ML20128Q238
| ML20128Q238 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/24/1985 |
| From: | Johnson G NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#385-952 OL, NUDOCS 8507260624 | |
| Download: ML20128Q238 (51) | |
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July 24, 1985 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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CLEVELAND ELECTRIC ILLUMINATING DocketNo.50-440kb,{({D COMPANY, ET AL.
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50-441 OL
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(Perry Nuclear Power Plant,
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- e5 JUL 25 P4.42 Units 1 and 2)
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TfGY5$f oHANCH NRC RESPONSE IN OPPOSITION TO OCRE MOTION TO RE0 PEN THE RECORD AND ADMIT CONTENTION ON PARTIAL EXEMPTION TO APPENDIX J I.
_ INTRODUCTION By its Motion to Reopen the Record and to Submit a New Contention
(" Motion") dated July 5, 1985, Intervenor Ohio Citizens for Responsible Energy (OCRE) seeks to reopen to record in this proceeding for the purpose of admitting the following contention:
Applicants have not demonstrated, pursuant to 10 CFR 2.758, that the application of Section III.D.2(b)(ii) of Appendix J to 10 CFR 50 to the Perry facility does not serve the purposes for which that regulation was adopted.
Motion, at 2.
OCRE's Motion is in response to the Applicants' request for partial exemptien, pursuant to 10 C.F.R. 50.12, from the requirement in Para-graph III.D.2(b)(ii) of Appendix J to 10 C.F.R. Part 50 which states that:
Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such periods at not less than P.
3 8507260624 850724 PDR ADOCK 05000440 G
PDR 1)Sf
k As part of the exemption request, Applicants proposed to substitute the seal leak test of Appendix J, Paragraph III.D.2(b)(iii), for the test otherwise required by Paragraph III.D.2(b)(ii). Although OCRE claims that the requested exemption has been granted, the Staff has only indi-cated an intent to consider and evaluate the exemption request in an SER Supplement and to document the exemption, it granted, in the Perry Unit 1 operating license.
See Letter, dated June 21, 1985, from B. J.
Youngblood to M. R. Edelman, transmitting Environmental Assessment and Finding of No Significant Impact (Motion, Exh. 3).1/
OCRE's newly proffered contention claims a failure to meet the requirements of 10 C.F.R. b 2.758. OCRE asserts that it is this regulation, rather than 10 C.F.R. 5 50.12, which applies to Applicants' request; never-theless, OCRE addresses the Section 50.12 exemption standards in arguing that the requested exemption ought not be granted. OCRE claims that the Commission has no authority to grant exemptions, that even if it does have such authority, the subject request cannot meet Section 60,12 requirements, and that financial burden is not permitted to form a basis for granting the subject exemption. Motion, at 2-3.
OCRE further asserts that the proposed substitution of the seal leak test specified in Para-graph III.D.2(b)(iii) of Appendix J is inadequate to assure the facility is "as safe as if it were in compliance" with the full pressure test requirement of Paragraph III.D.2(b)(ii), citing Long Island Lighting Co.
-1/
As noted by OCRE, the Staff has undertaken an environmental assessment of the proposed exemption and published in the Federal Register a finding of "No Significant Impact" as required by 10 C.F.R. 5 51.35. These actions did not constitute a grant of the exemption.
9
\\ (Shoreham Nuclear Power Station), CLI-84-8, 19 NRC 1154 (1984). 2/ OCRE, in reference to the Section 50.12 criteria, argues that the substitution of the seal leak test "would clearly endanger life and property by increasing the probability of containment leakage during an accident" and thus "is obviously not in the public interest." Motion at 6.
More directly pertinent to its contention, OCRE argues that Appli-cants, as "a party to a proceeding" must seek waiver of the Appendix J regulatory requirement pursuant to 10 C.F.R. 6 2.758, showing that applica-tion of the regulation in this case would not serve the purposes for which the rule was adopted and that special circumstances of the case justify such waiver or exception. Motion at 7, 8.
Although no such waiver has been sought in this case, OCRE argues that there is nothing unusual about the request for exemption which would constitute a showing of "special circumstances" under 6 2.758. 3/
Intervenor acknowledges that it is required to meet the standards for reopening the record, and for admitting late-filed contentions, and 2/
The Commission has determined to limit the applicability of CLI-84-8 to the Shoreham case, and to consider changes to the current exemp-tion process in separate rulemaking. Memorandum for W. J. Dircks, H. H. E. Plaine from S. J. Chilk, dated July 27,1984(reporting posed rule to amend 10 CTR 6 5U.T2)_eg. 16506 (April 26, 1986) (pro-vote of Commission). See, 50 Fed. R See also, Mississippi Power &
Light Co. (Grand Gulf Nuclear Station, Un1I 17, CLI-84-19, 20 NRC 1055, 1059 n.7 (1984); Philadelphia Electric Company (Limerick Generating Station, Units I and 2), ALAB-809 (June 17,1985) (slip op.at12n.10).
-3/
Intervenors also argue that exemptions are granted only in extraordinary cases upcn a showing of exceptional circumstances, citing Shoreham, supra, CLI-84-8, 19 NRC 1154, and that, under that decision, applications for exemptions in a contested case must be submitted to the Licensing Board. Motion, at 8.
k makes brief arguments asserting that the pertinent requirements are satisfied. Motion at 8-11.
Specifically, OCRE claims its motion meets the reopening standards because it was filed within 10 days of the Statt's Notice of Environmental Assessment and Finding of No Significant Impact, because it addresses the significant issue of " containment integrity," and because it alleges non-compliance with the regulations which would preclude making the requisite safety findings under 10 C.F.R. 5 50.57 (presumably leading to a different licensin'g decision). Motion, at 9.
Similarly, OCRE argues, pursuant to 10 C.F.R. 6 2.714(a)(i)-(v), 4/
that (1) the contention could not have been filed earlier, (2) relief (and thus representation of OCRE's interest) by the Staff is unlikely, (3) OCRE will be able to contribute to a sound record based on its
" considerable research on the subject of containment performance and integrity" relating to Issue #8, (4) no other party can represent OCRE's interest in the matter, and (5) while some delay might result, any delay should be attributed to the Applicants. Motion, at 10-11.
4_/
The factors considered in determining whether to admit late-filed contentions are:
(1) Good cause, if any, for failure to file on time.
(ii) The availability of other means whereby the petitioner's interest will be protected.
(iii)The extent to which the petitioner's participation may reasonably be expected to assist in developing a sound record.
(iv) The cxtent to which the petitioner's interest will be represented by existing parties.
(v)
The extent ot which the petitioner's participation will broaden the issues or delay the proceeding.
10 C.F.R. 9 2.714(a)(1).
k As discussed below, OCRE's arguments seeking to justify reopening the record and to admit its late-filed contention are non-specific and unpersuasive, and fall far short of meeting the applicable standards.
While it is therefore unnecessary to reach the question whether 0CRE's contention is otherwise admissible, the contention is, in any event, inadmissible, since it seeks to impose requirements on Applicants which are not found in Commission regulations.
10 C.F.R. 5 2.758. Similarly, to the extent OCRE's contention might be read as challenging the Commission's exemption authority or the grounds which may form the basis for issuance of an exemption under Section 50.12, the contention is also a challenge to the regulations, as interpreted by the Coninission.
Finally, Intervenor has failed to provide the requisite basis for a contention.
For all these reasons, the Staff opposes the OCRE motion.
II. DISCUSSION A.
OCRE Has Failed to Satisfy the Requirements for Reopening the Record It is well-settled under Commission case law that before a motion to reopen a closed record will be granted the motion must (1) be timely presented (except in the case of an exceptionally grave issue, which may be entertained even if untimely presented), (2) address a s1gnificant safety or environmental issue, and (3) establish that a ditterent result would have been reached initially had the newly proffered material been considered. See, e.g., Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units I and 2), ALAB-598, 11 NRC 876, 879, 883 (1980); jd., ALAB-775,19 NRC 1361,1365-67, n.18 (1984), af f'd, sub nom.
< San Luis Obispo Mothers for Peace v. NRC, 751 F.2d 1287 (D.C. Cir. 1984),
vacated in part and rehearing on banc granted on other grounds, 760 F.2d i
1320(1985). See also, Metropolitan Edison Co. (Three Mile Island Nuclear Station, Unit No. 1), CL1-85-07, 21 NRC _ (May 9, 1985) (slip op at 3); Proposed Rule:
Criteria for Reopening [Recordsj in Fomal Licensing Proceedings, 49 Fed. Reg. 50189 (December 27,1984)(proposing to codify the case law standard and to provide specific particularity requirements for affidavits).
For a motion to reopen to be timely presented, the movant must show that the issue sought to be raised cculd not have been raised earlier. Diablo Canyon, supra, ALAB-775, 19 NRC at 1366. On the matter of significance, mere allegations or broad assertions that the regulations have been violated, without specific information which "would permit realistic appraisal of the safety significance," fail to establish the safety significance of the issue, notwithstanding the statement of serious charges. Louisiana Power &
Light Company (Waterford Steam Electric Station, Unit 3), ALAB-/86, 20 NRC 1087, 1090-91 (1984). Neither bare allegations nor simple submission of new contentions will suffice. Diablo Canyon, supra, ALAB-775, 19 NRC at 1366. The material submitted in support of the motion must be in excess of that required to support an admissible contention.
Id.
In order to show that the riew information would materfally affect the decision, the movant must submit material which would constitute " relevant, material, and reliable" evidence under 10 C.F.R. 5 2.743(c).
Id. at 1366-67, n.18.
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6 Against these heavy burdens _/ OCRE comes up short.
First, OCRE is 5
silent as to why it waited until July 5, 1985, two months after the May 3, 1985 closing of the hearing record, before raising its contention.
OCRE's May 8, 1985 letter to the Staff concerning the April 8, 1985 exemption application shows that OCRE was aware of the pertinent details of the exemption request for at least two months prior to submitting its contention. Motion. Exh. 2.
Inasmuch as OCRE's underlying concern is whether the plant will be safe should the exemption be granted (see Motion, at 9), it need not have waited for Commission action on the request before raising its concerns. See also, discussion infra, at 9-10.
Although a two-month delay in filing the motion ordinarily might not be fatal, at this late stage of the Perry proceeding it is not insignificant and it has not been justified.
OCRE's Motion is not timely.
Second, although OCRE notes that containment integrity is vital to protection of the public health and safety, and that non-compliance with Commission requirements for finding the plant to be safe might prevent licensing, OCRE draws no cogent link between such statements and whether the air lock seal leak test may adequately substitute for a full pressure test when no maintenance on the air lock has been performed. Contrary to OCRE's argument, Motion, at 4-5, ' hat the 3-day seal test may not sub-stitute for the full pressure test prescribed in Appendix J, Para-5/
See, Waterford, su ra, ALAB-786, 20 NRC at 1090, citing Kansas Gas and Electric Co.
o f Creek Generating Station, Unit No.1),
ALAB-462, 7 NRC 320, 328 (1978).
L.
t graphIII.D.2.(b)(ii),thelanguageinParagraphIII.D.2(b)(111) prohibits only the substitution of the 3-day seal test for the 6-month test pre-scribed in Paragraph III.D.P.Ib)(1), implying that substitution of the seal leak test for the full pressure test in Paragraph III.D.2(b)(ii) is not prohibited. OCRE's statement that air lock doors have failed to close or latch does not support its argument and does not address whether the seal leak test is a safe and acceptable substitute.
