five separate reictor vessel nozzles indeoendent of the feedwater and main steam piping.
In order to protect the LRG-II plants emergency core cooling systems against the effects of waterhammer, the ECC systems are provided with jockey pumps. These jockey pumps keep the emergency core cooling system lines water-filled so that the emergency core cooling system pumps will not start pumping into voided lines and steam will not collect in the emergency core cooling system piping. To ensure that the emergency core cooling system lines remain water-filled, vents have been installed and further assurance fcr filled discharge piping is provided by pressure instrumentation at the piping high point. An alarm souncs in the main control room if the pressure falls below a predetermined setcoint l
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A-1 Waterhammer (Cont'dl indicating difficulty maintaining a filled discharge line. Shoulc t.lis occur, or if an instrument becomes inoperable, the required action is
.- identified in the Technical Specifications.
With regard to additional protection against potential waterhammer events currently provided in plants, piping design codes-requ' ire consideration of impact loads. Approaches used at the design stage include: (1) increasing valve closure times, (2) piping layout to preclude water slugs in steam lines and vapor formation in water lines, (3) use of snubbers and pipe hangers, and (4) use of vents and drains, In addition, LRG-II' participants will conduct a preoperational vibratio'n and cynamic effects test program in accordance with Standard OM-3 of the American Society of Mechanical Engineers for all Class 1, Class 2, Class 3 and other piping systems and piping restraints during startuo and initial operation. These tests will provide adequate assurance that the piping restraints have been designed to withstand dynamic effects due to valve closures, pump trips, and other operating modes.
Nonetheless, in the unlikely event that a large pipe break did result from a severe waterhammer event, core cooling is assured by the emergency core cooling system and protection against the dynamic effects of such pipe breaks inside and outside of containment is provided.
In the event that Task A-1 identifies potentially significant water-hammer scenarios which haved not explicitly been acccunted for in'the
, design and operation of LRG-II plants, corrective measures will'be implemented at that time. The task has not identified the need for
~
i measures beyond those already implemented.
Based.on the foregoing, we conclude that the LRG-II plants can be operated prior to ultimate resolution of this generic issue without undue risk to the health and safety of the public.
A-9 Anticioated Transients Without Scram
, Nuclear plants have safety and control systems to limit the consequences of temporary abnormal operating conditions or " anticipated transients."
Some deviations from normal operating conditions may be minor; others, occurring less frequently, may impose significant demands on plant equipment. In some anticipated transients, rapidly shutting down the nuclear reaction (initiating a " scram"), and thus rapidly reducing the generation of heat in the reactor core, is an important safety measure.
, If there were a potentially severe " anticipated transient" and the reactor shutdown system did not " scram" as desired, then an " anticipated transient without scram," or ATWS, would have occurred.
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A.-9 Anticicated Transients Without Scram (Cont'd)
All boiling water reactors, including LRG-II plants, have been required to provide recirculation pump trip in the event of a reactor trip and to provide additional operator training for recovery frcm anticipated transients without scram events. In addition, LRG-II plants will implement emergency procedures and operator training to cope with potential anticipated transients without scram events.
Operator training and action as described, in conjunction with the autcmatic recirculation pump trip, significantly improves the capability of the facility to withstand a range of anticipated transient without scram events, such, that operation of this facility presents no undue risk to the health and safety of the public while this matter is under review.
The anticipated transient without scram issue is currently under review through the rulemaking proceedings. Notice of the proposed rule for ATWS was puolished in the Federal Register on Novemoer 24, 1981. LRG-II plants will comply with any further recuirements on anticipated transient without scram which may be imposed as a result of the rulemaking.
Based on our review, we conclude that there is reasonable assurance that the LRG-II plants can be operated prior to ultimate resolution of this generic issue without endangering the health and safety of the public.
A.-11 Re&ctor Vessel Materials Touchness Resistance te brittle fracture is described quantitatively by a material
, property generally denoted as " fracture toughness." Fracture toughness has different values and characteristics depending upon the material being considered. For steels used in a nuclear reactor pressure vessel, three considerations are important. First, fracture toughness increases with increasing temperature; second, fracture toughness decreases with increasing load rates; and third, fracture toughness decreases with neutron irradiation.
