ML20206G405

Applicant Answer to Petitioner Board of Commissioners of Orange County Contentions.* Requests That Technical Contentions in Section III & Environ Contentions in Section IV Not Be Admitted.With Certificate of Svc

ML20206G405
Person / Time
Site:	Harris
Issue date:	05/05/1999
From:	Oneill J CAROLINA POWER & LIGHT CO., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	Atomic Safety and Licensing Board Panel
References
CON-#299-20353 99-762-02-LA, 99-762-2-LA, LA, NUDOCS 9905100006
Download: ML20206G405 (150)

Text

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r NP,C May 5,1999 u) my -7 P 2 52 OD hk UNITED STATES OF AMERICA ,

l NUCLEAR REGULATORY COMMISSION 1

Before the Atomic Safety and Licensine Board l

In the Matter of )

)

l CAROLINA POWER & LIGHT ) Docket No. 50-400-LA .

i COMPANY )

) ASLBP No. 99-762-02-LA (Shearon Harris Nuclear Power Plant) 1 l

l APPLICANT'S ANSWER TO PETITIONER BOARD OF COMMISSIONERS OF ORANGE COUNTY'S CONTENTIONS OfCounsel: John H. O'Neill, Jr.

Steven Carr William R. Hollaway Legal Department SHAW, PITTMAN, POTTS CAROLINA POWER & LIGHT & TROWBRIDGE COMPANY 2300 N Street, N.W.

411 Fayetteville Street Mall Washington, D.C. 20037-1128 Post Office Box 1551 - CPB 13A2 (202) 663 8148 Raleigh, North Carolina 27602-1551 Counsel For CAROLINA POW" (919) 546-4161 LIGHT COMPANY b

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yb 00CKETED ty: ARC May 5,1999 99 MY -7 P 2 :52 J

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UNITED STATES OF AMERICA y, NUCLEAR REGULATORY COMMISSION ,

i Before the Atomic Safety and Licensine Board j In the Matter of ) f

) l CAROLINA POWER & LIGHT ) Docket No. 50-400-LA l COMPANY )

) ASLBP No. 99-762-02-LA l (Shearon Harris Nuclear Power Plant)  :

i l

APPLICANT'S ANSWERTO PETITIONER BOARD OF COMMISSIONERS OF ORANGE COUNTY'S CONTENTIONS OfCounsel: John H. O'Neill, Jr.

Steven Carr William R. Hollaway Legal Department SHAW, PITTMAN, POTTS CAROLINA POWER & LIGHT & TROWBRIDGE COMPANY 2300 N Street, N.W.

3-411 Fayetteville Street Mall Washington, D.C. 20037-1128 Post Office Box 1551 -CPB 13A2 _

(202) 663-8148 Raleigh, North Carolina 27602-1551 Counsel For CAROLINA POW" (919)546-4161 LIGHT COMPANY E

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I 70353 l ]

May 5,1999 i

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensine Board In the Matter of )

i )

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

(Sheaun Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA APPLICANT'S ANSWER TO PETITIONER BOARD OF COMMISSIONERS OF ORANGE COUNTY'S CONTENTIONS OfCounsel: John H. O'Neill, Jr.

Steven Carr William R. Hollaway Legal Department SHAW, PITTMAN, POITS CAROLINA POWER & LIGHT & TROWBRIDGE COMPANY 2300 N Street, N.W.

411 Fayetteville Street Mall Washington, D.C. 20037-1128 Post Office Box 1551 - CPB 13A2 (202) 663-8148 l Raleigh, North Carolina 27602-1551 Counsel For CAROLINA POWER &

(919) 546-4161 LIGHT COMPANY f,

m p TABLE OF CONTENTS D

I. INTRO D U CTI ON . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . . . . . . . .

II. STANDARDS FOR ADMISSIBILITY AND SCOPE OF CONTENTIONS.................. 2 3 A. Overview Of Admissibility Requirements......................................... ............... .. 2 B. Pleading Requirements and General Limitations On The Admissibility Of Contentions..........................................................................................................4

1. B asis with Specificity .. . . . . .. . . . . .. . . ... .. . .. . . . . .. . ... . .. . . .. . . .... .. ... .. .. . . .. . .. . . . . . . . . . . . . . . 4 y 2. Opposition to Applicant's Position...................................... ............ ..... ...... 5

3. Petitioner's Mistake Cannot Be the Basis of a Contention......................... ... 6

4. Challenges to Regulations Barred .................. ..................... .......... .............. 6 l

S. . Scope of Proceeding and Materiality............... .... ........................ ............... 7

6. Health and Safety Significance.... ............................................. ............... .... 9 D C. The Scope of a Contention Is Limited by Its Specific Bases.. ........................ . . 9 III. Group I: Technical Contentions ........... ... ...... . . ................ .. .................. ....................... 1 1 l

A. Contention 1: Inadequate Emergency Core Cooling and Residual Heat l

0 Removal...........................................................................................................12

1. The Co ntenti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. Applicant's Response to the Contention............................. ........ ................ 14 B. Contention 2: Inadequate Criticality Prevention............................................ ... 28 g 1. The C onte ntio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. Applicant's Response to the Contention... ........... . .................. ....... ........ 29

C. Contention 3

Inadequate Quality Assurance .................... ........... .................... 36

1. The C o ntenti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .

) 2. Applicant's Response to the Contention.............. ..................... ......... ........ 3 8

3. Summary of Response to Contention 3 ......... ........ ................... . ............. 4 8 j IV. Group II: Environmental Contentions ..................... ......................... ..... .................... 49

A. Contention 4

Proposed License Amendment Not Exempt from NEPA......... .. 50 j 1. The C o ntention . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .

2. Applicant's Response to the Contention......................... ............ ........... .... 51 B. Contention 5: Environmental Impact Statement Required .... ............... ... . . . 51

1. The Co ntentio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .!

2.' . Applicant's Response to the Contention..................... ...... .. ............ ....... 52 1

.m.

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o o i 1

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, C. Contention 6: Scope of EIS Should Include Brunswick and Robinson l

l Storage..............................................................................................................-53 i

> 1 I

lO 1. The C o nte nt ion . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. Applicant's Response to the Contention................................... ......... .. ..... 54 4

i D. Contention 7: Environmental Assessment Required ........................... .. ........... 59 )j

1. The Contention .... . ................... . ........................................................59

2. Applicant's Response to the Contention...... ................................................ 59

)

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! E. Contention 8: Discretionary EIS Warranted .............. ........ ...................... ........ 59 1

1. The C o nte ntior1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. ..

)

2. Applicant's Response to the Contention......................... .. . ... ... ................ 60

)

i IV. Conclusion ......................................................................................................65 1O  ;

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D May 5,1999 3 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensine Board D In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

(Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA D

APPLICANT'S ANSWERTO PETITIONER . BOARD OF COMMISSIONERS OF D ORANGE COUNTY'S CONTENTIONS I. INTRODUCTION O

In its Initial Prehearing Order of February 24,1999, the Atomic Safety and Licensing Board (" Licensing Board"), pursuant to 10 C.F.R. 2.714(b), required that Petitioner file by April 5,1999, a supplement to its hearing petition / intervention request, which supplement must D

include a list of contentions and supporting bases. The Licensing Board also directed that responses to Petitioner's supplement be filed by May 5,1999.

3 Contentions were filed on April 5,1999 by Petitioner Board of Commissioners of Orange County ( " Petitioner" or "BCOC").' Applicant Carolina Power & Light Company (" Applicant" or "CP&L") hereby submits the following Answer to BCOC's contentions. Prior to discussing 3 each contention, Applicant sets forth in Section II its statement oflaw on the relevant standards for admission of contentions. We address the Technical Contentions in Section III and Orange County's Supplemental Petition to Intervene ("BCOC Supp. Pet.")

D t

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Environmental Contentions in Section IV. For the reasons set forth with respect to each of the h

contentions, Applicant respectfully submits that none of the contentions should be admitted.

j II. STANDARDS FOR ADMISSIBILITY AND SCOPE OF CONTENTIONS  !

1 A. Overview Of Admissibility Requirements t

i O ]

The Commission's Rules of Practice at 10 C.F.R. { 2.714 set forth the requirements for the admission of contentions. In addition to demonstrating the required interest, a petitioner g; must submit at least one valid contention that meets the requirements of 10 C.F.R. { 2.714 in l 1

order to be permitted to participate in a licensing proceeding as a party. 10 C.F.R. @ 2.714(b)(1); l

. Yankee Atomic Electric Co. (Yankee Nuclear Power Station), CLI-96-7,43 NRC 235,248  ;

g (1996); Georgia Institute of Technolony (Georgia Tech Research Reactor, Atlanta, Georgia), l CLI-95-12,42 NRC 111,117 (1995).