See, Motion at 5.
That substitution of the seal leak test for the full pressure test for the limited purposes of Paragraph III.D.2(b)(ii) of Appendix J does not raise a significant safety issue is attested to by the numerous facilities which have received t'le very same exemption. Exemption from the requirement of Paragraph III.D.2(b)(ii) has been granted for the Catawba, McGuire, Grand Gulf, Limerick, Byron, Fermi, and Wolf Creek facilities. See Attachment.
In short, OCRE has presented nothing which establishes that Applicants' specific request for exemption is of such safety significance as to warrant reopening the record.
Finally, OCRE has offered no "significant new evidence," as the Appeal Board has defined it in ALAB-775. OCRE has submitted neither affidavits nor anything which might be characterized as " relevant, material, and relloble" evidence. As noted abovc, the reference to NUREG/CP-0056 concerning failure of air lock doors to close or latch does not raise a significant issue regarding whether the door seals will leak when closed, or whether seal leak tests will serve the purpose of full pressure tests at the close of periods when containment integrity is not required. OCRE thus fails to show how the information presented is I
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relevant or material to containment integrity at Perry, and offers no information regarding the Perry plant which may be considered reliable evidence.
Thus, OCRE has failed to show that this new information or issue might change the overall result in the proceeding. SI In sum, OCRE's motion to reopen is not timely, does not show that Applicants' request for exemption raises a matter of safety significance, and does not demonstrate that were the matter considered, a different overall result might be reached, thus, OCRE has not met any of the standards for reopening the record and its motion to reopen should be denied.
B.
The Five-Factor Test for Late-Filed Contentions Balances Against Admission of the OCRE Contention In moving to reopen the record for litigation of a new contention, OCRE must not only demonstrate that the standards for reopening are met but it must show that the criteria for admission of late-filed conten-tions balance in its favor.
Pacific Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 & 2), CLI-82-39, 16 NRC 1712, 1714-15 (1982). Those criteria are set forth in 10 CFR 6 2.714. OCRE's showing 4
l on each of the late-filing factors is addressed below.
Factor i.
It is conceded that OCRE could not have raised a conten-tion on the safety of granting an exemption to Paragraph III.D.2(b)(ii) i s/
OCRE's argument Motion at 8-9. that the standard relating to whether a different result would have been reached does not apply where an initial decision has not yet issued, was implicitly rejected in Public Service Company of Oklahoma, et al. (Black Fox Station, Units 1and2),ALAB-573,10NRC775,BUTIT975).
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4 of Appendix J prior to filing of the application therefor on April 8, 1985. As noted in the previous section, however, OCRE sent a letter to the Staff, dated May 8, 1985, commenting on the exemption application.
It is therefore clear that 0CRE had enough information available to it no later than May 8, 1985 with which to file the subject contention. OCRE argues that it did not file its contention until after the Federal Register Notice of the Environmental Assessment and Finding of No Signi-t1 cant Impact because "it was not [until] then known whether the Staff would grant the exemption." Motion, at 9.
Similar arguments (i.e., that intervenor had to await availability of Staff positions before being able to file an adequate contention), however, have been rejected by the Commission.
AstheCommissionnotedinDukePowerCompany,etal.(Catawba Nuclear Station, Units 1 and 2), CLI-83-19, 17 NRC 1041, 1049 (1983), it is the applicant who carries the burden of proof on safety issues, and the possibility that a Staf f Safety Evaluation Report may lead to modificatio1 of an applicant's safety analysis in support of its license application "does not provide a reasonable basis for deferring of safety-related contentions until the Staff issues its SER." M. While, under Catawba, supra, CLI-83-19, 17 NRC at 1047, good cause may be found for filing a late contention where the late-filed contention is based solely on information contained in licensing-related documents which had previously been institutionally unavailable, late contentions must be tiled promptly after information sufficient to formulate an adequate contention becomes available. See Catawba, supra, ALAB-687, 16 NRC 460, 469(1982).
In this case, the document on which a contention could have
% been formulated came into existence on April 8,1985 and was accessible for public examination shortly after April 15, 1985. OCRE's May 8, 1985 letter demonstrates that it had sufficient information to file its contentien at least two months prior to its actual submission. While g000 cause may exist for filing OCRE's contention promptly after April 15, 1985, good cause has not been shown for the subsequent several-month delay in filing, and therefore the good cause factor (10 C.F.R. 92.714(a)(1)(1))weighssomewhatagainstadmission.
Factors 11 and iv.
It does not appear that there is another party or another adjudicatory forum through which OCRE can have its interest in the exemption represented, and therefore the second and fourth late-filing factors (10 C.F.R. 65 ?.714(a)(1)(ii), (iv)) appear to favor admission. Cf. Washington Public Power Supply System et al. (WPPSS Nuclear Project No. 3), 18 NRC 1167, 1173-// (1983). However, these factors ordinarily are accorded "relatively minor importance" in the balancing of the five factors. The Detroit Edison Company, et al.
(Enrico Fermi Atomic Power Plant, Unit 2), ALAB-707, 16 NRC 1760, 1767 (1982).
Factor 111. With respect to whether OCRE can be expected to assist in developing a sound record (10 C.F.R. 5 2.714(a)(1)(iii)), Intervenor has made a very weak showing, giving no indication that it has access to experts, or that Intervenor itself is technically qualified or prepared to address the pertinent issue -- i.e., whether safety is adversely l
attected by the substitution of a seal leak test for a full pressure test at the conclusion of pertods when containment integrity is not required.
l Cf. WPPSS, supra, ALAB-747,18 NRC at 1177.
It is the ability to i
y p-
,ym y-y
,g-
e y
4 contribute sound evidence, rather than asserted legal ur litigation skills that is of significance to this late-tiling factor. Houston Lighting & Power Co. (Allens Creek Nuclear Generating Station, Unit 1),
ALAB-671, 15 NRC 508, 513 n.14 (1982). OCRE's preparation for other issues involving containment integrity appears to be largely irrelevant.
OCRE has shown little in the way of its ability to contribute to the development of a sound record on the exemption issue, and this factor weighs against admission of OCRE's proffered contention.
Factor v.
Finally, OCRE concedes that admission at this late date of a new contention would delay the proceeding. Motion, at 10. Since the adequacy of air lock tests was not a subject of a previous contention or hearings, litigation of this matter would also broaden the proceeding.
It should be noted that the question before the Licensing Board on factor tive is not whether licensing will be delayed, E but whether the proceeding will delayed.
10 C.F.R. 5 2.714(a)(1)(v).
See, WPPSS, supra, ALAB-/47,18 NRC at 1178-1180. Admission of the proffered contention will both broaden the issues and delay completion of the proceeding.
Ihus, factor five weighs against admission.
~/
In point of fact, the Staff does not believe that licensing would 7
have to be delayed for litigation of OCRE's late-filed contention.
Were the Licensing Board to determine that the standards for re-opening and late-tiled contention admission were met, and the con-tention were deemed otherwise admissible, the matter would no longer be before the Statt for determination. However, in such circumstances, a decision on the exemption need not hold up l
licensing. A hearing on the exemption could be held after the Licensing Board issues its initial decision on issues already liti-gated.
If that initial decision were favorable to licensing, the operating license for Perry Unit 1 could be issued without the requested exemption (and with technical specifications imposing Appendix J Paragraph III.D.2(b)(ii) test requirements) and the matter of whether the exemption ought to be granted could be liti-gated subsequently.
% In sum, the three heavily weighted late-filing factors weigh against admission of the contention, while only the two lightly weighted factors weigh in favor.
In these circumstances, a balancing of the factors under 10 C.F.R. Q 2.714(a)(1) should lead to rejection of the proffered contention.
C.
OCRE's Contention Is Not Litigable Because It Challenges Commission Regulations and Lacks Basis and Specificity 1.
Limiting Applicants to Invoking Section 2.758 is Inappropriate The legal premise underlying OCRL's contention is that, in order to be relieved of the requirements of paragraph III.D.2(b)(ii) of Appendix J to part 50, Applicants must demonstrate under 10 C.F.R. b 2./58 that application of the subject provision should be waived because it does not serve the purposes for which that regulation was adopted. OCRE's attempt to place such a requirement upon Applicants would, contrary to 10 C.F.R.
Q 2.758, impose obligations not found in Commission regulations and preclude pursuit of other avenues of relief authorized by the regulations.
Section 2.758(b) provides that a party to a proceeding "may petition that application of a specified Commission rule or regulation... be waived or an exception made for a particular proceeding." (Emphasis added.) Section 2.758(b) is, on its face, permissive, applies only to application of requirements in initial licensing proceedings, and in no way precludes a party to a proceeding from seeking an exemption pursuant a
=-
% toSection50.12(a).8/ Indeed, by its terms, Section 2.758 focuses more on the mechanism by which a party may challenge or attack an NRC regula-tion in an adjudication (10 C.F.R. il 2.758(a), (b)) than on the manner in which a party may be relieved from compliance with a regulation which it admits is valid and applicable. Moreover, inasmuch as what Applicants seek is an exemption from a regulatory requirement for the life of the Perry plant, Section 2.768 does not provide the relief they seek.
In addition, Applicants do not contend that the regulation would not serve its purpose, but, rather, that 11teral compliance would entail unnecessary burdens on the facility without concomitant safety benefits. El Motion, Exh. 1.
Finally, since the subject matter of the exemption request is not a matter in controversy in this proceeding, E it would not have been appro-8/
Both the Commission and the Appeal Board have tacitly approved the approach of seeking a 10 C.F.R. I 50.12 exemption rather than a 10 C.F.R. 5 2.758 waiver in similar circumstances. Shoreham, supra, CLI-84-8, 19 NRC at 1155; Limerick, supra, ALAB-809, at 6 el seq.
J 9/
Relief from a regulatory requirement pursuant to 10 C.F.R. 5 2./b8 is appropriate only "when an interpretation or application of a regulation to particular tacts is called into question, which is not the case in the instant request." Washington Public Power Supply System (WPPSS Nuclear Project Nos. 3 and 5), CLI-77-11, 5 NRC 719, 723(1977).
10/ A licensing board has before it only those matters put in controversy by the parties.
10 C.F.R. 5 2.760a and Section VIII, Appendix A to Part 2.
Under i 2.760a, a licensing board may raise issues on its own initiative, but only upon a determination that "a serious safety, environmental, or comon defense and security matter exists." However, as discussed in Section II.A. of the text, supra, there has been no such showing, nor is there any reason to believe that the issuance of an exemption from Paragraph III.D.2.(b)(ii) of Appendix J to Part 50 raises a substantial safety matter.
i
V I
' l l
priate to submit the exemption request to the Licensing Board. E Thus, i
l the relief provided by Section 2.758 is not particularly pertinent here and there is no basis for requiring or limiting Applicants to resort to I
that provision.
In sum, since OCRE's contention seeks to require Applicants to i
petition for a Section 2.758 waiver where such waiver is not required, and to limit reliet to that available under Section 2.758 where such relief is neither appropriate nor adequate, and where other regulatory l
relief is permitted and available, the contention itself appears to constitute a challenge to Commission regulations (both Sections 2.758(b) and50.12(a))inviolationofSection2./b8(a),andshouldberejected.
l 2.