Inrecognitionoftheseconsiderations,powerreacthrsareoperated within restrictions imposed by the Tecnnical Specifications on -he pressure during heatup and cooldown operations. These restrictions assure that the reactor vessel will not be subjected to a combination of pressure and temperature that could cause brittle fracture of the vessel if there were significant flaws in the vessel material. The effect of neutron radiation on the fracture toughness of the vessel material over the life of the plant is accounted for in Technical Specification limitations.
The principal objective of Task A-11 is to develop safety criteria to allow a more precise assessment of safety margins during normal operation,
, transients and accident conditions in older reactor vessels with marginal fracture tougnness.
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A.-11 Reactor Vessel Materials Touchness (Cont'd)
Based upon evaluation of the LRG-II reactor vessel's materials, toughness, we conclude that adequate safety margins exist for brittle failure during operating, testing, maintenance, and anticipated transient conditions over the life of the units. Since Task Action Plan A-11 is projected to be completed well in advance of LRG-II plants reactor vessels reaching a fluence level which would notably reduce fracture resistance, acceptable vessel integrity for the postulated accident conditicas will be assured. When Task Action Plan A-11 is completed and explicit fracture evaluation criteria for accident conditions are defined, all vessels will be reevaluated for accept-ability over their, design lives. .
The materials of the LRG-II reactor vessels meet the fracture toughness requirements of NB-2300 of the ASME Code. Based on this fact and the fabrication techniques employed on the vessel, we estimate that the total fluence over the design life would result in a final fracture toughness value above the minimum charpy impact requirement of 50 foot pounds. In addition, the surveillance program required by Appendix H of 10CFR Part 50 will afford an cpportunity to reevaluate the fracture toughness periodically during a minimum of the first half of the design life.
To assure adequate safety margins, adjustment to the nil ductility transient temperature (NDTT) and the development method for pressure /
temperature curves are specified in 10CFR50 Appendices G and H. The amount of adjustment to the operating curves is a function of reference temperature, RT which depenas upon the fast neutron (1 Mev) fluence and copper and h,bbsphorus content in the RPV material. For BWR/6's, th:-
copper and phospnorus content of the material is closely controlled.
Furthermore, hign upper shelf toughness is specified and all values for core belt line material were in excess of 75 ft-lbs. The fast neutron fluence is low with respect to other reactor types because of the additional moderator (water) in the annulus between the core shroud and the RPV. Therefore, the reactor pressure vessel material toughness (A-11) issue is of relatively low concern for BWR/6's.
Therefore. based ucon the foregoing, we conclude that LRG-II plants can be operated prior to resolution of this generic issue without undue risk to the health and safety of the public.
A-17 Systems Interaction in Nuclear Power Plants The licensing requirements and procedures used in the LRG-II plant safety reviews address many different types of systems interaction.
Current licensing requirements are founded on the defense-in-depth principle. Adherence to this principle results in requirements such as physical separacion and independence of redundant safety systems, and protection against events such as high energy line ruptures, missiles, high winds, flooding, seismic events, fires, operator errors, and sabotage.
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A-17 Systems Interaction in Nuclear Power Plants (Cont'd)
These design provisions supplemented by the current review procedures of the Standard Review Plan, which require interdisciplinary reviews and which acccunt, to a large extent, for review of potential I systems interactions, provide for an adequately safe situation with respect-to such interactions. The quality assurance program which is followed during the design, construction, and operational phases for each plant is expected to provide addea assurance against the potential for adverse systems interactions.
In. mid-1977, Task A-17 was initiated to confirm that present review procedures and safety criteria provide an acceptable level of redundancy and independence for systems required for safety by evaluating the potential for undesirable interactions between and among systems.
The NRC staff's current review procedures assign primary responsibility for review of various technical areas and safety systems to specific organizational units and assign secondary responsibility to other units where there is a functional or interdisciolinary relationship. Designers follow somewhat similar procedures and provide for interdisciplinary reviews and analyses of systems. Task A-17 provide an indeoendent study of methods that could identify important systems interactions adversely impacting safety, and which are not considered by current review procedures.
The fir:' phase of this study began in May, 1978, and was completed in February, 1330, by Sandia Laboratories under contract to the NRC staff.