For a contention to be admitted, it must meet the standards set forth in 10 C.F.R.

O 2.714(b)(2), which provides that "[e]ach contention must consist of:"

"a specific statement of the issue oflaw or fact to be raised or controverted", accompanied by O (i) a "brief explanation of the bases of the contention";

(ii) a " concise statement of the alleged facts or expert opinion" supporting the contention together with references to " specific sources and documents . . . on which the petitioner intends to rely g to establish those facts or expert opinion"; and (iii) "[s]ufficient information . . . to show that a genuine dispute exists with the applicant on a material issue oflaw or fact," which showing must include " references to the specific portions of the application . . . that the petitioner disputes and the supporting O reasons for each dispute . . . ."

10 C.F.R. 2.714(b)(2); Duke Energy Coro. (Oconee Nuclear Station, Units 1,2, and 3), CLI-99-11,49 NRC (1999)(slip op. at 5,6). The failure of a contention to comply with any one of these requirements is grounds for dismissing the contention. 10 C.F.R. 2.714(d)(2)(i);

O

O Private Fuel Storane. L.L.C. (Independent Spent Fuel Storage Installation), CLI-99-10,49 NRC O (1999)(slip OP at 10); Arizona Public Service Co. (Palo Verde Nuclear Generating Station, Units 1,2, and 3), CLI-91-12,34 NRC 149,155-56 (1991). Further, a contention must also be dismissed where the " contention, if proven, would be of no consequence . . . because it would O

not entitle [the] petitioner to relief." 10 C.F.R. 2.714(d)(2)(ii);XaDkee Atomic Electric Co.

(Yankee Nuclear Power Station), LBP-96-2,43 NRC 61, 78, a[.d, CLI-96-7,43 NRC 235

. (1996); Lone Island Lightinn Co. (Shoreham Nuclear Power Station, Unit 1), LBP-91-39,34 0

NRC 273,280-81 (1991).

The standards governing the admissibility of contentions are the results of the 1989 amendments to 10 C.F.R. Q 2.714. These amendments were intended "to raise the threshold for dmission of contentions." Rules of Practice for Domestic Licensing Proceedings -- Procedural Changes in the Hearing Process; 54 Fed. Reg. 33,168 (1989); see also Oconee, CLI-99-11, agga, slip op, at 5-6; Palo Verde, CLI-91-12, sup.ta,34 NRC at 155-56; Lone Island Lighting Co.

(Shoreham Nuclear Power Station, Unit 1), LBP-91-35,34 NRC 163,167 (1991). The requirements of the new rule are to be enforced rigorously: "[i]f any one . . . is not met, a contention must be rejected." Palo Verde, CLI-91-12, aggg,34 NRC at 155; see also Shoreham, LBP-91-39, sugg,34 NRC at 279. A licensing board must not overlook a deficiency in a contention or assume the existence of missing information.11; see also Policy on Conduct of Adjudicatory Proceedings,63 Fed. Reg. 41,872,41,874 (1998). The Commission has also recently reaffirmed its position that the 1989 amendments " effectively work to bar ill-defined

' anticipatory' contentions . . . Our revised rules do not permit ' vague, unparticularized contentions,' or ' notice pleading, with details to be filled in later.'" Oconee. CLI-99-11, sugg, slip. op, at 12 (citations omitted); Duke Power Co. (Catawba Nuclear Station, Units 1 and 2),

ALAB-687,16 NRC 460,468 (1982), vacated in oart on other arounds, CLI-83-19,17 NRC

-1041 (1983).

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)

B. Pleading Requirements and General Limitations On The Admissibility Of Contentions D

The detailed pleading requirements of 10 C.F.R. 2.714(b)(2)(i)-(iii) added by the 1989 amendments " heighten [ed] the specificity requirements for pleadings filed by parties seeking to 3 intervene in (formal] licensing proceedings." Yankee Nuclear, CLI-96-7, suora,43 NRC at 248, gitjug Union of Concerned Scientists v. NRC. 920 F.2d 50,51-52 (D.C. Cir.1990). Commission regulations and precedent include several general limitations on the scope ofissues that may 3 properly be raised and litigated in a licensing proceeding.

1. Basis with Snecificity Under the amended Rules of Practice, a petitioner must set forth "[a] brief explanation of D

the bases of the contention." 10 C.F.R. s 2.714(b)(2)(i). Further, a petitioner must provide:

A concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing, together with references to 3 those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion.

10 C.F.R. 2.714(b)(2)(ii).

The Commission has made clear that the requirement of 10 C.F.R. @ 2.714(b)(2)(ii) for the provision of specific reference to documents or other sources ofinformation has the effect of 3

overturning prior precedent which had previously held that 2.714 did not require a petitioner to describe facts which would be offered in support of a proposed contention. 54 Fed. Reg. at

)

33,170. The Rules of Practice now require that a petitioner include facts in support ofits position in order to demonstrate that a genuine dispute as to a material issue oflaw or fact exists.

b e

)

IsL; Oconee, CLI-99-11, EHER, slip op. at 5-6.2 As the Commission further observed, a

) contention therefore is not to be admitted "where an intervenor has no facts to suppon its l

position and where the intervenor contemplates using discovery or cross-examination as a fishmg expedition which might produce relevant supporting facts." 54 Fed. Reg. at 33,171. Thus, under

)

the amended Rules of Practice a statement "that simply alleges that some matter ought to be considered" does not provide a sufficient basis for an admissible contention. Sacramento i

( Municioal Utility District (Rancho Seco Nuclear Generating Station), LBP-93-23,38 NRC 200,

)

246 (1993), review declined, CLI-94-02,39 NRC 91 (1994).

2. Onnosition to Anolicant's Position

)

Under the Rules of Practice as amended,10 C.F.R. s 2.714(b)(2)(iii) requires a petitioner to provide:

) Sufficient information . . . to show that a genuine dispute exists with the applicant on a material issue oflaw or fact. This showing  !

must include references to the specific portions of the application  ;

(including the applicant's environmental report and safety report) that the petitioner disputes and the supporting reasons for each  ;

dispute, or, if the petitioner believes that the application fails to 1

) contain information on a relevant matter as required by law, the identification of each failure and the supporting reasons for the petitioner's belief.

l The Statement of Considerations to the 1989 amendments states this provision "will require the intervenor to read the pertinent portions of the license application, including the Safety Analysis Report and the Environmental Report, state the applicant's position and the petitioner's opposing l

) view." 54 Fed. Reg. at 33,170. If the petitioner does not believe these materials address a l

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As observed by the Commission, such a requirement is consistent with judicial decisions, such as j Connecticut Bankers Ass'n v. Board of Governors,627 F.2d 245,251 (D.C. Cir.1980).

)

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l relevant issue, the petitioner is "to explain why the application is deficient." E See also Palo O' Y.udt, CLI-91-12, EP.ta, 34 NRC at 155-56. Thus, a contention that does not directly controvert

=

1 a position taken by the applicant in the license application is subject to dismissal. See Texas Utilities Electric Co. (Comanche Peak Steam Electric Station, Unit 2), LBP-92-37,36 NRC 370, 384 (1992).  !

3. Petitioner's Mistake Cannot Be the Basis of a Contention 9 Further, a contention that mistakenly claims that the applicant failed to address a relevant issue in the application must also be dismissed. Ets, s&, Georgia Power Co. (Vogtle Electric Generating Plant, Units 1 and 2), LBP-91-21,33 NRC 419,424 (1991); Rancho Seco. LBP-93 O

23 HEB,38 NRC at 247-48 (the claim that the "EA's findings are inadequate because there is no discussion" of the licensee's decommissioning activities or the associated environmental impacts ignores that the " entire EA discusses the decommissioning activities to be performed" O by the licensee as well as the associated environmental impacts and "makes no showing that any of these matters are misstated . . . "). In such circumstances, relative to the purported lack of information or lack of analysis, "there is no material factual dispute that warrants further Q inquiry." General Public Utilities Nuclear Coro. (Oyster Creek Nuclear Generating Station),

LBP-96-23,44 NRC 143,163 (1996).

4. Challenges to Regulations Barred d

It is well established that "a licensing proceeding . . . is plainly not the proper forum for an attack on applicable statutory requirements or for challenges to the basic structure of the -

g Commission's regulatory process." Philadelohia Ele'c tric Co.,(Peach Bottom Atomic Power Station, Units 2 and 3), ALAB-216,8 AEC 13,20 (1974). Thus, a contention which collaterally attacks a Commission rule or regulation is not appropriate for litigation and must be rejected. 10 C.F.R. 2.758; Potomac Electric Power Co. (Douglas Point Nuclear Generating Station, Units 1 and 2), ALAB-218,8 AEC 79,89 (1974).