Intervenors May Not Challenge Section 50.12 or Commission Interpretations Thereof in this Proceeding Although OCRE's contention does not itself rai:e the question whether Applicants' exemption request satisfies the requirements of 10 C.F.R. I 50.12 (presumably because OCRE believes it does not apply),
OCRE argues at some length that granting of an exemption to Appendix J is unauthorized, ultra vires, and, in any event, not consistent with the findingsrequiredtobemadeunderSection50.12(a). Motion, at 2-6.
Since it is clear from Commission regulations that the Commission i
is, indeed, authorized to issue exemptions under Section 50.12 when the 11/ This case is distinguishable from Shoreham, supra, CLI-84-8, 19 NRC at 1155, where a request for an exemption from General Design Criterion 17 arose directly from consideration of an admitted con-tention challenging the reliability of Shoreham's diesel generators.
TIT-Tl4~, WPPSS Project Nos. 3 and 5, supra, CLI-77-11, 5 NRC at See also l
L..
i
{
. appropriate findings are made, to the extent OCRE's Motion can be read to i
raise a contention that issuance of exemptions is not authorized by law, such contention is also a challenge to Connission regulations, contrary to 10 C.F.R. Q 2.758. E Second, OCRE argues that the Commission may not consider the finan-cial burden of literal compliance with a regulation when determining l
whether to grant an exemption under Section 50.12. Motian at 3.
Here l
l again, however, the prohibition OCRE seeks to read into the Section 50.12(a) goes well beyond the regulation. Moreover, in the recently published notice of proposed amendment to 10 C.F.R. 6 50.12, which the Appeal Board has noted provides helpful insight into the application and purpose of the existing rule, E the Commission rationale for amending the regulation relies substantially on Federal case law authority which holds that administrative agencies have as part of their authority to proceed "'by means of rules of general application a concomitant authority to provide exemptions procedures in order to allow for special circumstances.'" Proposed Rule, Specific Exemption, 50 Fed. Reg.16506 (April 26,1985), citing, U.S. v. Allegheny-Ludlum Steel Corp., 406 U.S.
l l
12/ OCRE's interpretation of " authorized by law" as requiring specific statutory authority to authorize a particular exemption is clearly
~
at variance with the Commission's interpretation of that language.
I See, 50 Fed. Je[. 16506 (April 26, 1985).
("As in the existing R
rule, an exempt'on must be authorized by law. Apart from the very l
fact of granting the exemption relief itself, the granting of the exemption cannot be in violation of other applicable laws, such as the Atomic Energy Act or the National Environmental Policy Act.")
13/ Philadelphia Electric Company (Limerick Generating Station, Units 1 and2),ALAB-809(June 17,1985)(slipop,at7n.5).
1
I t 742,755(1972). The authority to issue exemptions is based on just such considerations as inequity or financial hardship that agency rules or i
regulations may cause in particular circumstances. M.
Intervenor has simply misinterpreted the requirements which must be satisfied in granting an exemption under Section 50.12 as being the only factors which the Comission may consider.
For example, in the consideration of an Exemption in the Shoreham proceeding, which Intervenor references, the Comission specifically determined that " exigent circumstances" considered should include financial or economic hardship.
I_d.; Long j
Island Lighting Company (Shoreham Nuclear Power Station, Unit 1 (CLI-84-8,19NRC1154,1156,n.3(1984). Although the Comission's directives in Shoreham have been limited to the Shoreham proceeding, EI they do indicate the Comission's agreement that financial or economic burdens may form the basis of need for an exemption. OCRE's attempt to preclude consideration of financial burden is clearly inconsistent with the purpose of exemptions, generally, and Comission interpretation of current requirements. EI To the extent OCRE's Motion is read to raise a l
l l
M / See n.2 supra.
M/ Since the authority to grant exemptions from rules of general l
applicability based on equitable considerations peculiar to a particular regulated person is not based on the underlying purpose l
of the agency regulation, but upon fairness, the fact that the underlying purpose of Comission regulations is the public health j
and safety and comon defense and security does not itself preclude considering economic or financial hardship in determining whether an i
exemption is needed. On the other hand, such equitable consider-l ations may not be balanced against the health and safety and comon defense and security. Limerick, supra, ALAB-809, at 11-12, citing (FOOTNOTE CONTINUED ON NEXT PAGE) l l
. contention that consideration of financial burden is unauthorized, such I
contention also const1tutes a challenge to 10 C.F.R. 9 50.12, as interpreted by the Commission, contrary to 10 C.F.R. 9 2.758.
Finally, OCRE argues that granting Applicants' exemption request would endanger life and property by increasing the probability of containment leakage during an accident. Motion, at 4-6.
Again, this matter is not raised in OCRE's contention, but appears to be in support of OCRE's position that since no exemption may properly be granted under 10 CFR 9 50.12, the Licensing Board must now apply 10 C.F.R. 9 2.758.
Motion, at 6.
Even assuming that compliance with Section 50.12 were the subject of OCRE's contention, it would, in any event fail, for lack of the required basis with reasonable specificity. OCRE's Motion references several provisions of the Perry FSAR, but aside from these, it is devoid of l
reference to the Perry facility and the safety of granting an Appendix J exemption to Applicants. Surely, citation to the Proceedings on the Second Workshop on Containment Integrity (Motion, at 5) fails to provide the requisite nexus to a safety concern at Perry. OCRE does not state l
(F00lNOTECONTINUEDFROMPREVIOUSPAGE)
Shoreham, supra, CLI-84-8,19 NRC at 1156, n.3.
However, the fact that the Commission may not consider the financial investment of a utility in construction of a facility in determining whether an application for an operating license can be granted does not pre-i l
clude granting exemptions where the requisite safety findings have been made. Cf. Power Reactor Development Co. v. International l
Union of Elect'rical, Radio and Machine Workers, AFL-CIO, 367 U.S.
396, 415 (1961); Pacific Gas and Electric Company v. State Energy Resources Conversation & Development commission, 461 U.S. 190, 207 l
(1983).
I
any basis for questioning the safety of granting the subject exemption tor the Perry facility. As a result, even if the Motion were treated as raising a contention on the sufficiency of. the exemption under.
Section 50.12, it must be rejected for failing to meet the basiv and specificity requirement in 10 C.F.R. 9 2.714(b).
In sum, OCRE has not presented a contention which meets Commission requirements for admissibility, and its proffered contention should be rejected.
I III. CONCLUSION OCRE has failed to meet Commission requirements applicable to reopening a closed record, admitting late-filed contentions, and submission of contentions with basis and reasonable specificity. As a result, OCRL's Motion and its contention should be denied.
Respectfully submitted, x$--
George E. Johnson Counsel for NRC Staff Dated at Bethesda, Maryland, this 24th day of July,1985.
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ATTACHMENT UNITED STATES
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,E W ASHINGTON, D. C.,20S$5 NUCLEAR REGULATORY COMMISSION
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January 17, 1985 Docket No. 50-413 Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street
(,harlotte, North Carolina 28242 ~
Dear Mr. Tucker:
Subject:
Issuance of Facility Operating License No. NPF Catawba Nuclear Station, Unit 1 The NRC has issued the enclosed Facility Operating License NPF-35 together with Technical Specifications and Environmental Protection Plan for the Catawba Nuclear Station, Unit 1.
The license authorizes operation at 100 percent power (3411megawattsthermal).
A safet'y evaluation supporting changes to license. conditions 2.C.(6) and 2.C.(22) and deletion of license ' conditions 2.C.(20) and 2.C.(25) contained
._f in Facility Operating License NPF-31 is enclosed. This safety evaluation
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will b.e incorporated in Supplement No. 5 to the Ca,tawba Safety Evaluation Report.
f Alto enclosed is a copy of a related notice, the original of which has been forwarded to the Office of the Federal Register for publication.
Five signed copies of Amendment No. 3 to Indemnity Agreement No. B-100 which covers the activities authorized under License No. NPF-35 are enclosed.
Please sign all copies and return one copy to this office.
Sincerely, w
ilen u,
1 ec or DarrellG.f. Licensing Division n Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operating License NPF-35
.?.
Safety Evaluation 3.
Federal Reoister Notice 4
Amandment No. 3 to Indemnity M... ~ C QWh Agreement B-100 Cortified Ey G
cc w/ enclosures:
See next page
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NUCLEAR REGOLATORY COMMISSION
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DUKE POWER COMPANY NORTH CAR'0 LINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC. COOPERATIVE, INC.
DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 6
FACILIT OPERATING LICENSE License No. NPF-35 1.
The Nuclear Regulatory Commission (the Comission or the NRC) has found that:
A.
The application for license filed by the Duke Power Con:pany acting for itself and North Carolina Electric Membership Corporation and
. Saluda River Electric Cooperative, Inc. (the licensees) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set agencies or bodies have bee,and all required. notifications to other forth in 10 CFR Chapter I; n duly made; k
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B.
Construction of the Catawba Nuclear Station, Unit 1 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-116 and the application, as' amended, the provisions of the Act and the regulations of the Commission; C.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Com-mission (except as exempted from compliance in Section 2.D. below);
D.
There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Sec-tion 2.D. below);
E.
Duke Power Company
- is technically qualified to engage in'the activ-ities authorized by this license in accordance with the Consnission's regulations set forth in 10 CFR Chapter I; F.
The licensees have satisfied the applicable. provisions of 10 CFR Part
- Duke Power Company is authorized to act as agent for the North Carolina
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Electric Membership Corporation and the Saluda River Electric Cooperative, Inc., and has exclusive responsibility and control over the physical con-struction, operation and maintenance of the facility.
999G030099-850117 f'DR ADOCK 05000413
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(22) Progress of Offsite Emergency Preparedness (Section 13.3, SER, SSER #1, SSER #2, SSER #3, SSER.W4)
In the event that the NRC finds that the lack of progress in com-pletion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a na.ior substantive problem exists in achieving or maintaining an adequate state of preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.
1 (23) Emergency Preparedness Issu~es (ASLB PID, 9/18/84)
By June 4,1985, Duke Power Company shsll have submitted for staff review and received staff approvel on the following items:
1.
The Public Information Brochure shall state that high levels of radiation are harmful to health and may be life threatening.
Such statements shall be contained within that portion of the brochure that deals with actions.to be taken in the event of an emergency.
2.
The warning signs.and decals shall specify the types of emergencies they cover including nuclear.-
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3.
The warning signs and decals shall. notify transients as to where they can obtain local emergency information, as provided in NUREG-0654 Evaluation Criterion II.G.2.
4.
The emergency plans shall reflect the kinds of locations within the plume exposure EPZ wherein the warning signs and decals and emergency response information will be placed and the pro-
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cedures employed to assure that sufficient numbers are being distributed to effectively reach transients, and that the plans are implemented.
5.
Comprehensive plans shall provide for early notification to Carowinds of a radiological emergency at Catawba and for evacuation of Carowinds.
The plans shall describe the responsibilities of the emergency response organizations of
'Mecklenburg and York Counties and provide for the coordina-tion of their efforts avong themselves and with Carowinds' officials. The plans anall provide for inanediate notifica-tion of patrons and staff of Carowinds at the time of the precautionary closing of the Dark, of the cause of the emergency.
The means to implement the plans shall be made available.
D.
The facility requires exemptions from certain requirements of Appendices A, E and J to 10 CFR Part 50.