The Phase I /estigation was structured to identify areas where interactio~ are possible between and among systems and have the
, potential for negating or seriously degrading the performance of safety
' functions. Tie study concentrated on common cause of linking failures among systems that could violate a safety function. The investigation then identifi2d where NRC review procedures may not have properly accounted for tnese interactions.
The Sandia Study used fault-tree methods to identify component failure combinations (cut-sets) that could result in loss of .a safety function.
The cut-sets were reduced to minimal ccmcinations by' incorporating six ccmmon or linking systams failurss into the analysis. Tha results of the Phase I effort indicate that, within the scope of the study, only a few areas of review procedures need improvement regarding systems interaction. However, the level of detail needed to identify all examples of potential system interaction candidates observed in some operating plants are not within the Phase I scope of the Sandia Stucy.
The Systems Interaction Branch, formed in the Office of Nuclear Reactor Regulation in April,1980, has been studying state-of-the-art methods that can be used to predict systems interactions. The initial ef fort, supported by three laboratory contracts, is underway; a range of methocs is being considered and tested for feasibility againt a sample of some systems interaction candidates derivea from Licensee Event Report evaluations.
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A-17 Svstems Interaction in Nuclear Power Plants (Cont'd) .
1 It is expected that the development of systematic ways to identify and evaluate systems interactions will reduce the likelihood of common cause failures resulting in the loss of plant safety functions. However, the studies to date indicate that current review procedures and criteria supplemented by the application of post-TMI findings and risk studies provide reasonable assurance that the effects of potential systems interaction on plant safety will be within the effects on plant safety previously evaluated.
LRG-II participants will provide for a systematic visual inspection by a multidisciplinary team to review the "as-built" condition of the plant areas where physical interactions could potentially result in adverse effects on safety grade equipment. Visual inspections of interaction areas are performed for spatially coupled systems interactions initiated by seismic events. Any spatial separations that do not meet established design criteria are reported for disposition by analysis and/or hardware modification. LRG-II participants are improving their programs basec on the experience gained in the industry's efforts, but will maintain the multidisciplinary team concept which the staff considers essential to a systems interaction analysis.
Therefore, we conclude that these is reasonable assurance that the LRG-II plants can be operated prior to final resolution of this generic issue without endangering the health and safety of the public.
A-39 Safetv Relief Valve Hvdredvnamic Loads
. All BWR plants are equipped with a number of SRVs to control primary system pressure transients. The SRVs are mounted on the main steam lines inside the drywell with discharge lines routed through the drywell into the suppression pool. When an SRV is actuated, the steam released frem the primary system is discharged into the suppression pool where it is condensed.
Actuation of an SRV can be either automatic, at a preset pressure, or manual by means of an external signal. A preselected number of SRVs are used for the Automatic Decressurization System (ADS) which is designed to reduce the reacter pressure and permit operation of the low pressure emergency core cooling systems. The ADS performs this function by automatic actuation of the specified SRVs following receipt of specific signals from the reactor protection system.
Upon actuation of an SRV, the air column within the partially submerged discharge line is compressed by the high pressure steam and, in turn, accelerates the water leg into the sucoression pool. The water jets thus formed create cressure and velocity transients which are manifested as drag or jet impingement loads on submerged structures.
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A-39 Safety Relief Valve Hvdrodynamic loads (Cont'd)
Following water clearing, the coepressed air is also accelerated into the suppression pool, forming high pressure air bubbles. These bubbles execute a number of oscillatory expansions and contractions while rising to the suppression pool surface. The associated transients again create drag loads on submerged structures as well as pressure loads on the submerged boundaries. These loads are referred to as SRV air clearing loads. Containment structures, equipment and piping at LRG-II plants have been designed to accommodate these loads.
In July, 1976, the staff issued acceptance criteria for SRV loads for the Mark III containments. These criteria were established on the basis of our evaluation of the methodology for predicting the SRV loads which was proposed by the General Electric Company. In late 1980, however, GE proposed a revised method, which will result in substantial reduction of SRV loads. This improved method was based.on the Caorso inplant SRV tests which were performed in January, 1979, in Italy. NRC has approved the revised GE method in NUREG-0802. The LRG-II plants have used the revised SRV loads accepted by the NRC and will review the inplant testing results from Kuo Sheng 1 and Grand Gulf-1, to determine their applicability and confirm the conservatism of their designs. This concern is addressed in further detail in LRG-II issue 1-CSB.