O i

i Further, a contention which " advocate [s] stricter requirements than those imposed by the O regulations" is "an impermissible collateral attack on the Commission's rules" and must be rejected. P_ublic Service Comoany of New Hamnshire (Seabrook Station, Units 1 and 2), LBP-82-106,16 NRC 1649,1656 (1982); see also Arizona Public Service Co. (Palo Verde Nuclear O _ Generating Station, Units 1,'2,'& 3), LBP-91-19,33 NRC 397,410, aff'd in oart and rev'd in oart on other arounds. CLI-91-12,34 NRC 149 (1991). Likewise, a contention that seeks to ,

litigate a generic determination established by Commission rulemaking is " barred as a matter of /

O law." Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), LBP-93-1,37 NRC 5,30 (1993); Yankee Nuclear. CL1-96-7, spIg,43 NRC at 251.

re Ver, a peWoner cannot chauenge 6e apphcant's Sture comphance & clear O

regulatory constraints without a particularized basis. A contention asserting that an applicant will violate " clear regulatory constraints" will not be admitted unless the petitioner has made a

" particularized demonstration that there is a reasonable basis to believe the applicant would act O

contrary to their explicit terms." Oyster _ Cash, LBP-96-23, supr_a,44 NRC at 164. ,

a

5. Scope of Proceeding and Materiality l 0 Licensing boards "are delegates of the Commission" and, as such, they may " exercise only those powers which the Commission has given [them]." Public Service Co. (Marble Hill Nuclear Generating Station, Units 1 and 2), ALAB-316,3 NRC 167,170-71 (1976); accord 0- Portland General Electric Co. (Trojan Nuclear Plant), ALAB-534,9 NRC 287,289-90 n.6 (1979). Accordingly, matters outside the scope of a proceeding do not provide a basis for a j cognizable contention. Marble Hill. ALAB-316, spig,3 NRC at 170-71. A contention is not )

.O cognizable unless it is material to a matter that falls within the scope of the proceeding for which the licensing board has been delegatedjurisdiction as set forth in the Commission's Notice of l Opportunity for Hearing. E; see also Commonwealth Edison Co. (Zion Station, Units 1 and 2),

O ALAB-616,12 NRC 419,426-27 (1980); Commonwealth Edison Co. (Carroll County Site),

0:

1 i

ALAB-601,12 NRC 18,24 (1980). The Notice of Opportunity for Hearing in this case 3

delineates the scope of the present licensing proceeding as follows:

The [NRC] is considering issuance of an amendment to Facility ,

Operating License No. NPF-63 issued to Carolina Power &

Light . . . for operation of the Shearon Harris Nuclear Power ,

3_ Plant. . . . The proposed amendment would support a modification i to the plant to increase the spent fuel storage capacity by adding rack modules to spent fuel pools (SFPs) "C" and "D" and placing the pools in service.

) 64 Fed. Reg. 2237,2238 (1999). Thus, an admissible contention here is limited to the very ,

narrow set ofincremental health and safety or environmental impacts resulting from the

' activation and use of already-installed spent fuel storage at the Shearon Harris Nuclear Power 3'

Plant (" Harris plant" or " Harris" or "HNP") in spent fuel pools C and D, as requested in CP&L's license amendment application.

)

The specific issues oflaw or fact raised or controverted by a contention must be material to the granting or denial of the license amendment at issue. This general limitation on the admission of contentions is expressly provided for by the 1989 amendments to 10 C.F.R. Q 2.714

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and is implicit in NRC precedent prior to the 1989 amendments. In the Statement of Considerations to the 1989 amendments, the Commission defined a " material" issue as meaning that the " resolution of the dispute would make a difference in the outcome of the licensing

)

proceeding." 54 Fed. Reg. at 33,172 (emphasis added). Thus, immaterial issues are subject to dismissal under 10 C.F.R. Q 2.714(d)(2)(ii) because, even if proven, they "would not entitle [the) petitioner to relief." See also Notice of Consideration, spa,64 Fed. Reg. at 2240 (1999).

The requirement that contentions raise issues material to the granting or denial of the license subject of the licensing proceeding ensures that contentions have concrete application to j the facility in question and precludes the litigation of generalized claims unrelated to the facility.

Sit,9A, Peach Bottom, ALAB-216, sata,8 AEC at 21, n.33 ("if someone wants to advance

)

r-

l e l generalizations regarding his particular views of what applicable policies ought to be, a role

1. other than as a party to a trial-type hearing should be chosen"), auctine Duke Power Co.

(William B. McGuire Nuclear Station, Units 1 & 2), ALAB-128,6 AEC 399,401 (1973);

accord. Oconee, CLI-99-11, sup_r_a, slip op, at 6.

1. 6. Health and Safety Significance

For a contention raising non-environmental issues to be material, it must assert a

, significant health and safety concem with respect to the license application. The contention "must either allege with particularity that an applicant is not complying with a specified [ safety]

regulation, or allege with particularity the existence and detail of a substantial safety issue . . . ."

, Seabrook. LBP-82-106, suora,16 NRC at 1656 (footnote omitted); accord Duke Power Co.

(Catawba Nuclear Station, Units 1 and 2), LBP-82-116,16 NRC 1937,1946-1947 (1982). For example, contentions conceming alleged deficiencies in a decommissioning plan must not only allege and provide sufficient bases to show the deficiencies, but also show that the purported deficiencies have "some independent health and safety significance" such that reasonable assurance of the public health and safety with respect to decommissioning is no longer assured.

Yankee Nuclear, LBP-96-2, suora,43 NRC at 75; see also Yankee Nuclear. CLI-96-7, supj:a,43 NRC at 258 (" Petitioners must show some specific, tangible link between the alleged errors in the plan and the health and safety impacts they invoke").

, C. The Scope of a Contention Is Limited by Its Specific Bases Certain of the contentions filed by Petitioner allege a general inadequacy in the license amendment application (eg, Contention 3: " Inadequate Quality Assurance") followed by 3 specific assertions in the basis as to the manner in which the application is allegedly deficient. It is well established under Commission precedent that the scope of a contention is determined by its literal terms, coupled with its stated bases. See, eg, Public Service Comoany of New D

7 0;

Hamnshire (Seabrook Station, Units 1 and 2), ALAB-899,28 NRC 93,97 (1988). In that case, M 'in assessing the scope of the intervenor's contention, the Appeal Board stated that:

The reach of a contention necessarily hinees uoon its terms couoled with its stated bases. . . . [0]ne purpose of the requirement i' in [{]2.714(b) that the bases of a contention be set forth with

.O reasonable specificity is to put the other parties on notice as to .

what issues they will have to defend against or oppose. Thus,

]

-_where a question arises as to the admissibility of a contention, we l look to both the contention and its stated bases. . . . [W]here the  !

issue is the scope of a contention, there is no good reason not to I

O construe the contention and its bases together in order to get a sense of what precise issue the party seeks to raise.

E at 97 (emphasis added)(citations omitted).

O Similarly in Illinois Power Co. (Clinton Power Station, Unit 1), LBP-81-61,14 NRC 1735,1737 (1982), the Licensing Board held that contentions must be narrowed to fit their stated bases. ' In analyzing the admissibility of contentions, " making broad allegations plus specific

. allegations that provide the bases for the broad range," the Board ruled that Where a contention is made up of a general allegation which, standing alone, would not be admissible under 10 C.F.R.

O {2.714(b), plus one or more alleged bases for the contention set forth with reasonable specificity, the matters in controversy raised by such contention are limited in scope by the soecific alleged basis or bases set forth in the contention.

.O E at 1736-37 (emphasis added). Accord, Cleveland Electric Illuminatino Co. (Perry Nuclear Power Plant, Units 1 and 2), LBP-81-35,14 NRC 682,685-86 (1981).

O Thus, the scope of a broadly worded contention is limited by the specific assertions made in its bases. Accordingly, in analyzing the admissibility of the Petitioner's contentions, the Applicant has proposed that certain contentions be restated to incorporate the specific allegations O fr m their bases. This serves to focus the analysis whether each contention is admissible and, in g

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[

l the event the contention were admitted, to better define the precise issues to be litigated within

_ the scope of the contention.

III. GROUP I: TECHNICAL CONTENTIONS Petitioner has advanced three technical contentions, alleging (1) inadequate emergency core cooling and residual heat removal; (2) inadequate criticality prevention; and (3) inadequate quality assurance. None of the technical contentions should be admitted for the reasons summarized below.