These include (a) partial exemption from General Design Criterion 1 of Appen. dix A, with respect to the upgrade to safety-related of the pressurizer power
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I operated relief valves (P0DVs) and steam generator PORVs until first refueling (Section 5.4.4 of SER and SSER 2, and Section 15.4.4 of SSERs 3 and 4), (b) exemption from the requirements of Appendix E, IV.F, insofar as they may require the active participa-tion o,f all Crisis Management Center personnel for the Catawba Station emergency preparedness exercises (Section 13.3 of SSER 4),
(c) partial exemption from the requirement of paragraph III.D.2(b)(ii) of Appendix J, the testing of containment airlocks at times when the containment integrity is not required (Section 6.2.6 of the SER, and SSERs 3 and 4), (d) exemption from the requirement of paragraph III.A (d) of Appendix J, insofar as it requires the venting and draining of lines. for type A tests (Section 6.2.6 of SSER 3), and (e) partial exemption from the requirements of paragraph III.B of Appendix J, as it relates to bellows testing (Section 6.2.6 of the SER and SSER 3). These exemptions are authorized by law and will not endanger life or property or the comon defense and security and are otherwise in the public interest. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12.
With the granting of these exemptions, the facility. will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission.
- p E.
Duke Power Company shall fully implement and maintain in effect all provisions of the Commission approved physical security, guard training and qualification and safeguards contingency plans, including amendments made pursuant to the authority of 10 CFR 50.54 (p).
The approved plans which contain 10 CFR 73.21 infor-mation are collectively entitled: " Catawba Nuclear Station Physical Security Plan" Revision 1 dated February 1, 1982, with additional pages dated May 17, 1982 (transmittal letter dated May 27, 1982), Revision 2 dated May 17, 1982 (transmittal letter dated May 18,1982), Revision 3 dated July 20, 1982 (transmittal letter dated August 18,1982), Revision 4 dated June 1, 1983 (transmittal letter dated June 20,1983), Revision 5 dated April 13,1984 (transmittal letter dated April 16, 1984) and Revision 5 additional pages dated June 25, 1984 (transmittal letter dated July 3, 1984); and the " Catawba Nuclear Station Safeguards Contingency Plan" dated June 29, 1981 (transmittal i
letter dated June 30,1981), Revision 1 dated February 1, 1982 (transmittal letter dated February 10,1982), Revision 2 dated January 3,1983 (transmittal letter dated January 25,1983),
Revision 3 dated April 13,1984 (transmittal letter dated April 16, 1984); and the " Catawba Nuclear Station Training and Qualification Plan" dated October 21, 1981, and Revision 4 dated October 25, 1983.
F.
Reporting to the Commission i
Duke Power Company shall report any violations of the requirements contained in Section 2. Items C.(1), C.l3) through. C.'(23) of this license.
Initial notification shall be made within twenty-four 1
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NUREG-0954 Supplement No. 4 I'
Safety Evaluation Report i
related to the operation of i
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 i
ti lj Duke Power Company, et al.
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i, U.S. Nuclear Regulatory l
Commission Offi i
ce of Nuclear Reactor Regulation December 1984
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The staff has completed the review of the information submitted concerning the operability of the ITT Grinnell 4-in. diaphragm valves and the Borg Warner 4-in.
gate valves used to purge and vent the containment at Catawba.
The staff finds that the information submitted demonstrated the ability of these valves to close against the buildup of pressure in the containment in the event of a DBA/LOCA, and therefore they may be used to purge and vent the containment during operat-ing conditions 1, 2, 3, and 4.
6.2.5 Combustible Gas Control System By letter dated August 31, 1984, the licensee confirmed that the Unit 1 distrib-uted ignition system used to control hydrogen released during a LOCA is installed and operable and that the system will be energized following verification of a valid safety injection signal.
On this basis, the staff concludes that the first element of proposed License Condition 11 of the July 18, 1984, proposed low power license has been satisfactorily resolved and is no longer required.
With regard to the second element of License Condition 11, the licensee submitted responses to all outstanding staff questions on hydrogen control measures on May 22, 1984.
The staff has reviewed the material provided in the licensee's submittal and concludes that the responses do not adequately resolve the staff's concerns.
Accordingly, by letter dated October 3, 1984, the staff requested additional information and will modify the license to require the licensee to submit upgraded analysis for staff review and approval by April 1, 1985.
The staff has determined that full power operation is acceptable before final resolution of these matters because the staff has reviewed a virtually identi-cal hydrogen mitigation system installed at the McGuire Nuclear Station and sup-plementary analyses submitted for the Catawba facility and concludes that there i
is reasonable assurance that the hydrogen control measures used at Catawba will satisfactorily mitigate the consequences of the hydrogen release generated during the more probable degraded core accident sequences.
The staff concludes that a date of April 1, 1985, will provide adequate time for resolution of the remaining issues without undue risk to public health and safety.
Final resolution of this item will be addressed on completion of the staff's review.
6.2.6 Containment Leakage Testing By letter dated September 19, 1984, the licensee requested a partial exemption from the requirements of paragraph III.D.2(b)(ii) of Appendix J to 10 CFR 50 for the life of the facility.
Paragraph III.D.2(b)(ii) of Appendix J states:
l Air locks opened during periods when containment integrity is not re-quired by the plant's Technical Specifications shall be tested at the end of such periods at not less than Pa.
As discussed in Supplement 3 to the SER, whenever the plant is in cold shutdown (mode 5) or refueling (mode 6), containment integrity is not required.
- However, if an air lock is opened during modes 5 and 6, paragraph III.D.2(b)(ii) of Ap-pendix J requires that an overall air lock leakage test at not less than the i
Catawba SSER 4 6-5 i
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3 9W calculated peak containment pressure from a design-basis LOCA (Pa) be conducted h
b fore plant heatup and startup (i.e., entering mode 4).
The existing air lock
@c d:crs are so designed that a full pressure (i.e., Pa = 14.7 psig) test of an-cntire air lock can only be performed after strongbacks (structural bracing)
(P-have been installed on the inner door.
Strongbacks are needed because the pres-sure exerted on the inner door during the test is in a direction opposite to ilk-that of the accident pressure direction.
Installing strongbacks, performing
&c th test, and removing strongbacks requires at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per air lock (there are two air locks) during which access through the air lock is prohibited.
W lC If the periodic 6 month test of paragraph III.D.2(b)(i) of Appendix J and the t:st required by paragraph III.D.2(b)(iii) of Appendix J are current, no mainte-p@'
nance has been performed on the air lock, and the air lock is properly sealed, W
there should be no reason to expect the air lock to leak excessively just because jJ it has been opened in mode 5 or mode 6.
4r Accordingly, the staff has concluded that the licensee's proposed approach of M
substituting the seal leakage test of paragraph III.D.2(b)(iii) for the full-Q pressure test of paragraph III.D.2(b)(ii) of Appendix J is acceptable when no 5/
taintenance has been performed on an air lock. Whenever maintenance has been is p:rformed on an air lock, the requirements of paragraph III.D.2(b)(ii) of Ap-E p:ndix J must still be met by the licensee.
i Therefore, on the basis of the foregoing and in accordance with 10 CFR 50.12(a),
k the staff has concluded that the partial exemption from paragraph III.D.2(b)(ii) p.
of Appendix J to 10 CFR 50, as discussed above, is authorized by law, will not 4
endanger life or property or the common defense and security, and is otherwise t
in the public interest.
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j Catawba SSER 4 6-6
i Federal Register / Vol. 49. No.199 / Friday. October 12, 1984 / Notices 40133 0
h period of time due to the release of IlCl Advisory Panel for Cellular Physiology; 11
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and AbOs.
Meeting 10 CFR Part 50. Appendix J, paragraph A isory o mi e ct. as a
- nded, e 1t req i e o e !! cont.ainment 4 gh s sp yea) n a
launch transition of 5/ year in FY 1984 to Pub. L 92-403, the National Science afrlock tests to be performed if an 24/Icar in FY 1988-Foundation announces the following airlock is opened during Modes 5 or 6.
An overall airlock leakage test at not t "I.
tess than P must be conducted prior to j
DATE Comments must be received on or J
before November 13.1984.
Name: Advisory Panel for Cellular plant heatup and startup (i.e., entering Ph iokg M0dC 4}'
b ADDRE5s: National Aeronautics and
.te nd time: Monday. Tuesday and Space, Admmistration. Headquarters.
Wednesday. October 29. 30. and 31.1984.
(([
i Weshmgton. DC 20546.
from 900 a.rn. until 6ou p.m.
FOR FURTHER INFORMATION CONTACT:
Place: Room 338. National Science By letter dated August 2.1983. the Paul Wetzel. 202-453-1872.
Foundation,1800 C Street. NW Washington.
licensee requested a change in the
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C.
CWre % Char Madon Mndal SUPPLEMENTARY INFORM.1 TION:"Ihe I
Environmental Assessment for SRM Type of meeting-Closed.
Specification 4.6.1.3.b which currently Contact person: Maryanna P. llenkart.
requires overall containment airlock Production was completed in July 1984 EIS's for the SRM DDT&E Program, anEl h2.Telep 8'
aka es t e
.. if e
P the Space Shuttle Program, and the Purpose of advisory panel: To provide INTEGRITY was not required...
subject Environmental Assessment are advice and recommendations concerning cvailable upon request.
eupport for research in Cellular Physiology.
The proposed change would be to
==
Conclusion:==
The SRM production Agenda:To review and evaluate research require the overall airlock leakage test program will not result in any significant proposals as part of the selection process for to be performed **... when
- werde.
maintenance has been performed on the adverse environmental impacts. No EIS is required.
E'***"I ' closing:De Proposals being airlock that could affect the airlock reviewed include information of a proprietary sealing capability." This change requires Dated: October 4.1964.
or confidential nature, including techrucal an exemption from the requirements of C. Robert N smith, information. financial data, such as salaries; Appendix j. to 10 CFR Part 50.
3 Associate AdminisuntorforManagement.
and personal information concerning
.Re existing airlock doors are so i duaIsasa a d i e p' Po ab
.e ption ) nd designed that s' full pressure,i.e '(14.8 Int ow. s+-mm nieho-n-at ass'..]
(6) of 5 U.S.C. 552b(c). Govemment in the psig), test of an entire airlock can only
- * #5" Sunshine Act.
be performed after strong backs NATIONAL SCIENCE FOUNDATION Authority to close meeting His (structural bracing) have been installed determination was made by the Committee on the inner door. Strong backs are Advisory Cornrnittee for International Management Officer pursuant to provisions needed since the pressure exerted on Programs; Meeting of Section 10(d) of Pub. L 92-463. ne the inner door during the test is in a Cmmlun Management Oscer was direction opposite to that of the accident In accordance with the Federal Advisory Comrr.ittee Act. Pub. L 92-463, f
pressure direction. Installing strong on b Di e o onkly the National Science Foundation back. performing the test, and removing e, 3979, announces the following meeting:
Dated: OctoberD.1984.
8'.rong backa requires at least 6 ours per adock @ere are 2 aMs) lock is%g Name: Advisory Committee for M.a. w;nkler*
which access through the air International Programs.
C##*i ## M#"###*#"
ff"^
prohibited.