A-40 Seismic Desian Criteria - Short-Term Prooram NRC regulations required that nuclear pcwer plant structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes. Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in i the NRC regulations and in regulatory guides issued by the Commission.
However, these are a number of plants with construction permits and operating licenses issued before the NRC's current regulations and regulatory guidance were in place. For this reason, re-reviews of the seismic design of various plants are being undertaken to assure that these plants do not present an undue risk to the public. Task A-40 is, in effect, a compendium of short-term efforts to support such reevaluation efforts of the NRC staff, especially those related to older operating plants. In addition, some revisions to sections of the Standard Review Plan and regulatory guidas to bring than more in line with the state-of-the art will result.
The seismic design basis and seismic design of the LRG-II plants have been established on current licensing criteria and requirements. Should the resolution of Task A-40 indicate a cnange is needed in these licensing requirements, all operating reactors including Clinton, Perry and River Bend will be reevaluated on a case by case basis.
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A-40 Seismic Desion Criteria - Short-Term Procram (Cont'dl Accordingly, we have concluded that the LRG-II plants can be operated prior to ultimate resolution of this generic issue without endangering the health and safety of the public.
A-43 Containment Emercency Sumo Reliability Following a postulated loss-of-coolant accident, i.e. , a break in the reactor coolant system piping, the water flowing from the break would be collected in the suppression pool. This water would be recirculated through the reactor system by the emergency core cooling pumps to maintain core cooling. Loss of the ability to draw water from the -
suppression pool could disable the emergency core cooling system.
The concern addressed by this Task Action Plan for boiling water reactors is primarily focused on the potential for degraded emergency core cooling system performance as a result of thermal insulation debris that may be blown into the suppression pool during a loss-of coolant accident and cause blockage of the pumo suction lines. A second concern, potential vortex formation, is not considered a serious concern for Mark III containment due to the large depth of the pool and the low approach velocities.
LRG-II plants have a minimum suction submergence for the ECCS systems of over 7 feet. This concern is addressed in further detail in LRG-II issue 7-RSB.
With regard to potential blockage of the intake lines, the likelihood of any insulation being drawn into an emergency core cooling system pump suction line is very small. The potential debr s in the drywell could i
I only be swept into the suppression pool via the horizontal vents. Any pieces reacning the pool would tend to settle on the bottom and would not be drawn into the pump suction since the suction center line is minimum of 4 feet and the approach velocity is only 1 foot /second above the pool bottom.
In addition, boiling water reactor designs-employ strainers on the suction sized with flow areas 200 percent larger than the suction piping.
Accordinglyc we conclude that LRG-II' plants can be. operated prior to ultimate resolution of this generic issue without endangering the health and safaty cf tha public.
A-44 Station Blackout Electrical power for safety systems at nuclear power plants must be supplied by at least two redundant ano independent divisions. The systems used to remove decay heat to cool the reactor core following a reactor shutdown are included among the safety systems that must meet these recuirements. Eacn electrical division for safety systems includes two offsite alternating current power connections, a stancey emergency diesel generator alternating current power supply, and direct current sources.
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A-44 Station Blackout 3' '
Task.A-44 involves a study of whether or not nuclear power plants should be designed to accommodate a complete loss of all alternating current f power, i.e. , a loss of both the offsite and the emergency diesel generator
{ alternating current power supplies. . .This issue arose because of operating experienc'e regarding the reliability of alternating current power supplies.
A number of operating plants have experienced a total loss of offsite electrical power, and.more occurrences are expected in the future.
During each of these loss of-offsite power event, the onsite emergency alternating current power supplies were available to supply the power needed by vital safety _ equipment. However, in some instances, one of the. redundant emergency power supplies have been unavailable. In addition, there have been numerous reports of emergency diesel generators failing to start and run in operating plants during periodic surveillance tests.
A loss of .all alternating current power was not a design basis event for the LRG-II facilities. Noneth'eless, a combination of design, operating, and testing requirements have been imposed to assure that these units will have substantial resistance to a loss of all alternating current and that, even if a loss of all alternating current should occur, there is reasonable assurance the core will be cooled. These design, operating, and testing requirements are discussed below.