Eitst, Petitioner has demonstrated a misunderstanding of the Harris Final Safety Analysis Report ("FSAR") and the license amendment application. Indeed, some of the information k discussed regarding the Harris FSAR was incorrect information from outdated versions of the L Harris FSAR. Thus, the source of a number of concerns raised by the Petitioner results from misunderstandings or mistakes. Sesond, Petitioner has claimed that certain analyses have not

) . been performed or inspections not planned, where in fact the opposite is true. Third, in a number ofinstances the scope of the contention was much broader than the basis offered by Petitioner in support. Closer inspection shows that the basis does not actually support the contention in any event. Eounh, often the purported basis for a contention lacked specificity. Eifth, Petitioner has raised issues outside the scope of this proceeding. S_i_th,

.x a subpart of one contention is no more than a unparticularized and unsupported assertion that Applicant will fail to comply with a clear h regulatory requirement. Seventh, in some instances Petitioner attempted to support a contention l

l with Applicant's own statements in the license amendment application. There is no genuine issue in dispute with Applicant's own statements. Eighth, both of the bases for Contention 2 are h impermissible collateral attacks on the Commission's rules. Finally, in Contention 3, Petitioner has not shown any nexus between the alleged inadequacies in the 50.55a Alternative Plan and public health and safety.

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While Applicant's response to Petitioner's technical contentions is somewhat lengthy, it 3 is necessary to explain why, for example, Petitioner is mistaken, or that there is no statement of basis with specificity for the proffered contentions, or that there are no health and safety impacts raised by the proposed contentions. l D

We address Contentions 1,2 and 3 in the remainder of this section.

1 A. Contention 1: Inadequate Emergency Core Cooling and Residual l Heat Removal D .

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1. The Contention i

BCOC Contention 1 asserts the following:

In order to cool spent fuel storage pools C and D, CP&L proposes 3 to rely on the Unit 1 Component Cooling Water ("CCW") system, coupled with administrative measures to ensure that the heat load from the pools does not overtax the CCW system. CP&L's reliance on the Unit 1 CCW system and administrative measures  !

for cooling spent fuel storage pools C and D will unduly l D compromise the effectiveness of the residual heat removal l

("RHR") system and the Emergency Core Cooling System

("ECCS") for the Shearon Harris plant, such that the plant will not l comply with Criteria 34 and 35 of Appendix A to 10 C.F.R. Part '

l 50. q 3

BCOC Supp. Pet. at 4. BCOC begins its statement of basis by quoting General Design Criteria 1

34 and 35 and summarizing a number of statements made by CP&L in its license amendment application.3 Isla at 4-7. BCOC then identifies six specific issues that form the basis for its 3

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j For example, BCOC notes that CP&L has stated: some CCW system flow will be sent to the pool C&D heat exchangers many hours into a LOCA event (citing Applicant's license amendment application ("Lic. Amend. App."), Encl. 9); the CCW system has adequate margin to accommodate pools C and D (jsL); a technical specification will be added to limit the heat load in pools C and D to 1.0 MBTU/ hour (isL, Encl. 5). Needless to say, these citations to CP&L's own statements in its license amendment application do not establish a genuine dispute with CP&L, and therefore do not provide a basis for Contention 1. S_ee BCOC Supp. Pet. at 4-7.

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I l contention. To facilitate the determination of admissibility of this contention, the Applicant has

) summarized BCOC's six bases asserted for Contention 1 as follows:d f Basis 1 - Even without the amendment to add pools C and D, the Harris FSAR shows that the CCW system is incapable of i l accommodating the heat load from the recirculation phase of a 3 design-basis LOCA; ,

Basis 2 - The analysis of CCW margin supporting the license amendment application does not address the time dependence of the CCW system heat load during a design-basis LOCA; D

Basis 3 - The analysis of CCW margin supporting the license i amendment application does not address degradation of CCW and RHR heat exchanger performance due to heat exchanger fouling and plugging;

] Basis 4 - The license amendment application does not address the potential for failure to comply with the administrative measure

limiting the heat load in pools C and D to 1.0 MBTU/ hour; i

i l Basis 5 - The license amendment application does not address the h potential for increased operator error in diverting CCW system l flow to meet the cooling needs of pools C and D during a LOCA event;

Basis 6 - The analysis supporting the license amendment i

application does not address the ability of Unit 1 electrical systems h to meet the needs of pools C and D while also supporting essential safety functions.

BCOC Supp. Pet. at 7-9. BCOC closes its statement of basis with a discussion of different equipment approaches for cooling pools C and D that BCOC would have liked CP&L to have pursued.5 Sn & at 9-10. BCOC notes that "CP&L has not proceeded with this option." him at

) '

Su { II.C., sgg.

8 For example, the Petitioner would have preferred that CP&L provide cooling for pools C and D by

"[c]onstruction of an independent cooling system for pools C and D, supported by dedicated

! emergency diesel generators." E at 9. The Petitioner also notes that a " future upgrade of the CCW system . . . contemplated [by CP&L] is not described in the present license amendment application"

) and that "CP&L has made no commitment to undertake [this] upgrade." E at 9-10.

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9. BCOC's statement of design attematives that the Applicant has not pursued is beyond the O

scope of this proceeding and does not provide a basis for an admissible contention. S_eg e Q II.B.5., i E!PR

2. Applicant's Response to the Contention O

Contention 1 should be rejected in its entirety because each of the six bases asserted by the Petitioner fail to provide the support required for an admissible contention within the O

Commission's regulations. In general, BCOC's Contention 1 is based on a misreading and mistaken understanding of the facts supporting the Harris FSAR and license amendment a

application and thus must be dismissed. See s II.B.3., supn. Each of the six bases asserted by  !

O l BCOC are addressed, in turn, below. I 1

Basis 1 - Even Without the Amendment to Add Pools C and D, the Harris FSAR Shows that the CCW System is Incapable of Accommodating the Heat Load From the Recirculation Phase O of a Design-Basis LOCA l

Basis 1 for this contention must be rejected both because it is beyond the scope of this

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proceeding and because it is supported by a mistaken understanding of the facts in the Harris O

FSAR.

Basis 1 must first be rejected because it is beyond the scope of this proceeding. BCOC's O statement to support Basis 1 alleges that " design information in the Final Safety Analysis Report l l

('FSAR') for the Harris plant suggests that accommodating a design-basis LOCA will alread_y i exploit the margin of the CCW system, without any additional load from nools C and D."

O BCOC Supp. Pet. at 7 (emphasis added). This statement demonstrates that this basis is challenging the already existing and approved accident analyses for the Harris plant, before the license amendment request in this proceeding to activate pools C and D. A contention which j O. attempts to raise issues outside of the scope of the current proceeding, as set forth in the Commission's Notice of Opportunity for Hearing, must be rejected by the Board. Sgg Q II.B.5.,

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O EDIg The Commission defined the scope of this proceeding in the Notice of Opportunity for j O' Hearing, which states that:

The [NRC) is considering issuance of an amendment . . . to  ;

increase the spent fuel storage capacity by adding rack modules to spent fuel pools (SFPs) 'C' and 'D' and placing the pools in O service.

64 Fed. Reg. at 2237-38. This proceeding therefore addresses the addition of spent fuel pools C and D to HNP. It does not address the existing status of the Harris plant prior to the addition of  !

pools C and D. Thus, the scope defined by the Commission for this proceeding does p_qt include

' the validity of the current licensinn basis of the Harris plant (11, prior to the addition of pools C and D). Because Basis 1 raises issues outside the scope of this proceeding, it must be rejected by the Board.

Even if Basis I were within the scope of this proceeding, it must also be rejected because O it is supported by a mistaken understanding of the facts in the Harris FSAR. In support of Basis 1, the Petitioner states that:

The CCW system has two heat exchangers, each with a design heat transfer rate of 50 million BTU / hour. During the recirculation O phase of a design-basis LOCA, the estimated maximum load on the CCW system is 160 million BTU / hour.

BCOC Supp. Pet. at 7 (footnotes omitted). From this, the Petitioner concludes that M0 0 MBTU/ hour " exceeds the heat transfer rate of 100 million BTU / hour" provided by the two CCW heat exchangers, and therefore the CCW system, even prior to the activation of pools C and D, is incapable of handling a design-basis LOCA. Ecs li(emphasis added). However, the O Petitioner's statements are based on a mistaken and outdated understating of the Harris FSAR.