Dater October 29.1984.900 a.m. to 500 p m.:
(71t Dec.e4-D0:2 Fned so-n* a.s I
If the periodic 6-month test of October 30.1984, sto a.m. to 3o0 p.rn.
eumo cooc ress es.*'
paragraph !!!.D.2(b)(i) of Appendix l and Place: National Science Foundation.1800 G the test required by paragraph f CLEAR REGULATORY
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!!!.D.2(b)(iii) of Appendix ] are current.
Street. NW. Washington, DC: Room 1224.
Type of meeting: Open.
NU not maintenance has been performed on Contact person: Dr. Bodo Bartocha-CCMMISSION the airlock, and the airlock is properly Director. Dms, ion of International Programs.
National Science Foundation. Washington.
(Docket Nos.60-369 and 50-370]
sealed, there should be no reason to DC 20550 Tetephone (202) 357.-9552.
i expect the airlock to leak excessively i
I just because it has been opened in Mode Summaryof minutes: May be obtained Duke Power Co. (McGuire Nuclear 1
from contact Person.
Station, Units 1 and 2); Exemption i 5 or Mode 6. Accordingly, the staff finds Purpose of meeting:To provide advim
- that the licensee's proposed approach of recomrnendations. and oversight related to I
relying on the sealleekage test of i
peratsain Duke Power Company (the licensee)is paragraph !!!.D.2(b)(iii) of Appendix j is scie I
ne i g Agende: October 29 Welcome andinitial j the holder of Facility Operat,ng Licenses acceptable when no maintenance has i
briefing by NSF officiels. Science activities !
NPF-9 and NPF-17 which authorize the been performed on an airlock, and will l
with developing countrice. UNESCO.
operation of the McGuire Nuclear not adversely affect maintaining l
committee disonssions.
Station. Unita 1 and 2 (the facilities). The containment integrity. Whenever October 30: Eastern European program.
licenses provide, among other things.
maintenance has been performed on an uport controls update. Discussion-that they are subject to all rules.
airlock. the test requirement of l
casignments. and future agendas.
regulations, and Orders of the Nuclear paragraph I!!.D.2(b)(ii) of Appendix l l
Dated. October 9.1944.
Regulatory Commission (the must still be met.
M. Rebecca Wink!*'-
Commission).The facilities comprise O,
Committee M.mqyement Officer.
two pressurized waterreactors at the l
p o.a.,non ra.a no um.s i
licensee's site ip Mecklenburg County.
Accordingly, the Commisslon has mumocoocrsau sas North Carolina.
determined that an enemption in l
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Federal Register / Vol. 49. No.199 / Friday October 12. 1984 / Notices 40134 accordance with 10 CFR Mt12 in pneumatic testing as un acceptable Finding of No Significant Impact S.
authorized by law, will not endanger life method of pressure testing.
He Commission has determined not or property or the comrnon de fense and The licensee proposes to meet the to prepare en environmental impact security and is otherwise in the pubhc following alternative requirements:
statement for the proposed relief.
B P # "the fo #
interest.Therefore, the requested
..neumatic testing may be used in place of d
we concMe exemptinn from the containment airhick hydretatic when the system is not enwromnental anosment.,ll not have a lest requirements nf to CFR Part 50.
compatible with a liquid test medium or the
"'.the proposed act,on wi i
Appendix J. Section ll).D.2(b)(ii) is sysicm is normally charged with a gas.
significant effect on the quality of the hereby granted.
provided code-required pressures and human environment.
Pursuant to 10 CFR 51.32. the temperatures are met.-
For further details with respect to this
- [,
Commission has determined that the The Needfor the Proposed Action action. see the application for the reliefs issuance of this exemption will have no The request for relief from Article dated June 21.1984 and supplement significant impact on the environment IWD 5222 and IWC 5200 is required dated August 15.1984, wh!ch are
.i (49 FR 38425).
because there is no viable means of available for public inspection at the Commission's Public Document Room.
t' This exemption is efective upon heating certain components, a e stems. or 1717 H Streets. NW., Washington. D.C.
.j issuance.
portions of syst' ems.
and at the Salem Free Ubrary;112 West Dated at Bethesda. Maryland this 5th day The request for relieI from Article Broadway. Salem County. New Jersey f
of October 1964.
IWA 5200. ASME Code,1974 Edition.
08079.
,s For the Nuclear Regulatory Commisston.
stems from the fact that no provision Dated at Dethesda Maryland, this 2nd day i
Darrett c. usenhut, Director. Division o/ Licensing. Office of syskrns not compatible with a hqu.d of October 1964.
j i
NucleorReactor Regulation.
test medium or systems normally For the Nuclear Regulatory Commission.
tI ni car.
r nas rd.na.n* an I charged with gas.This section of the Darreu G. Eisenhut.
I Code was charEed in the 1977 Edition to Director. Division of Licensing. office of perrnit pneumatic testing in lieu of guef,,, g,,,,,, g,guf,,j,,,
hydrostatic test;ng for those systems.
. 7 m,,y IDocket Noa. 50-272 and 50-3111 EnvironmentalImpacts of the Proposed eau.c cooe tsoo-et-u Public Service Electric and Gas Co, et. Action al.; Environmental Assessment and
%e proposed relief will provide a IDocket No.50-4061 Finding of No Significantimpact degree of assuratice of operability that is The U.S. Nuclear Regulatory equivalent to that prescribed by the Tuskegee In'stit'ute; Order Authorizing I
Commission (the Commission)is ASME Code. Consequently, the Disposition of Component Parts and I
consi,dering Issuance of relief from probability of the components not Termination of Facility Ucense certam requirements of ASME Code operating properly will not be increased By application dated August 9.1983.
Section XI to Public Service Electric and and post. accident radiological releases Tuskegee Institute (the licensee)
Gas Company Philadelphia Electric will not be greater than previously requested authorization from the I
Company, Delmarva Power and ught determined nor does the proposed relief Nuclear Regulatory Commission (the Company, and Atlantic City Electric otherwise affect radiological plant Commission or NRC) to dismantle the Company (the licensees). for the Salem effluents.Therefore, the Commission Nuclear Generating Station, Units 1 and concludes that there are no significant Tuskegee AGN-201 Reactor (the
- 2. Salem County, New Jersey.
radiological environmental impacts facility). a research and training reactor associated with these proposes reliefs.
loca'ed on the Institute's campus at Tuskegee. Alabama, to dispose of the Environmental Assessment With regard to potential non.
Identification ofPmposed Action radiological impacts, the propa ied relief component parts, and to terminate Facility I.icense No. R-122.The involves features located entiteiy within authorization would allow the licensee The relief pertains to the Ten. Year the restricted area as defined in 10 CFR to dismantle the facility, dispose of the System Leakaga a.nd Hydrostatic In.
Part 20. It dors not affect non.
component parts in accordance with the Service Inspection programs for Class 1.
radiological plant effluents and has no application. and terminate Facility
- 2. and 3 components. One relief relates other environmental impact. Therefore, License No. R-122. A " Notice of to Articles IWD 5222 and IWC 5200.
the Commission concludes that there are Proposed Issuance of Order Authorizing
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which require system leakage and no significant non-radiological Disposition of Component Parts and e
hydrostatic pressure test temperatures environmentalimpacts associated with Termination of Facility license." dated be "not less than 100 'F."
%e licensee proposes to meet the the proposed relief.
June 5.1984 was published in the followicg alternative requirements.-
Alternats.re use of Resources Federal Register on June 12,1984. No request for a hearing or petition for
" "8 This action involves no use of leave to intervene was filed following resources not previously considered in notice of the proposed action.
fer t c sie I ponents a mee i e requirements specified by fracture prevention the Final Environmental Statement The Commission has found the facility criteria. If fracture toughness criteria were (construction permit and operating has been decontaminated, that j
neither specified not required in the construction of the components, the owner license) for the Salem Nuclear satisfactury disposition has been made will determine the test temperature. No limit Cencrating Station. Units 1 and 2.
of the fuel, and the component parts can on system test temperature in required for Aycucies and Persons Consulted lie disposed of without restriction in systems or portions of systema constructed entirely of austenitic siccl.
De NRC staff reviewed the nccordance with the Commission's De second relief relates to Article lire, sees
- request and did not consult regulations in 10 CFR Chapter 1. and in a manner not inimical to the common IWA 5200. which does not include nther agencies or persons.
/
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f jo, UNITED STATES y
' y ( (,p, NUCLEAR REGULATORY COMMISSION E
WASHING TON, D. C. 20555
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,o November 1, 1984 Docket No. 50-416 Mr. J. B. Richard Senior Vice President, Nuclear Mississippi Power & Light Company P.O. Box 23054 Jackson, Mississippi 39205
Dear Mr. Richard:
Subject:
Issuance of Facility Operating License NPF Grand Gulf Nuclear Station, Unit 1 The U.S. Nuclear Regulatory Comission has issued the enclosed Facility Oper-ating License NPF-29, together with Technical Specifications and Environmental Protection Plan for the Grand Gulf Nuclear Station, Unit 1.
This action has been taken in accordance with the Comission's Order dated October 25, 1984.
This license authorizes operation at 100% power (3833 megawatts thenaal).
Also enclosed are copies of a related notice, the original of which has been forwarded to the Office of the Federal Register for publication, and of an evaluation of the effect of 40 years of license duration with respect to environmental matters.
Four signed copies of Amendment No. 2, Indemnity Agreement No. B-72 which covers the activities authorized under License No. NPF-29 are also enclosed.
Please sign all copies and return one copy to this office.
Sincerely, Darrell 6 Eis
, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operating License NPF-29 2.
Federal Register Notice 3.
40-year Evaluation 4.
Amendment No. 2 to Indemnity Agreement B-72 cc w/ enc 1:
See next page W
. - - - - - = - - - - - - -
/pa uc%'o, UNITED STATES
,. ( [ h NUCLEAR REGULATORY COMMISSION s E WASHINGTON. D. C. 20555
+....
MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 FACILITY OPERATING LICENSE License No. NPF-29 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found that:
A.
The application for license filed by Mississippi Power & Light Company, for itself and Middle South Energy, Inc., and South Mississi Power Association (hereinafter referred to as the licensees)ppi Electric complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B.
Construction of the Grand Gulf Nuclear Station, Unit 1 (the facility),
has been substantially completed in conformity with Construction Permit No. CPPR-118 and the application, as amended, the provisions of the Act, and the regulations of the Comission; C.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission (except as exempted from compliance in Section 2.D. below);
D.
There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be con-ducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below);
E.
The Mississippi Power & Light Company (MP&L)* is technically qualified to engage in the activities authorized by this operating license in accordance with the Commission's regulations set forth in 10 CFR Chapter I;
- The MP&L is authorized to act as agent for the co-owners and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.
9
.- (b)
Provide the second level undervoltage protection for Division 3 power supply (Item No. 373, T.S. Table 3.3.3-2).
(c)
Incorporate a bypass or coincident logic in all Division 1 and 2 diesel generator protective trips, except for trips on diesel engine overspeed and generator differ 2ntial cur-rent (Item No. 808, T.S. 4.8.1.1.2.d.16.d).
(38)
Control Room Leak Rate (Section 6.2.6, SSER #6)
MP&L shall operate Grand Gulf Unit I with an allowable control room leak rate not to exceed 590 cfm.
Upon restart of construc-tion of Unit 2 control room, MP&L will be permitted to operate at a leak rate of 760 cfm as evaluated in SSER No. 6.
D.
The facility requires exemptions from certain requirements of Appendices A and J to 10 CFR Part 50.