A loss of offsite alternating current power involves a loss.of both the 1
preferred and backup sources of offsite power. Our review and basis for acceptance of the design, inspection, and testing provisions for the offsite power system are described in Section 8.2 of the SER.
If offsite alternating current-power is lost, three diesel generators i and their associated distribution systems will deliver emergency power to safety related equipment. Our review of the design, testing, sur-
, veillance, and maintenance provisi.ons for the onsite energency diesels is described in Sections 8.3 and 9.5 of the SER. The requirements include preoperational testing to assure the reliability of the installed diesel generators. .
If both offsite and onsite alternating crr ent power are lost, boiling water reactors may.use a combination of safety / relief valves and the reactor core isolation cooling system to remove core decay heat without reliance on alternating current power. These systems assure tnat adequate cooling can be maintained for at least two hours, which allows time for restoration of alternating current power from either offsite or onsite Jources.
The issue of station blackout was considered by the Atomic Safety and Licensing Appeal Board (ALAB-603) for the St. Lucie No. 2 facility. In addition, in view of the ccmoletion schedule for Task A-44 (Octcber, 1982), the Appeal Board recommended tnat the Ccmmission take expeditious action to acccmmodate a station blackout event. The Commission has reviewed their recommendations and determined that some interim measures should be taken at all facilities including LRG-II plants while Task A-aa MJA: rf:csc/113A16
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A-44 Station Blackout (Cont'd) is being conducted. NRC Generic Letter 81-04 requested a review of plant capability to mitigate a station blackout event and prompt implementation, as necessary, of emergency procedures and a training program for station blackout events. Consequently, interim emergency procedures and operator training for safe operation of the facility and restoration of alternating current power will be implemented. This action will be completed by fuel load date.
Based on the above, we have concluded that there is reasonable assurance that LRG-II plants can bo operated prior to the ultimate resolution of this generic issue:without enoangering the health and safety of the public.
A-45 Shutdown Decav Heat Removal Recuirements Following a reactor shutdown, the radioactive decay of fission creducts continues to product heat (decay heat) which must be removed from the primary system. The principal means for removing this heat in a boiling water reactor while at high pressure is via the steam lines to the turbine condenser. The condensate is normally returned to the reactor vessel by the feedwater system; however, the steam turoine-driven reactor core isolation cooling system is provided to maintain primary system inventory, if alternating current power is not available. When the system is at low pressure, the decay heat is removed by the residual heat removal systems. This " Unresolved Safety Issue" will evaluate the benefit of providing alternate means of decay heat removal which could substantially increase the plants' capability to handle a broader
,, spectrum of transients and accidents. The study will consist of a generic system evaluation and will result in recommendations regarding the desirability of and possible design requirements from improvements in existing systems or an alternative decay heat removal method if the improvements or alternative can significantly reduce the overall risk to the public.
The LRG-II plants:are designed with various methods for the removal of decay heat. 'As discussed above, the decay heat is normally rejected to the turbine condenser and condensate is returned to the vessel by the
- feedwater system. The reactor core isolation cooling (RCIC) system i
provides an alternate means of supplying makeup water to the vessel.
This turbine driven pump takes suction from the condensate storage tank and pumps to the vessel. If the condenser is not available (e.g., loss of offsite power), heat can be removed via the safety / relief valves to the suppression pool. Also, the high pressure core spray (HPCS) system I
is provided if the reactor core isolation cooling system is not available.
Both of these systems (RCIC and HPCS) can supply water to the vessel from either the concensata storage tank or the suporession pccl.
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A-45 Shutdown Decav Heat Removal Recuirements (Cont'd)
If the reactor core isolation' cooling and high pressure core spray are unavailable, the reactor system pressure can be reduced by the autcmatic
- depressurization system so that cooling by the residual heat removal systec can be initiated. When the condenser is not used, the heat rejected to the suppression pool is subsequently removed by the residual heat removal system.
The reactor core isolation cooling and high pressure core spray systems for the LRG-II plants have improvements over ccmparable systems at older boiling water reactors. The reactor core isolation cooling system has been upgraded to safety grade quality (now required for all boiling ,
water reactors), and the high pressure core spray is powered by its own dedicatea diesel 50 it can operate with an assumed loss of all other sources of alternating current power. Also, the residual heat removal system contains three pumps; the flow capacity of any single pumo (A or B) is sufficient to remove the decay heat.