BCOC has confused the capability of the CCW system to reiect heat with the heat load int the CCW system. BCOC characterizes 160 MBTU/ hour as the " estimated maximum heat O

load 2n the CCW system." li(emphasis added). BCOC cites this number from " Table 9.2.1-3, A~

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Amendment No.15" of the Harris FSAR. E at note 12. The table cited by BCOC, however, is from' the Service Water System ("SWS") section of the FSAR and establishes the maximum design capability of SWS to accept heat loads during accident conditions from a variety ofinput l sources, one system of which is the CCW. Sss Harris FSAR at 9.2.1-5 to 11 (SWS Safety Evaluation). Furthermore, the number offered by BCOC, M0 MBTU/ hour cited from  :

i Amendment No. M, is considerably out of date. The current version of the Harris FSAR is l Amendment .41. Pursuant to Amendment 48, the maximum capability of the SWS to handle heat

)- loads rejected from'the CCW system heat exchangers is 2216 MBTU/ hour. See Harris FSAR, Table 9.2.1-3 (" Maximum Service Water System Heat Loads Following LOCA," Amendment No. 48). Thus, the capability to reject heat out of the CCW system during the recirculation phase h of a design-basis LOCA is 272.6 MBTU/ hour. To ensure there is sufficient margin, the FSAR must show that the heat load into the CCW system is less than 272.6 MBTU/ hour.

Following a design-basis LOCA, the only heat loads on the CCW system are heat outputs l from the RHR heat exchanger and the RHR pump; every other load on the CCW system is on the l "non-essential header," and is isolated when the RHR system is initiated.6 Sag Harris FSAR at 9.1.3-5 to 6. The Harris FS.AR defimes the maximum heat output from the RHR system (and therefore heat input to the CCW system) following a LOCA to be 222.2 MBTU/ hour. Sset Harris FSAR at 9.2.1-7 to 8; Table 9.2.1-11 (RHR heat exchanger heat rejection rate).7 Since 2216

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Isolation of the CCW non-essential header is performed by operating four switches in the Harris control room. Sfe Harris FSAR at 6.3.2-13d.

l The "50 million BTU / hour" number cited by BCOC for heat input to the CCW system is the " design heat transfer" rate for CCW during normal plant operation. Ssg Harris FSAR, Table 9.2.2-1. This l number represents normal olant operation. rather than accident conditions. and is based on a CCW outlet temperature of 105 F during normal plant operation. E at 9.2.2-1 and Table 9.2.2-1. Normal plant operation is D9.t the correct heat load assumption for analyzing a LOCA event, which is an accident condition. During cooldown following a LOCA event, the CCW system operates under different temperature conditions, including a CCW outlet temperature of 120 F. E at 9.2.2-1. The actual maximum heat load from RHR to CCW following a LOCA event is 222.2 MBTU/ hour from both RHR trains (111.1 MBTU/ hour from each of two RHR trains). E at 9.2.1-7 to 8 and Table

) 9.2.1 11.

r

O MBTU/ hour, the ability of CCW to Isigra heat to SWS is greater than 222 2 MBTU/ hour, the j heat innut to CCW from RHR, it is clear that the CCW system can handle the RHR system load 1

- during the recirculation phase of a design-basis LOCA. BCOC's assertions supporting Basis 1 l

are based on a mistaken and outdated understanding of the Harris FSAR. A petitioner's

,O mistaken understanding of the facts regarding an application do not provide a basis for an -

admissible contention. Sgg { II.B.3., EUKR Therefore, even if Petitioner's Basis I were within the scope of this proceeding, it must still be rejected for lack of basis. ,

'O Basis 2 - The Analysis of CCW Margin Supporting the License Amendment Application Does Not Address the Time l Dependence of the CCW System Heat Load During a Design-Basis LOCA O

In Basis 2, BCOC alleges that the Applicant has failed to take into account the time dependence of the CCW system heat load during a design-basis LOCA. BCOC Supp. Pet. at 7.

Since "the heat load on the CCW system from the RHR system will change over time," BCOC states that:

Analysis must demonstrate that the CCW system has sufficient margin to accommodate both the RHR system and fuel pool heat 1 ads over time, during the LOCA event and subsequently.

O Ld, In fact, the Applicant's analysis supporting the license amendment application does demonstrate that the CCW system has sufficient margin to accommodate the RHR system and

O spent fuel pool heat loads over time. Applicant agrees with BCOC that the heat load on the CCW system changes over time following a LOCA, and that the analysis must take this into account. These facts do not raise a genuine dispute with the Applicant. 4 l0 Enclosure 9 of Applicant's license amendment application states that:

[a] new thermal-hydraulic cnalysis was performed to evaluate the

, 1.0 MB[TU]/hr heat load that would be added to spent fuel pools LO C' and 'D' as a result of this [ amendment].

l. 1 I

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LO'- I 1 1 Lic. Amend. App., Encl. 9 at 2. Based on this analysis, the license amendment application l O concludes that "the CCW system has adequate thermal-hydraulic capacity" to meet the additional l

heat loads from spent fuel pools C and D. E at 2-3. Enclosure 9 addresses the time dependent nature of the heat loads on the CCW system following a LOCA event, indicating that the O . thermal-hydraulic analysis addressed the significant post-LOCA heat loads as a function of time, including "the beginning of the sump recirculation phase," E at 3, and "[t]he addition of spent fuel pools 'C' and 'D' to the CCW system," E at 4. While copies of the detailed calculations O

! supporting a license amendment application are not required to be included in the application itself, the Applicant has attached the CCW Calculation as Exhibit 1 to this pleading to aid in l clarifying the Petitioner's concems.:

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! CP&L's analysis supporting this license amendment accounts for the time dependence of the CCW system heat load during a design-basis LOCA, and demonstrates that the CCW system has sufficient margin to accommodate both the RHR system and spent fuel pool heat loads over O

time. The CCW Calculation evaluates the capability of the CCW system during each significant phase of post-LOCA operation to ensure the CCW system has adequate margin to handle both the RHR system load and the spent fuel pool cooling load, specifically including the additional O

1.0 MBTU/ hour from pools C and D. CCW Calculation at 1 (Section 1.0," Purpose"). There are three significant phases to evaluate in order to demonstrate the post-LOCA adequacy of the CCW system:

1. LOCA: Safety Injection Phase;'

'O Sgs CP&L SF-0040, Soent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis.

Shearon Harris Nuclear Power Plant (Rev. O,1998)("CCW Calculation")(attached as Exhibit 1).

During this initial phase of post-LOCA operation, the RHR heat exchangers are not yet operating 0: and the corresponding load on CCW is minimal. CCW Calculation at 3.

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2. LOCA: Recirculation Phase, Containment Sump Recirculation with CCW Nonessential HeaderIsolated;'

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3. LOCA: Recirculation Phase, Containment Sump Recirculation with Limited Fuel Pool Cooling."

The CCW Calculation demonstrates that the CCW system has adequate capacity to handle the maximum heat loads during each of the three major phases of post-LOCA operation.12 _S f e 11, Tables 7(g) (LOCA-Safety Injection), 7(h) (LOCA-Recirc., RHR Only), and 7(i) (LOCA-Recirc., RHR/SFP). Specifically for Phase 3, in which the spent fuel pool heat load (including

. pools C and D) is added to the CCW system, the analysis demonstrates that the maximum RHR heat load at the time the spent fuel pools are added is down to 10JJ 0 MBTU/ hour, which yields more than enough margin for the CCW system to accommodate the additional heat load ofifil MBTU/ hour from the spent fuel pools (112 MBTU/ hour from pools A and B, .10 m MBTU/ hour from pools C and D)." li at 14. Each of the three major CCW post-LOCA operating conditions was analyzed by CP&L in reaching its conclusion that:

The analysis demonstrates that adequate margin exists during all normal and accident modes of system operation and that the CCW system has adequate thermal-hydraulic capacity to provide the minimum flow required by the fuel pool heat exchangers after the O activation of Pools 'C' and 'D'.

3 During this second phase of post-LOCA operation, the RHR heat exchanges are activated to perform core cooling, and the CCW non-essential header (which includes the spent fuel pools) is isolated.

The maximum post-LOCA heat load on RHR occurs when recirculation is first initiated since core

.O heat load continually decreases as a function of time following shutdown. CCW Calculation at 7.

During this third phase of post-LOCA operation, in addition to the RHR heat load, some CCW flow

- is diverted to the spent fuel pool heat exchangers. This occurs several hours after the LOCA when the core, and hence RHR, heat load has dropped. No other loads are added to the CCW system until the LOCA event has terminated. CCW Calculation at 9.

O l2 Of course, as discussed for Basis 1, the ability of CCW to handle post-LOCA heat load prior to the addition of spent fuel pools C an D had already been demonstrated.