These include: (a) exemption from General Design Criterion 17 of Appendix A until startup following the first refueling outage, for (1) the emergency override of the test mode for the Division 3 diesel engine, (2) the second level undervoltage protec-tion for the Division 3 diesel engine, and (3) the generator ground over current trip function for the Division 1 and 2 diesel generators (Section 8.3.1 of SSER #7) and (b) exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J for the containment airlock testing following normal door opening when containment integrity is not required (Section 6.2.6 of SSER #7).
These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
There-fore, these exemptions are hereby granted pursuant to 10 CFR 50.12.
With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act and the rules and regulations of the Comission.
E.
MP&L shall maintain in effect and fully implement all the provisions of the Commission-approved physical security plan, guard training and qualification plan and safeguards contingency plan, including amend-ments made pursuant to the authority of 10 CFR Section 50.54(p).
The approved plans, which are safeguards information protected under 10 CFR 73.21, are collectively entitled Grand Gulf Nuclear Station
" Physical Security Plan," Revision 1, 2 and 3; the Grand Gulf Nuclear Station " Security Training and Qualification Plan," and the Grand Gulf Nuclear Station " Safeguards Contingency Plan." The identification of vital areas and measures used to control access to these areas, as described in the physical security plan, may be subject to amendments in the future based upon a confirmatory evaluation of the plant to determine those areas where acts of sabotage might cause a release of radionuclides in sufficient quantities to result in dose rates equal to or exceeding 10 CFR Part 100 guidelines.
g 3,.
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.g UielTED STATES NUCLEAR REGULATORY COMMISS N
,., V;,,#., j WASHINGTON, D. C. 20555
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P FEB 141985 J
Docket No. STN 50-454 Mr. Deniils 1. Tar ~r'ar
~
Director of Nuclear Licensing Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690
Dear Mr. Farrar:
~
Subject:
Issuance of Facility Operating License NPF Byron Station, Unit 1 The U.S. Nuclear Regulatory Comission (NRC) has issued the enclosed Facility Operating License NPF-37, together with Technical Specifications and Environ-mental Protection Plan for the Byron Station, Unit 1.
Based upon the findings of the Comission as reflect 4d in the enclosed license and the favorable vote by the Comission on full-power operation, License No. NPF-37 authorizes oper-ation of the Byron Station, Unit I at reactor core power levels not in excess of 3411 megawatts thennal (100% power), and supersedes License No. NPF-23, issued on October 31, 1984.
Enclosed is a copy of a related notice, the original of which has been for-warded to the Office of the Federal Register for publication.
Two signed copies of Amendment No. 2 to Indemnity Tgreement No. B-97 which covers the activities authorized under License No. NPF-37 are also enclosed.
Please sign both copies and return one to this office.
An Assessment of the Effect of License Duration on Matters Discussed in the Final Environmental Statement for the Byron Station, Unit 1 is enclosed as Sincerely,
)Y rahia, eting Director Frank J.
Division of Licensing
/
Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operating License NPF-37 2.
Federal Reoister Notice 3.
Amendment No. 2 to Indemnity Agreement No. B-97 4.
Assessment of the Effect of License Duration on Matters Discussed in the FES cc w/ enclosures:
See next page
Uh*lTED TTATEs f
~e r;UCLEAR REGULATORY COMi!.;SS!ON y.,/.V;,9 E
WASHINGTON, D. C. 20555
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. - p COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. I' I
FACILITY OPERATING LICENSE I
License No. NPF-37 1.
The Nuclear Regulatory Commission (the Commission) has found thet:
A.
The application for license filed by Commonwealth Edison Company (licensee), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commis-sion's regulations set forth in 10 CFR Chapter I, and all requiced notifications to other agencies or bodies have been duly made; B.
Construction of the Byron Station, Unit No. 1 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-130 and the application, as amended, the provisions of the Act, and the regulations of the Commission; C." The facility will operate in conformity with the application, as am. ended, the provisions of the Act, a,nd the regulations of the Com-mission (except as exempted from compliande in Section 2.D below);
D.
There is reasonable assurance:
(i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in
'Section 2.D below);
E.
Commonwealth Edison Company is technically qualified to engage in the activities authorized by this license in accordance with the Commis-sion's regulations set forth in 10 CFR Chapter I; F.
Commonwealth Edison Company has satisfied the applicable provisions of 10 CFR Part 140 " Financial Protection Requirements and Indemnity Agree-j/
ments," of the Commission's regulations; G.
The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; e
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(15) Operating Staff Experience Requirements (Section 13.1.2.1, SSER #5)
" 'TheI' licensee shall have a licensed senior operator on each shift who has had'at least s'ix months of hot operating experience on a
'similar type plant, including at least six weeks at power levels greater than 20% of full power, and who has had start-up and shut-down experience, except as follows.
For those shifts where such an individual is not available on the plant' staff, an advisor shall be provi'ded who hYs had at least four years of power plant experience, including two years of nuclear plant experience, and who has'had at least one year of experience on shift as a licensed senior operator at a similar type facility.
Use of advisors who were licensed only at the R0 level will be evaluated on a case-by-case basis. Advisors shall be trained on plant procedures, tech-nical specifications and plant systems, and shall be examined on these topics at a level sufficient to assure familiarity with the plant.
For each shift, the remainder of the shift crew shall be trained as to the role of the advisors. These advisors shall be retained until the experience levels identified in the first sentence above have been achieved. The NRC shall be notified at least 30 days prior to the date that the licensee proposes to release the advisors from further service.
D.
The' facility requires exemptions from certain requirements of Appendices A, E and.J.to 10 CFR Part 50. These include.(a) An exemption from the require-ment of Paragraph III.D.2(b)(ii) of Appendix J, the testing of containment air locks at times when containment integrity is not required (Section 6.2.6 of the SER), (b) an exemption from GDC-2 of Appendix A, the requirement that structures, systems and components important to safety be designed to with-stand the effects of natural phenomena such as earthquakes (Section 3.'O of SSER #5), (c) an exemption from GDC-13 and GDC-17 of Appendix A, the require-ment that instrumentation be provided to monitor variables and systems over their anticipated ranges, and the requirement that provisions be included to minimize the probability of losing electric power (Section 9.5.4.1 of SSER #5), (d) an exemption from GDC-19 of Appendix A, the requirement that the control room have adequate radiation protection to permit access and occupancy under accident conditions (Section 6.5.1 of SSER #6), and (e) an exemption from the requirement of Section IV.F of Appendix E that a full participation emergency planning exercise be conducted sithin one year be-fore issuance of the first operatin operation above 5% of rated power (g license for full power and prior to
'/
Section 13.3 of SSER #6). These exemp-tions are authorized by law and will nnt endanger life or. property or the
common defense and security and are otherwise in the public interest.
Therefore, the.se exemptigns are hereby granted pursuant to 10 CFR 50.12.
~
With"ths granting of these exemptions the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
E.
The licensee shall maintain in effect and fully implement all provisions.
of the Commission approved Physical Security Plan, Guard Training and Quali-fication Plan, and Contingency Plan, including amendments made pursuant to the authority of 10 CFR 50.54(p). The approved plans which contain Safe-guards Information and are required to be protected against unauthorized disclosure in accordance with 10 CFR 73.21 are collectively entitled:
Commonwealth Edison Company, Byron Nuclear Power Station Physical Security -
Plan, Security Personnel Training and Qualification Plan *, and Safeguards Contingency Plan *, Revision 2 (May 1980), transmitted by letter dated May 2, 1980, as revised by Revision 3 (June 1980) transmitted by letter dated June 27, 1980, as revised by Revision 4 (August 1980) transmitted by letter of August 11, 1980, as revised by Revision 5 (January 1982) transmitted by letter of January 25, 1982, as revised by Revision 6 (April 1982) transmitted by letter dated April 19, 1982, as revised by Revision 7 (September 1982) transmitted by letters dated October 8 and December 22, 1982, as revised by Revision 8 (August 1983) transmitted by letters dated Septe~mber 16, 1983 and October 28, 1983, as revised by Revision 9 (October 1983) transmitted by letter dated November 17, 1983, as revised by Revision 10 (January 1984) transmitted by letter d,ated December 30, 1983, as revised by Revisions 11 and 12 (July and August 1984)'Eransmitted by letter dated August 29,' 1984.
F.
Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, the licensee shall report any violations of the requirements contained in Section 2.C of this license in the following manner:
initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency h'otification System with written followup within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c) and(e).
G.
The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims -
l l
- The Security Personnel Training and Qualification Plan and the Safeguards -
Contingency Plan are Appendices to the Security Plan.
i i
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the containment that will be isolated during the Type A tests.
However, if i
modifications should result in any such instrumentation lines beirg isolated during Type A tests, the applicant has committed to locally (Type C) leakage l
testing the lines and adding the measured leakage to the total leakage resulting from the Type A test.
Fluid systems which may be open to the containment atmosphere under postaccident l
conditions and become extensions of the boundary of the containment must, in general, be vented and drained during Type A testing so as to ensure exposure iy of the system containment isolation valves to containment air test pressure and lj postaccident differential pressure.
Certain exceptions are allowed, as noted
'f in paragraph III.A.1(d) of Appendix J to 10 CFR 50.
The applicant has not yet q
completed preparation of the Type A test procedures concerning venting and g
draining, but has committed to comply with the appropriate requirements of N
Appendix J to 10 CFR 50.
When the procedures are completed, the staff will confirm, prior to initial fuel loading, that proper venting and draining pro-y visions are employed.
The applicar.t will perform Type B tests (local leakage tests of containment i
J penetrations) preoperationally and also periodically at least once every two y
years, as required by Appendix J to 10 CFR 50.
For Type B testing of the
)y electrical penetrations, a permanently installed leakage surveillance system q
will be used for pressurization between the closure flanges of all the elec-9 trical penetrations at the same time.
The total measured leakage will be added M
to the total of Type B and Type C leakage. This method will be used in lieu of more conventional Type B testing.
The staff finds this acceptable.
'[
The leakage testing provisions for the two containment air locks for personnel are in compliance with the requirements of Appendix J to 10 CFR Part 50, with one exception which is discussed below.
Th$ applicant has requested an exemption for the Byron Units 1 and 2 from certain requirements of 10 CFR 50, Appendix J, paragraph III.D.2(b)(ii), which states:
" Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such period at not less than P '"
a Whenever the plant is in Mode 5 (cold shutdown), containment integrity is not required.
Hence, if an air lock is opened during Mode 5 operations, paragraph III.D.2(b)(ii) requires that an overall air lock leakage test at not less than P be conducted prior to entry into Mode 4.
a Even if the periodic 6-month test required by paragraph III.D.2(b)(i) of Appendix J has been satisfied, to meet the requirement of paragraph III.D.2(b)(ii),
no access to the containment can be allowed while preparing to leave Mode 5 until every air lock that has been opened in Mode 5 is first tested and the plant has entered Mode 4.
The test would effectively be required every time Mode 5 was entered.
The containment would have be cleared of personnel during performance-of this test or they would be required to remain inside containment during the test and until the plant reached Mode 4.
Often there are several minor operational and maintenance problems that require containment entry just prior to entering Mode 4; the special air lock test would have to wait until Byron SER 6-24 2
all problems requiring containment entry were first corrected.
This is a very restrictive requirement and would slow the process of returning to operation.