Following the TMI accident, the industry performed and documented extensive analyses of feedwater transients and small-break loss-of-coolant accidents to support acceptability of current designs. A report of these analyses was provided to the NRC in NED0-24708A Revision 1, dated i
December, 1980.
Based on the above, we have concluded that the LRG-II plants.can be operated prior to the ultimate resolution of the generic issue without endangering the health and safety of the public.
A-46 Seismic Qualification of Ecuicment in Oceratinc Plants The design criteria and methods for the seitmic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant change during the course of the commercial nuclear power program. Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. TN.c -csismic qualification of the equipment in operating plants must, therefore,.be reassessed to ensure the ability to bring the plant to a safe shutdown condition when subject to a seismic event. The ocjective of this "ljnresolved Safety Issue" is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants in lieu of attemoting to backfit current design criteria for new plants. This guidance will concern equipment required to safely shutdown the plant, as well as equipment wnose functions is not recuired for safe snutdown, but whose failure could result in adverse conditions which might impair shutdown functions.
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A-46 Seismic Oualification of Eouicment in'Oceratino Plants (Cont'd)
LRG-II plants were designed using current seismic design criteria, anti methods for seismic equipment qualification are to be latest codes and
/ .stand ard s. Requirements for seismic equipment qualification include !
IEEE 344-1975 and Regulatory Guides-1.92 and 1.100. ~ Standard Review Plans 3.2.2, 3.9.2, 3.9.3, and.3.10 have also been considered in the qualification efforts.
Since identification of hydrodynamic loa'd effects on LRG-II plant structures,.an effort was initiated to assess'the effects of these loads (in combination with previously established seismic loads) on equipment required to safely. shut down the plant. This reassessment involved validation of equipment qualification througn both analytical methods and additional testing, where required.
It is concluded that LRG-II plants can be operated prior to resolution of this generic issue without undue risk to the health and safety of the public.
A-47 Safety Imolications of Control Systems This issue concerns the potential for transients or accidents being made more severe as a result of control system failures or malfunctions.
These failures or malfunctions may occur independently or as a result of the accident or transient under consideration. One concern is the potential for a single failure such as a loss of power supply, short circuit, open circuit, or sensor failure to cause simultaneous malfunction
', of several control features. Such an occurrence would conceivably
. , result in a transient more sever than those transients analyzed as anticipated operational occures. A second concern is for a postulated accident to cause control system failures which would make the 4ccident more severe than analyzed. Accidents could conceivably cause control system failures by creating a harsh environment in the area of the control equipment or by physically damaging the control equipment.
Although it is generally believed that such control system failures would not. lead to serious events or result in conditions that safety systems cannot safely handle, in-deoth studies have not been' rigorously performed to verify this belief. The cotential for an accident that would affect a particular control system, anc effects of the control system failures, may differ from plant to plant. Therefore, it is not possible to develop generic answers to these concerns, but rather plant-specific reviews are 4 required. The purpose of this " Unresolved Safety Issue" is to define generic criteria that will be used for plant-specific reviews.
The LRG-II plant control and safety systems have been designed with the goal of ensuring that control system failures (either single or multiple failures) will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe shutdown condition following any " anticipated operational occurrence" or "accicent". This has been accomolished by either providing MJA: rf:csc/118A19
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A-47 Safety Imolications of Control Svstems (Cont'd) independence between safety and nonsafety systems or providing isolating devices between safety and nonsafety systems. These devices preclude e the propagation of.nonsafety system equipment faults such that operation of the safety system equipment is not impaired.
A systematic evaluation of the control system design, such as contemplated for this " Unresolved Safety Issue," has not been performed to determine whether postulated accidents could cause.significant control system failures which would make the accident consequences more severe than
~
presently analyzed. However, a wide range of bounding transients and accidents is presently analyzed.to assure that the postulated events would be adequatel mitigated by the safety systems.' In addition, systematic reviews of safety systems have been performed with the goal of ensuring that control system failures (single or multiple) will not defeat system action. Specifically, these reviews include identification and evaluation of the potential adverse impacts to plant safety as a result of control system failures, effects from loss of non-Class 1E i
power sources, and harsh environments following high energy line breaks.