Recall from the discussion of Basis 1 that the CCW system has a total heat rejection capability of )

272.6 MBTU/ hour. Even one of the two CCW loops (136.3 MBTU/ hour) is capable ofhandling i both the maximum RHR load at the time pool cooling is restarted (80.53 MBTU/ hour) and the }

O maximum heat load from the pools (16.2 MBTU/ hour).

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f-Lic. Amend. App., Encl. 9 at 2-3.

O Thus, the Applicant's analysis supporting the license amendment did, in fact, take into account the time dependence of the CCW system heat load following a LOCA event, and demonstrated that the CCW system has adequate capacity to handle both the RHR load and the spent fuel pool load following a LOCA. BCOC's assertion that the Applicant has failed to take into account the time dependence of the CCW systr. heat load during a design-basis LOCA is 3 mistaken. A mistaken claim that the applicant failed to address a relevant issue in an application must be dismissed. See & II.B.3., spa. BCOC's Basis 2 must be rejected because it lacks the basis for an admissible contention and fails to establish a genuine factual dispute warranting g further inquiry.

Basis 3 - The Analysis of CCW Margin Supporting the License Amendment Application Does Not Address Degradation of CCW and RHR Heat Exchanger Performance Due To Heat

' Exchanger Fouling and Plugging

?

In Basis 3, BCOC alleges that the Applicant has failed to:

address the sensitivity of CCW and RHR system performance to factors that may degrade performance from nominal levels.

O Relevant factors include heat exchanger fouling and plugging.

BCOC Supp. Pet. at 8. BCOC acknowledges that "CP&L itself has previously recognized that

[the CCW system analysis should] include fouling factors and tube plugging limits," citing 3

viewgraphs shown by CP&L to the NRC during a public meeting on this license amendment in March 1998. R Yet, BCOC mistakenly assumes that "CP&L fails to address this issue in Enclosure 9." E i

As discussed above, the CCW Calculation is addressed in Enclosure 9 of the license amendment application and forms the basis for CP&L's conclusion that:

adequate margin exists during all normal and accident modes of system operation and that the CCW system has adequate thermal-j I l D

' hydraulic capacity to provide the minimum flow required by the ,

fuel pool heat exchangers after the activation of Pools 'C' and 'D'. i 3

Lic. Amend. App., Encl. 9 at 2-3. In reference to the CCW Calculation, Enclosure 9 states that ,

"[i]n support of this design change package, a thermal-hydraulic model was created to analyze D the overall impact of this additional heat load." E at 1. The Applicant stated in its handout at the March 3,1998 public meeting, cited in BCOC's contention, that analysis of the CCW system would include, intgr alia, " fouling factors [and] tube plugging limits."" Thus, the Applicant 3 made it clear that the " thermal-hydraulic model [ discussed in Enclosure 9 of the license amendment application] . . . created to analyze the overall impact of this additional heat load" would include " fouling factors [and] tube plugging limits." Ses Lic. Amend. App., Encl. 9 at 1; j NRC Meeting Summary, Encl. 2 at 8.

The CCW Calculation shows that these factors are, in fact, included in the analysis. - The CCW Calculation clearly states that:

3 All heat exchanger thermal models use design fouling factors . . .

to ensure that design basis conditions can be met even with extreme fouline conditions. ,.

D CCW Calculation at 3 (emphasis added); ggs alsg isl a at 5 (section 4.1.2). For example, the CCW heat exchanger, which rejects heat to SWS, uses a design fouling factor that "significantly (50.4 percent) exceeds the current worst case trended tubeside fouling factor" and assumes no I additional tube plugging because any additional assumption on heat exchanger fouling and plugging would lead to results that are " overly conservative, given the excessive design fouling l factor." E BCOC's belief that CP&L "recoenized that exploitation of the margin in the CCW

) system could involve . . . foulino factors and tube oluooina limits," but then somehow failed to CP&L's viewgraph outlined options for providing cooling for pools C and D. The option implemented in this license amendment explicitly included consideration of fouling factors and tube plugging. Sag NRC Meeting Summary, Encl. 2 at 8 (PDR Accession # 9803200255)(Mar. I1, j 1998).

1 O i 1

1 include these factors in its thermal-hydraulic analysis, is mistaken. The Applicant's public O statement that its analysis will address heat exchanger fouling and plugging is fully consistent with its CCW Calculation that does address heat exchanger fouling and plugging. BCOC's Basis 3 must be rejected because for lack of sufficient basis and failure to establish a genuine factual '

n o

dispute warranting further inquiry. .S_ee e II.B.3., suora.

Basis 4 -The License Amendment Application Does Not Address the Potential for Failure to Comply with the Administrative Measure Limiting the Heat Load in Pools C O and D to 1.0 MBTU/ hour In Basis 4, BCOC states that CP&L has failed to address:

the potential for failure of administrative measures, such that the O heat load in pools C and D will exceed 1.0 million BTU /hnur.

BCOC Supp. Pet. at 8.

O Eirs, Basis 4 must be rejected because it lacks sufficient basis with specificity for an admissible contention. BCOC provides only three sentences in support of Basis 4. The Petitioner first generally alleges that such " administrative measures" "could be exceeded as a O result of human errors," and requests that "such errors . . . be carefully considered." Id2 However, the Petitioner fails to identify any specific " administrative measures" about which it is concerned, and fails to identify any specific " human errors" to consider. See id. Under the O Commission's regulations, a contention "that simply alleges that some matter ought to be considered" does not provide a sufficient basis for an admissible contention. Rancho Seco. LBP-93-23, suora,38 NRC at 246; see also II.B.l., suora. BCOC's general assertion regarding O failure of" administrative measures" due to " human errors" should be rejected by the Board due to its failure to provide sufficient basis for an admissible contention.

1 Second, Basis 4 must be rejected because, to the extent it challenges the Applicant's compliance with a Technical Specification, it must be rejected for lack of basis. The Applicant's l

-22 D

D' license amendment includes the addition of Technical Specification 5.6.3.d to the Harris

) operating license, which requires that "[t]he heat load from fuel stored in Pools 'C' and 'D' shall not exceed 1.0 MBtu/hr." Lic. Amend. App., Encl. 4 at 3. Pursuant to Commission regulations, Technical Specifications are reauired to be included in the " license authorizing operation" of a 3 power reactor issued by the Commission.10 C.F.R. Q 50.36(b). Compliance with the terms of an operating license is required by the NRC. Therefore, clear regulatory constraints mandate that CP&L must keep the heat load in pools C and D from exceeding 1.0 MBtu/hr. A contention

) asserting that an applicant will violate clear regulatory constraints must be rejected unless the petitioner has made some " particularized demonstration that there is a reasonable basis to_ believe

[the applicant] would act contrary to their explicit terms." Oyster Creek, LBP-96-23, s.up.ta,44 u

) NRC at 164; sge { II.B.4., supra. BCOC has provided no basis upon which to believe the Applicant will violate the Technical Specifications in the Harris operating license. Basis 4 must be rejected for failure to provide sufficient basis to establish a material dispute warranting further 3 l mqtry.

Basis 5 - The License Amendment Application Does Not Address the Potential for Increased Operator Error In Diverting CCW System Flow to Meet the Cooling Needs of D Pools C and D During a LOCA Event 1

In Basis 5, BCOC states that CP&L has failed to address:

the potential for increased operator error associated with the need 3 for the CCW system to meet the cooling loads of pools C and D while also serving other essential safety functions.

BCOC Supp. Pet. at 8. BCOC generally asserts that "[t]he operators' burden of observation, 3 decision-making and action would be increased by the use of the CCW system to cool pools C and D." Id, Specifically, BCOC alleges that:

The potential for operator error . . . during a LOCA event . . .

K would be further increased if, during this event, the operators were required to divert some CCW system flow from the RHR heat exchangers in order to meet the cooling needs of pools C and D.

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o M at 8-9. BCOC provides no other facts or discussion to support Basis 5.

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Basis 5 must be rejected for failure to provide sufficient basis with specificity to establish an admissible contention. In its four-sentence statement for Basis 5, the Petitioner fails to l

identify any specific operator errors that would occur or any particularized reason to believe that

!. the CP&L operators would fail to accomplish their actions in compliance with the Commission's i

regulations. Sg & at 8-9. The Commission's regulations establish a comprehensive set of regulations for power reactor operator training and licensing 10 C.F.R. Q 50.120; 10 C.F.R. Part l 55 (" Operator's Licenses"). BCOC provides no reasonable basis to believe that operators trained and licensed pursuant to the Commission's comprehensive regulatory scheme would be I

incapable of providing cooling flow to spent fuel pools C and D following a LOCA event.