If the periodic 6 month test of paragraph III.D.2(b)(i) and the test required by paragraph III.D.2(b)(iii) are current, no maintenance has been performed on the air lock, and the air lock is properly sealed, there should be no reason to expect the air lock to leak excessively just because it has been opened in Mode 5 or Mode 6.
Accordingly, the staff concludes that the applicant's proposed approach of substituting the seal leakage test of paragraph III.D.2(b)(iii) is acceptable when no maintenance has been performed on an air lock.
Whenever maintenance has been performed on an air lock, the requirements of paragraph II.D.2(b)(ii) of Appendix J must still be met by the applicant.
Therefore, an exemption from this requirement [10 CFR 50, Appendix J, paragraph III.D.2(b)(ii)] is justified and acceptable for Byron Units 1 and 2, and appropriate requirements will be added to the plant Technical Specifications.
The applicant will perform Type C tests (local leakage tests of containment isolation valves) preoperationally and also periodically at least once every two years, as required by Appendix J to 10 CFR Part 50.
Not all containment isolation valves are required by Appendix J to be Type C tested; those contain-ment isolation valves not receiving Type C tests will be identified and justified during the staff's review of the plant Technical Specifictions, and the staff will ensure that the valves to be Type C tested will be properly identified at that time.
The applicant will place under administrative control the test, vent, and drain connections that are used to facilitate Type A, B, and C testing.
These con-nections will be closed and signed off and under periodic surveillance to ensure their integrity and to verify the effectiveness of administrative controls.
Based on the above discussion, the staff concludes that the proposed reactor containment leakage testing program is acceptable and complies with the require-ments of GDC 52, 53, and 54; Appendix J to 10 CFR 50; and 10 CFR 100.
Such compliance provides adequate assurance that containment leak-tight integrity can be verified periodically throughout service lifetime on a timely basis to maintain such leakage within the limits of the Technical Specifications.
Maintaining containment leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the contain-ment, the loss of the containment atmosphere through the leak paths will not be in excess of acceptable limits specified for the site.
6.2.7 Fracture Prevention of Containment Pressure Boundary The staff safety review assessed the ferritic materials in the Byron Units 1 and 2' containment system that constitute the containment pressure boundary to determine if the material fracture toughness is in compliance with the require-ments ist GDC 51, " Fracture Prevention of Containment Pressure Boundary."
~
GDC 51 requires that under operating, maintenance, testing, and postulated accident conditions (1) the ferritic materials of the containment pressure Byron SER 6-25
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UNITED STATES g3 g NUCLEAR REGULATORY COMMISSI N r,,Yy",f;t
-c j WASHINGTON. D. C. 20555 je
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mR 2 01985 l
b Docket No. 50-341 Dr. Wayne Jens Vice President - Nuclear Operations The Detroit Edison Company 2000 Second Avenue
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Detroit, Michigan 48226
Dear Dr. Jens:
Subject:
Issuance of Facility Operating License NPF-33, Fermi-2 The U.S. Nuclear Regulatory Comission (NRC) has issued the enclosed Facility Operating License NPF-33, together with the Technical Specifications and the Environmental Protection Plan for the Femi-2 facility. License No.
NPF-33 authorizes operation of Fermi-2 at reactor core power levels not in excess of 3292 megawatts themal (100% power).
Pending Comission approval, operation is restricted to power levels not to exceed 5 percent of full power (165 megawatts thermal).
Enclosed is a copy of a related notice, the original of which has been forwarded i
to the Office of the Federal Register for publication.
l Three signed copies of Amendment No. 26 to Indemnity Agreement No. B-20 which covers the activities authorized under License No. NPF-33 are also enclosed.
Please sign all copies and return one to this office.
l l
l l
C;.R 2 i' 1555 Dr. Wayne Jens.
An Assessment of the Effect of License Duration on Matters Discussed in the Final Environmental Statement for the Fenni-2 is provided in Enclosure 4.
Sincerely, I ) ugh L tr hompson r., Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operating License NPF-33 2.
Federal Reaister Notice 3.
Amendment No. 26 to Indemnity Agreement No. B-20 4.
Assessment of the Effect of License Duration on Matters Discussed in the FES cc w/ enclosures:
See next page 6
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jo, UNITED STATES
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p,g NUCLEAR REGULATORY COMMISSION
/9j WASHINGTON, D. C. 20555 CiRl DETROIT EDISON COMPANY WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED DCCKET NO. 50-341 FERMI-2 FACILITY OPERATING LICENSE License No. NPF-33 4
1.
The Nuclear. Regulatory Comission (the Comission) has found that:
A.
The application for license filed by the Detroit Edison Company and Wolverine Power Supply Cooperative. Incorporated (licensees) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 4
10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B.
Construction of Fermi-2 (the facility) has been substantially com-pleted in conformity with Construction Permit No. CPPR-87 and the appli-cation, as amended, the provisions of the Act, and the regulations of the Commission; C.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Com-mission, (except as exempted from compliance in Section 2.D below);
D.
There is reasonable assurance:
(i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set Section 2.D. below)pter I (except as exempted from compliance in forth in 10 CFR Cha l
E.
The Detroit Edison Company
- is technical.ly qualified to engage in the activities authorized by this license in accordance with the Comis-sion's regulations set forth in 10 CFR Chapter I; L
l F.
Detroit Edison Company and Wolverine Power Supply Cooperative, Incor-porated, have satisfied the applicable provisions of 10 CFR Part 140 i
l
" Financial Protection Requirements and Indemnity Agreements," of the Comission's regulations; i
l
- Detroit Edison Company is authorized to act as agent for Wolverine Power Supply Cooperative, Incorporated, and has exclusive responsibility and control over the i
physical construction, operation and maintenance of the facility.
L.
- -, ~ + - - - - - - -,.. - -
. (17) _ Iodine / Particulate Sampling System (Section 22, Item II.F.1, SSER #5)
Prior to startup following the first refueling outage, DECO shall demonstrate that the operating iodine / particulate sampling system will perform its intended function.
(18) Emergency Planning In the event that the NRC finds that the lack of progress in completing the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of Section 50.54(s)(2) of 10 CFR Part 50 will apply.
(19) Emeraency Response Capability (NUREG-0737, Supplement No. 1)
DECO shall complete the required emergency response. capabilities as described in Attachment 2 to this license, which is incor-porated into this license.
(20) Generic Letter 83-28 (Required Actions Based on Generic Implications of Salem ATWS Events)
DECO shall submit responses to and implement the requirements of Generic Letter 83-28 on a schedule which is consistent with that given in its letters dated November 3, 1983, and November 29, 1984.
D.
An exemption from certain requirements of Appendix J to 10 CFR Part 50, is described in Supplement No. 5 to the SER. This exemption is author-ized by law and will not endanger life or property or the common defense i
and security and are otherwise in the public interest. Therefore, this exemption is hereby granted pursuant to 10 CFR 50.12. With the granting of this exemption, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
E.
DECO shall maintain in effect and fully implement all provisions of i
the Comission approved physical security, guard training and qualifi-l cation and safeguards contingency plans, including amendments made pursuant to the authority of 10 CFR 50.54(p). The approved plans, which contain information described in 10 CFR 73.21 are collectively entitled, "Enrico fermi Atomic Power Plant Unit 2 Physical Security Plan" Revision 4, dated April 1983 (transmitted June 29, 1983) (It is noted that Revision 4 is a completely reformatted revision which replaces all previous revisions); Amendment 5, dated.luly 1984
(
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NUREG-0798 Supplement No. 5 Safety Evaluation Report related to the operation of Fermi-2 Docket No. 50-341 Detroit Edison Company, et al.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l
l March 1985 l
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- I
In addition to the two innermost, quick acting isolation valve is located in each of the four main steam lines to pvalve leakage control volume for the back up pressurization rovide a redundant of the Fermi-2 Technical Specifications requires that functi L
system.
Paragraph 3.6.1.4 tests of the leakage control system be performed ever onal oporability J
the leak tightness integrity of this third isolation 3
y 18 months to verify that this leakage control system has the capability to mai t ivalve and
)
at a pressure greater than that in the containment
)
n a n the control volume We conclude that leak testing of the main steam isol ti described above is an acceptable alternative to the s on valves in the manner a
Appendix J.
ments of Appendix J as described above.Accordingly, we hereby c require-Air Lock Testing In its letter dated January 26, 1985, of Appendix J to 10 CFR Part 50.from the full pressure air lock quested an exemption Specifically, this paragraph states:ragraph III.D.2
" Air locks open during periods when containment i t at the end of such periods at not less than Pa." req I
n egrity is not ested Condition 4 or 5, it must be tested s'atisfactorily r ng either Operational test prior to plant entry into Operational Conditio overall air lock door n 3.
The applicant states in its January 26th letter th hours would be required to perform the overall test as me of between 11 to 13 This time interval is required since the following stepsspecified in to complete the overall test:
lock door; (2) pressurize the air lock to Pa(1) install 14 tie downs to the inte lock, remove and store the tie downs.to stabilize and perform epressurize the air The applicant proposes in its January 26th letter III.D.2(b)(ii) of Appendix J. verifying the air lock integrity in lieu of c
e in Paragraph testing the seals of the inner and outer doors at Pa r thThe applican entire door. provided no maintenance has been performed on the a
the last successful test.
er than testing the r lock since g
criteria for the seal test is proposed to betest since th e source.
The acceptance 5 standard cubic feet per hour at a pressure,a leak rate less than or equal to lock maintenance will have been performed since the lastPa, equal to 5 overall air lock leakage test at Pa shall be performed prior to tional Condition 3.
successful test, an performed as described above, will provide reasonable assuI entering Opera-tightness integrity of the air lock will be maintainedrance that the leak-I i
Fermi SSER 5 6-4 l
On this basis, we find that the applicant's request for an exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J to 10 CFR Part 50, is acceptable.
Accordingly, we have incorporated the applicant's proposed method of performing air lock tests as discussed above and in the applicant's letter of January 26,' 1985, into the Fermi-2 Technical Specifications.
6.4 Control Room Habitability Systems 6.4.1 Radiological Dose Protection In the SER we issued in July 1981, we concluded that the calculated dose esti-mates for the Fermi-2 control room meet the requirements of General Design Cri-terion (GDC) 19 of Appendix A to 10 CFR Part 50.
We further concluded that the design of the Fermi-2 control room provides an acceptable means of maintaining the control room in a safe and habitable condition by providing adequate protec-tion under radiation accident conditions.
Subsequently, the applicant submitted Amendment No. 58 to its FSAR in July 1984 in which it revised its estimate of the time to draw down the secondary contain-ment to a partial vacuum.
(Refer to Sections 6.2.3 and 15.2.3 of this supple-ment.)
Based on the applicant's latest estimate of the secondary containment drawdown time, we have revised Table 6.4-1 of the SER and replaced it in its entirety with Table 6.4-1 of this supplement.
The only significant change in this table is the calculated thyroid and whole-body doses to the control room operators in the first eight hours following the initiation of the postulated loss-of-coolant accident.
The net result is that the total thyroid dose increases from an estimated value of 15.7 rem to 16.1 rem while the estimated whole-body dose increases from 1.50 rem to 1.53 rem.
Both these revised doses are within the guideline values contained in GDC 19 of Appendix A to CFR Part 50.
Accordingly, we find that the calculated dose estimates for the Fermi-2 control room still meet the requirements for GDC 19.