These concerns are addressed in further detail in LRG-II issues 5-ICSB, 6-ICSB, and 7-ICSB.
A specific subtask of this " Unresolved Safety Issue" will be to study the reactor overfill transient in boiling water reactors to determine
[ the need for preventative and/or mitigating design measures to preclude e or minimize the consequences of this transient. Several early boiling j water reactors have experienced reactor vessel overfill transients with 4
subsequent two phase or liquid flow through the safety / relief valves.
Following these early events, commercial grade high-level trips (Level 8)
I have been installed at LRG-II plants to terminate flow from the appropriate systems. These high-level trips are single failure proof and periodic surveillance is required by the Technical Specifications. No overfilling events have occurred since the Level 8 trips were installed. In addition BWR/6's have a high level scram that precludes this concern.
l Based on the above, we have concluded that there is reasonable assurance
, that LRG-II plants can be operated prior to the ultimate resolution of this generic issue without endangering the health and safety of the public.
4 l A-48 Hydrocen Control Measures and Effects of Hydrocen Burns on Safetv Ecuiceent i Following a loss-of coolant accident in a light water reactor plant, combustible gases, principally hydrogen, may accumulate inside the primary reacter containment as a result of: (1) metal-water reaction involving the fuel element cladding; (2) the raciolytic deccmposition of 1
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- ' A-48 Hydrocen Control Measures and Effects of Hvdrecen Burns on Safety Eouioment (Cont
- c) of .certain construction materials by the spray solution; and (4) any f synergistic chemical, thermal, and radiolytic effects of post accidents environmental conditions on containment protective coating systems and electric cable insulation.
l Because of the potential for significant hydrogen generation as the result of an accident, 10CFR Section 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," and Criterion 41 of the General Design Criteria, " Containment Atmosphere Cleanup," in Appendix A to 10CFR Part 50, require that systems be provided to control hydrogen concentrations in the containment atmospnere following a postulated accident to ensure that containment integrity is maintainec.
Regulation 10CFR Section 50.44 requires that the combustible gas control system provided ce capaole of handling the hydrogen generated as a result of degradation of the emergency core cooling system such that the hydrogen release is five times the amount calculated in demonstrating compliance with 10CFR Section 50.46 or the amount corresponding to
- reaction of the cladding to a depth of 0.00023 inch, whichever amount is greater.
The accident at TMI-2 on March 28, 1979, resulted in hydrogen generation well -in excess of the amounts specified in 10CFR Section 50.44. As a result of this knowledge, it became apparent to NRC that specific design measures are needed for handling larger hydrogen releases, particularly for smail, low pressure containments. As a result, the Commission determined that a rulemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need to be taken into account in plant design. An advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Reoister on October 2,1980. -
Recogni::ing that a number of years may be required to complete this rulemaking proceeding, a set of short-term or interim actions relative to hydrogen control requirements was developed and iinglemented. These interim measures were described in a seccnd October 2, 1990, Federal Register notica.
For plants with Mark III containments, the proposed interim rule specified that either it must be demonstrated that the containment can withstand hydrogen burns or explosions or a detailed evaluation of possible hydrogen control measures must be performed and the selected measures installed.
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A-48 Hydrocen Control Measures and Effects of Hvdrocen Burns on Safety Eouiccent (Cont'd)
The LRG-II position is to comply with this interim rule through use of a hydrogen igniter system. This system consists of glow plug igniters distributed throughout the containment. This system is designed to ignite hydrogen at low concentrations, thereby maintaining the con-centration of hydrogen below its detonable limit and preventing potential containment overpressure.
To collectively evaluate the concerns associated with the Hydrogen issue f.or Mark III containments, LRG-II participants are involved in an owners group. This group is sponsoring analytical work with General Electric, Offshore Power systems and others. Current evaluations of this group '
have demonstrated that containment pressures will remain well below the failure point as the result of the anticipated hydrogen release and burn.
Based on the above, we conclude that LRG-II plants can be operated prior to resolution of the " Unresolved Safety Issue" and the proposed rulemaking without undue risk to the health and safety'of the public.
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