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l Moreover, a review of the operator actions in question further demonstrates that Basis 5 fails to establish any material factual dispute warranting further inquiry. As discussed sp.g, h when the RHR system is initiated following a LOCA event, the CCW system non-essential header, which includes the spent fuel pool cooling systems, is isolated by the operators from the control room. Sg Harris Plant Operating Manual, Emergency Operating Procedure EOP-EPP-

) 010 at 12 (steps 8.a. and 8.b.; shut CCW non-essential supply and return valves) ("EOP Operating Manual")." This operation isolates the entire CCW non-essential header, including j pools A and B and pools C and D, in a single step. Because this operator action is already included in the current Harris plant licensing basis, there is no incremental action added by pools C and D, and any challenge to this action is beyond the scope of this proceeding. The spent fuel pool cooling loads are added back to the CCW system 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the RHR system is l Relevant excerpts from the Harris Plant Operating Manual including both the EOP Operating Manual and the OP Operating Manual, as cited in this response, are attached to this pleading as y Exhibit 2.

1 I

O initiated.i6 EOP Operating Manual at 20; Harris Plant Operating Manual, Operating Procedure OP-145 at 30 ("OP Operating Manual") (Exhibit 2); ge also Harris FSAR at 9.1.3-6. Operator actions to add the spent fuel pool heat exchangers back to the CCW system first require operators to locally shut the isolation valves for all of the other (non-spent fuel pool) loads on the CCW 4

non-essential header. OP Operating Manual at 30-31. However, since this operation is already performed for spent fuel pools A and B under the current licensing basis, and no additional action would be required following the addition of spent fuel pools C and D, this operator action is also beyond the scope of this proceeding. In fact, the only additional operator action required "to meet the cooling needs of pools C and D" is for the operators in the control room to start the pool C and D spent fuel pool cooling system pumps (which requires turning two switches in the control room). Seg .e is,d at 31. The Petitioner provides no reasonable basis upon which to believe that turning two switches in the control room over five hours after a LOCA event would be a

" stressful event" for operators that would result in " operator error." See BCOC Supp. Pet. at 8.

Furthermore, because the maximum heat load of pools C and D is only about 7% that of pools A and B," pools C and D would only heat up a small fraction as fast as would pools A and B, and would take about 85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> to heat pools C and D up to the administrative limit of 137 F.'8 The

• Petitioner provides no reasonable basis to believe that, even if an operator error were to occur, it would not be readily corrected by the operators during the three days following a LOCA event.

p 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is the approximate time required to heat pools A and B, assuming their maximum heat load of 15.2 MBTU/ hour, up to a temperature of 137'F, the administrative limit for the pools. It is estimated to require an additional 2.97 hours0.00112 days <br />0.0269 hours <br />1.603836e-4 weeks <br />3.69085e-5 months <br /> to heat pools A and B up to 150 F. Note that both of p "

these administrative limits are far below the boiling temperature of water,212 F.

The maximum pool C and D heat load (1.0 MBTU/ hour) is about 7% of the maximum heat load for pools A and B (15.2 MBTU/ hour).

Pools A and B, at 15.2 MBTU/hr, require 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to heat up to 137 F; pools C and D, at 1.0 MBTU/ hour, would require about 85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> (15.2/1.0 times 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) to heat up to the same

^) temperature.

3

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l BCOC has failed to provide any support for its generalized assertion that the requirement to provide CCW system flow to pools C and D following a LOCA event would lead to operator errors and has failed to identify what operator errors would occur and why they are material.

Basis 5 must be rejected for failure to establish a valid basis for an admissible contention.

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Basis 6 - The Analysis Supporting the License Amendment Application Does Not Address the Ability of Unit 1 Electrical Systems to Meet the Needs of Pools C and D While Also Supporeing Essential Safety Functions

)

In Basis 6, BCOC alleges that the Applicant failed to address the ability of on-site and off-site electrical power to support the additional electrical load of spent fuel pools C and D, in addition to meeting the electrical loads of other essential safety functions at the Harris plant.

DCOC Supp. Pet. at 9. Specifically, BCOC states that an analysis must be performed:

to indicate that the available margin in the Unit 1 electrical systems, with or without offsite power, is adequate to meet the

) needs of pools C and D while also supporting residual heat removal, emergency core cooling and other essential safety functions, li

)

BCOC's assertion that the Applicant has not analyzed the ability of on-site and off-site electrical power systems at HNP to handle the additional load from pools C and D is mistaken.

3 The Harris FSAR evaluates the capability of the plant's emergency diesel generators to handle safety-related electrical loads following a loss of off-site power (" LOOP"). See Harris FSAR, Tables 8.3.1-2a and 2b. The Harris plant has two emergency diesel generators, EDG-A and

)

EDG-B. Each EDG has a continuous rating of 6500 kW, and an overload rating of 7150 kW for l two hours in any 24-hour period. Harris FSAR at 8.3.1-58. The FSAR evaluates electrical loads for both the LOOP and LOCA/ LOOP scenarios, with the purpose of"demonstrat[ing] that

) continuous loading is within the continuous rating of the emergency diesel generator." Harris FSAR at 8.3.1-58. The electrical loads on the EDGs are analyzed in CP&L Calculation E-(

y i

< 6000. Harris FSAR at 8.3.1-58. The tables in the current version of the Harris FSAR,  !

b Amendment 48, show the available margin of the two EDGs prior to the addition of pools C and D. EDG-A shows a margin of 243 kW and EDG-B shows a margin of 321 kW, both with respect to the continuous loading rating of 6500 kW. Harris FSAR at 8.3.1-54 to 55. The

) addition of pools C and D adds an additional load of one 150-hp pump to each EDG.2 The 150-hp pump load translates into an electrical load of 125-kW.2 Therefore, the FSAR demonstrates sufficient margin for both EDG-A and EDG-B to handle the additional 125-kW electrical load

) required to place spent fuel pools C and D in service.

A comprehensive evaluation of HNP electrical loads confirmed that, as the FSAR indicates, the EDGs do have sufficient margin to accommodate the additional electrical loads

)

from pools C and D. As the FSAR notes, there are " outstanding change documents posted . . .

against Calculation E-6000." Harris FSAR at 8.3.1-58. Therefore, Calculation E-6000 was revised to incorporate all " outstanding change documents" as well as the additional 125-kW

)

loads from adding pools C and D. Calculation E-6000 at ii. Table 6 of Calculation E-6000 shows that, even after the addition of pools C and D electrical loads, the two EDGs have l 22 remaining margin; EDG-1 has a remaining margin of 182.1 kW and EDG-B has a remaining

)

margin of 254.5 kW.23 Lda , Tables 6 and 7. Calculation E-6000 also evaluates the capability of off-site power to accommodate HNP safety-related loads, including spent fuel pools C and D. l

?

The relevant pages of Calculation E-6000 (Rev. 6,1999) are attached as Exhibit 3.

2o There are two independent spent fuel pool cooling system pumps for pools C and D, each rated at 150 hp. Each of the pumps is connected to a separate EDG train to ensure that the loss of a single '

EDG will not eliminate cooling of pools C and D.

2:-

A 150-hp mechanical pump converts to an electrical load of 125-kW (unit conversion of 1 hp =

0.746kW and the standard conversion factor of 0.90). There are no other safety-related (and hence EDG) electrical loads associated with pools C and D.

22 6500 kW continuous rating minus 6317.9 kW maximum load.

y 23 6500 kW continuous rating minus 6245.5 kW maximum load.

.O l

See ida at 3-4. Based on its comprehensive evaluation of all Harris safety-related electrical loads O

(including electrical loads from pools C a.id D), Calculation E-6000 concludes that the electrical auxiliary system meets design requirements for proper operation of equipment via both onsite (emergency diesel j generators) and offsite (230kV switchyard) power sources.

ld, at 5 (emphasis added). BCOC's assertion that the Applicant has failed to analyze the available margin in the Unit 1 electrical systems is mistaken. In support of this license O amendment, CP&L has analyzed the margin of the Harris plant electrical systems, both with and without off-site power, and has demonstrated that the Harris plant electrical systems are adequate to meet the needs of pools C and D while also supporting other essential safety functions. Again, O . a contention based upon a mistaken claim that an applicant failed to address a relevant issue should be dismissed because it fails to establish a genuine factual dispute warranting further inquiry. S_ee Q II.B.3., suora. BCOC's Basis 6 must be rejected due to its failure to provide a O sufficient basis to establish a genuine factual dispute warranting further inquiry.