With respect to the habitability of the Fermi-2 control room under postulated accident conditions, our inspection of the Fermi-2 control room ventilating ducts during the week of November 5 through 9,1984, indicated that one of the assumptions used in estimating the dose rates to the control room operators might not be valid.
(We assume in our analysis that no more than 10 cubic feet
~
per minute of unfiltered air infiltrates into the control room.) The results of this inspection effort is described in deport No. 50-341/84-43, dated January 11, 1985.
In this report, we expressed our concern that contrary to the guidelines in Regulatory Guide 1.52, silicone sealants were used to seal leaks found during leakage acceptance tests on the control room filter system ducts and housings.
Both of these components are engineered safety features (ESF) and, therefore, must meet applicable guidelines.
In response to our concerns, the applicant submitted additional information in its letter dated January 8, 1985, to clarify its FSAR commitment regarding Regulatory Guide 1.52.
The applicant stated that the Fermi-2 heating, ventilating and air-conditioning (HVAC) ducts, filters and filter housings were designed prior to issuance of both Regulatory Guide 1.52 and ANSI /ASME Standard N509-1976, which is referred to in Regulatory Guide 1.52.
The applicant further stated that the Fermi-2 HVAC systems were designed to the standards available at that time; namely, Sheet Metal and Air-Conditioning Contractors' National Association, Inc. (SMACNA) High Velocity Duct Construction Standards and ORNL-NSIC-65.
The Fermi SSER 5 6-5
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UNITED STATES
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g NUCLEAR REGULATORY COMMI IO
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WASHINGTON. D. C. 20655 Y
,o Docket No. STN 50-482 Mr. Glenn L. Koester Vice President - Nuclear Kansas Gas and Electric ~ Company 201 North Market Street Post Office Box 208 Wichita, Kansas 76201
Dear Mr. Koester:
Subject:
Issuance of Facility Operating License NPF Wolf Creek Generating Station, Unit 1 The U.S. Nuclear Regulatory Comission (NRC) has issued the enclosed Facility Operating License NPF-42,' together with Technical Specifications and Environ-mental Protection Plan for Wolf Creek Generating Station, Unit 1.
Based upon the findings of the Comission as reflected in the enclosed license and the favorable vote by the Commission on full-power operation License No. NPF-42 authorizes operation of the Wolf Creek Generating Station, Unit I at reactor core power levels not in excess of 3411 megawatts thermal (100% power), and supersedes License No. NPF-32, issued on March 11, 1985.
Enclosed is a copy of a related notice, the original of which has been for-warded to the Office of the Federal Register for publication.
Four signed copies of Amendment No. 3 to Indemnity Agreement No. B-99 which covers the activities authorized under License No. NPF-42 are also enclosed.
Please sign all copies and return one to this office.
Sincerely.
I H gh
. Thompson Jr.,
ctor D
ion of Licensing Office of Nuclear Reactor Regulation
Enclosures:
1.
Facility Operating License NPF-42 2.
Federal Resister Notice 3.
Amendment Po. 3 to Indemnity Agreement No. B-99 cc w/ enclosures:
See next page
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k, UNITED STATES
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g NUCLEAR REGULATORY COMMISSION 5
- j WASHINGTON, D. C. 20555 l
KANSAS GAS AND ELECTRIC COMPANY KANSAS CITY POWER & LIGHT COMPANY KANSAS ELECTRIC POWER COOPERATIVE, INC.
DOCKET NO. STN 50-482 WOLF CREEK GENERATING STATION, UNIT NO. I FACILITY OPERATING LICENSE Lii:ense No. NPF-42 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for license filed by Kansas Gas and Electric Company, Kansas City Power & Light Company, and Kansas Electric Power Cooperative, Inc. (licensees), complies with the standards and requirements of the' Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I, and all required notifica-tions to other agencies or bodies have been duly made; B.
Construction of the Wolf Creek Generating Station, Unit No.1 (the facility) has been substantially completed in conformity with Con-struction Pennit No. CPPR-147 and the application, a.s amended, the provisions of the Act, and 'the regulations of the Comission; C.
The facility will operate in confonnity with, the application, as l
amended, the provisions of the Ac~ t, and the -regulations of the Com-l mission, (except as exempted from compliance in Section 2.D below);
I D.
There is reasonable assurance:
(1) that the activities authorized by this operating license can be conducted without endangering the l
health and safety of the public, and (ii) that such activities will i
be conducted in compliance with the Comission's regulations set l
forth in 10 CFR Chapter I, (except as exempted from compliance in l
Section 2D below);
E.
Kansas Gas and Electric Company
- is technically qualified to engage in the activities authorized by this license in accordance with the Connission's regulations set forth in 10 CFR Chapter I;
- Kansas Gas and Electric. Company is authorized to act as agent for the Kansas City Power & Light' Company and the Kansas Electric Power Cooperative. Inc.,
and has exclusive responsibility and control over the physical construction.
l operation and maintenance of the facility.
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.g (11) Steam Generator Tube Rupture (Section 15.4.4, SSER #5)
Prior to restart following the first refueling outage, KG&E shall submit for NRC? review and approval an analysis which demonstrates that the steam generator single-tube rupture (SGTR) analysis pre-sented in the FSAR is the most severe case with respect to the t
release of: fission products and calculated doses. Consistent with the analytical assumptions, the licensee shall propose all necessary changes to Appendix A to this license.-
(12). LOCA Reanalysis (Section 15.3.7, SSER #5)
Prior to restart following the first refueling outage, KG&E shall submit for NRC review and approval a reanalysis for the worst large break LOCA using an approved ECCS evalua. tion model.
j (13) Generic Letter 83-28 KG&E shall submit responses to and implement the requirements of i
Generic Letter 83-28 on a schedule which'is consistent with.
that given in their February 29, 1984 and February 6, 1985 letters.
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(14) Surveillance of Hafnium Contirol Rods (Section 4.2.3.1(10),SER and 55ER #2)
KG&E shall perform a visual. inspection of a sample of hafnium control rods during one of the first five refueling outages. A sumary of the. results of these inspections shall be submitted to the NRC.
D.
Exemptions from certain requirements of' Appendix J to 10 CFR Part 50, and from a portion of the requirements of General Design Criterion 4 of Appendix A to 10 CFR Part 50, are described in the Safety Evaluation a
Report.
These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions the facility will operate, to the ext 9nt authorized herein, in confonnity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission.
E.
KG8E shall fully implement and maintain in effect all provisions of the Comission approved Physical Security, Guard Training and Quali-fication, and Safeguards Contingency plans, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR
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NUREG-0881 Supplement No. 5 i.
Safety Evaluation Report re ated to the operation of
- , Wolf Creek Generating Station, Unit No.1 Docket No. STN 50-482 Kansas Gas and Electric Company, et al.
U.S. Nuclear Regulatory
-Cornmission
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Office of Nuclear Reactor Regulation 1
March 1985 L
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safety injection.
acceptable because the hydrogen monitoring system wil 6
requirements of TMI Action Plan Items II.E.4.2 and II.F.1, Attachment 6 6.2.6 2
Containment Leakage Testing Containment Air Lock Surveillance 3
By letter dated November 6, 1984, the applicant requested an exemp the requirement of paragraph III.D.2(b)(ii) of Appendix J to 10 CFR 50 J
graph III.D 2(b)(ii) of Appendix J states:
Para-i
" Air locks opened during periods when containment integrity is not required by the plant's Technical Sp tions shall be tested at the end of such periods at not less than P after each and every shutdown regardless of the pu The a.
f air lock test only when maintenance is performed In l
affect the air lock sealing capability.
tion from the requirements 'of Appendix J to 10 CFR 50.This proposed change req tion of this exemption request follows.
The staff's evalua-Whenever the plant is in cold shutdown (Mode 5) or refueling (Mode 6) tainment integrity is not required.
, con-However, if an air lock is opened during Modes 5 and 6, paragraph III.D.2(b)(ii) of Appendix J requires that an o air lock leakage test at not less than P, be conducted before plant h startup (i.e., entering Mode 4).
j that a full pressure test (P,, 48.0 psig) of an entire air lock ca performed after strong backs (structural bracing) have been installed on th inner door.
Strong backs are needed because the pressure exerted on the inner door durin direction.g the test is in a direction opposite to that of the accident pressure Installing strong backs i
backs require at least 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per a,ir lockuperforming the test, and removing strong I
L ring which access through the air lock is prohibited.
If the periodic 6-month test of paragr i
test required by paragraph III.D.2(b)( 4. i k X 2(b)(i) of Appendix J and the iii) of Appendix J are current, no main-tance has been performed on the air lock, and the air lock is properly sea there should be no reason to expect the air lock to leak excessively just b cause it has been opened in Mode 5 or Mode 6.
Accordingly, the staff concludes that the applicant must still meet its approach of substituting the seal leakage test of paragraph III.D.2(b)(iii) for the full pressure test of paragraph III.O.2(b)(ii) of Appendix J g
Therefore, an exemption from this requirement [10 CFR 50, Appendix J graph III.D.2(b)(ii)] is justified and acceptable for Wolf Creek Unit 1
, para-this subject are acceptable.the applicant's proposed changes to the Te and 6.6 Inservice Inspection of Class 2 and 3 Components This section was prepared with the technical assistance of Departme (DOE) contractors from the Idaho National Engineering Laboratory (INEL)
Wolf Creek SSER 5 6-2 i
s UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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CLEVELAND ELECTRIC ILLUMINATING
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Docket No. 50-440 OLUCH*E30 COMPANY, ET AL.
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50-441 OL
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(Perry Nuclear Power Plant,
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'65 JUL 25 P4:42 -
Units 1 and 2)
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CFFlCE Or 3im ja.
CERTIFICATE OF SERVICE 00CXETi EPvlu.
I hereby certify that copies of "NRC RESPONSE IN OPPOSITION TO OCRE MOTION TO REOPEN THE RECORD AND ADMIT CONTENTION ON PARTIAL EXEMPTION TO APPENDIX J" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, by deposit in the Nuclear Regulatory Commission's internal mail system, this 24th day of July,1985:
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James P. Gleason, Chairman Donald T. Ezzone, Esq.
Administrative Judge Assistant Prosecuting Attorney 513 Gilmoure Drive 105 Main Street Silvet Spring, MD 20901 Lake County Administration Center Painesville, Ohio 44077
- Mr. Glenn 0. Bright Administrative Judge Susan Hiatt Atomic Safety and Licensing Board 8275 Munson Road U.S. Nuclear Regulatory Commission Mentor, OH 44060 Washington, DC 20555 Terry J. Lodge, Esq.
- Dr.' Jerry R. Kline 618 N. Michigan Street, Suite 105 Administrative _ Judge Toledo, OH 43624 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission John G. Cardinal, Esq.
Washington, DC 20555 Prosecuting Attorney Ashtabula County Courthouse day Silberg, Esq.
Jefferson, Ohio 44047 Shaw, Pittman, Potts and Trowbridge 1800 M Street, NW Janine Migden, Esq.
Washington, DC 20036 Ohio Office of Consumers Counsel 137 E. State Street Columbus, OH 43215 t
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- Atomic Safety and Licensing Board Panel U.S. Nuclear-Regulatory Commission Washington, DC 20555
- Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, DC 20555
- Docketing & Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555
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,f Gecrge E..Jppnson Counsel foVNRC Staff f
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