B. Contention 2: Inadequate Criticality Prevention

1. The Contention

.O BCOC asserts in Contention 2 that:

Storage of pressurized water reactor ("PWR") spent fuel in pools C and D at the Harris plant, in the manner proposed in CP&L's

.O license amendment application, would violate Criterion 62 of the General Design Criteria ("GDC") set forth in Part 50, Appendix A.

GDC 62 requires that: " Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations." In lO violation of GDC 62, CP&L proposes to prevent criticality of PWR fuel in pools C and D by employing administrative measures which limit the combination of burnup and enrichment for PWR

( fuel assemblies that are placed in those pools. This proposed reliance on administrative raeasures rather than physical systems or O processes is inconsistent with GDC 62.

g t

y l BCOC Supp. Pet. at 10-11. The asserted basis for the contention are set forth in three pages of

) discussion following the contention.24 In order to facilitate the determination of admissibility of this contention, the Applicant has summarized the bases asserted by BCOC for Contention 2 as follows:25

]

Basis 1 - CP&L's proposed use of credit for burnup to prevent 1 criticality in pools C and D is unlawful because GDC 62 prohibits )

the use of administrative measures, and the use of credit for burnup is an administrative measure;

) Basis 2 - The use of credit for bumup is proscribed because  ;

Regulatory Guide 1.13 requires that criticality not occur without l two independent failures, and one failure, misplacement of a fuel l assembly, could cause criticality if credit for bumup is used. l j

]Lat 12-13. I

2. Applicant's Response to the Contention j Contention 2 must be rejected in its entirety because it advocates stricter requirements than those imposed by the regulations, and therefore constitutes an impermissible collateral l

attack on the Commission's rules. Seg Q II.B.4., supra. Each of the two bases asserted by BCOC j in support of Contention 2 is addressed, in turn, below. l Basis 1 - CP&L's Proposed Use of Credit for Burnup to Prevent Criticality in Pools C and D is Unlawful Because GDC 62 Prohibits the Use of Administrative Measures, and the Use of Credit for Burnup is an Administrative Measure

)

In Basis 1, the Petitioner uses a textual analysis of General Design Criterion 62 ("GDC 62") to draw the conclusion that: I

)

2' The first page and a half of discussion in BCOC's basis recounts the criticality control features l

currently in place for Harris spent fuel pools A and B and those proposed by the Applicant for spent fuel pools C and D. Id a at Il-12.

25

)- S.s { II.C., sygra.

i 4

).l l

h GDC 62 is quite clear that any measures relied on must be ohysical

!. rather than administrative [and therefore] the administrative

/ measures proposed by CP&L [ credit for burnup] must be rejected l as unlawful under GDC 62.

BCOC Supp. Pet. at 12 (emphasis added). The 'only " administrative measure" that BCOC

) addresses in Contention 2 is " credit for bumup." Sss & at 13. BCOC also asserts that the Applicant's reliance on an NRC Regulatory Guide for the acceptability of taking credit for bumup is misplaced because Staff Regulatory Guides are "useful as guides" but "cannot be

) viewed as necessarily controlling." E at 12-13, citing Potomac Electric Power Co. (Douglas

. Point Nuclear Generating Station, Units 1 and 2), LBP-76-13,3 NRC 425,432 (1976).

BCOC's reading of GDC 62 is unsupported by the plain meaning of the words and is

)

- directly contrary to numerous Commission determinations. BCOC's Basis 1 is based on a loose and overbroad interpretation of the text of GDC 62, which interjects an additional term that is not part of the actual text. GDC 62 states: '

).

Criticality in the fuel storage and handling system shall be prevented by ohysical systems or nrocesses, preferably by use of geometrically safe configurations.

1 10 C.F.R. Part 50, App. A, Criterion 62 (emphasis added). A literal reading of the text shows that GDC 62 can be met through the use of either "ohysical systems" or " processes " & The Petitioner chooses to read GDC 62 by adding an additional term, assuming that the text instead

) states " physical systems" or "ohysical processes." Seg BCOC Supp. Pet. at 12. BCOC emphatically asserts that "GDC 62 is auite clear that any measures relied on must be ohynical."

E (emphasis added). BCOC's insistence, however, is not supported by the text of GDC 62.

) The use of the term "or" in GDC 62 indicates that " physical systems" and " processes" are attematives. Ssg American Heritage College Dictionary 959 (3d ed.1993). " Process"is defined as a "a series of actions, changes, or functions," and " processes" as "to put through the steps of a prescribed procedure." & at 1090-91. There is no indication in the common usage that l

i 1 i

" processes""must be ohysical." as BCOC attempts to assert. Credit for burnup is a process 1 which is implemented through a series of written procedures. _Sgg Lic. Amend. Appl., Encl. 7 at 4-4. BCOC provides no argument to counter the interpretation that " credit for bumup" is, in fact, consistent with the common dermition of" processes "

D' Moreover, the NR.C Staff has described the administrative measures required to implement credit for bumup as a " process" involving " written procedures." Draft Regulatory

) Guide 1.13 at 1.13 13,15 (Prop. Rev. 2,1981) (" Reg. Guide 1.13"). Reg. Guide 1.13 is the NRC Staff's specific guidance on acceptable methods for preventing criticality in spent fuel pool

\

storage and complying with the GDC. See jA at 1.13-7,9. With respect to compliance with the j GDC, the Commission has detennined that "[i]f there is confonnance with regulatory guides,

)

. there is likely to be compliance with the GDC." Petition for Emergency and Remedial Action, I CLI-78-6,7 NRC 400,407 (1978) (emphasis added). This is particularly true in the absence of other evidence. Long Island Lightine Co. (Shoreham Nuclear Power Station, Unit 1), LBP .

l 22,17 NRC 608,616 (1983). Here, BCOC has provided nothing to counter Reg. Guide 1.13's interpretation that credit for burnup comprises " processes."

3 BCOC's flawed textual interpretation of GDC 62 cannot provide a valid basis for a litigable contention. Basis 1 of Conte'ntion 2 must be rejected for failure to provide sufficient basis for an admissible contention.

3 More importantly, however, BCOC's interpretation of GDC 62 runs directly afoul of the i

Commission's implementation ofits regulations, and thereby advocates stricter requirements than the Commission's regulations require. In promulgating the General Design Criteria (including GDC 62), the Commission clearly stated that prior to issuing an operating license for a power reactor, or an amendment thereto, y the Commission will require assurance that these criteria have been satisfied in the detailed design and construction of the facility.

F  !

i 1

p j 36 Fed. Reg. 3255,3255 (1971).26 Therefore, when a license amendment is approved, the

)

g Commission has made a determination that the amendment complies with all of the GDC, including GDC 62. The Commission has approved numerous license amendments to allow the L use of credit for bumup to prevent criticality with the use of high-density storage racks in spent f fuel pools. ~ See. e.n.,63 Fed. Reg. 40,551,40,566 (1998) (Waterford); 61 Fed. Reg. at 7566

! i l (Comanche Peak); 59 Fed. Reg. 27,049, 27,703 (1994) (Salem); 58 Fed. Reg. 28,050,28,069 (1993) (Sequoyah). Each of these license amendment approvals is based on "[t]he

]

h' Commission's related evaluation of the amendments . . . contained in . . . a Safety Evaluation."

Sss, t.&,61 Fed. Reg. at 7565-66 (Comanche Peak). Asjust one example, the Commission's

Safety Evaluation for the Comanche Peak license amendment evaluates and approves "burnup dependent criticality analyses" using " reactivity equivalencing" based on " enrichment versus l

- bumup ordered pairs." Sp3 Letter from NRC to TU Electric (issuing license amendments 46 and

32) and enclosed Safety Evaluation at 3-4 (Feb. 9,1996) (PDR Accession ## 9602140197,207) h (" Comanche Peak Safety Evaluation"). In another, the Commission's Safety Evaluation for l

Waterford determines that " General Design Criterion 62 . . . is met" by "burnup reactivity equivalencing" using " enrichment versus bumup ordered pairs." Sse Letter from NRC to j

h Entergy Operations (issuing license amendment 144) and enclosed Safety Evaluation at 2-3

- (July 10,1998) (PDR Accession ## 9807140341,347) ("Waterford Safety Evaluation").27 Just as in the instant case, in Waterford, the Commission approved a Technical Specification change

) to allow spent fuel storage in high-density racks that were limited by administrative measures to I 2'

l . An amendment to a power reactor operating license must comply with the same requirements as ,

those applicable to an initial operating license. 10 C.F.R. 6 50.90. '

I 27 j There are numerous other Commission approvals of credit for burnup for preventing criticality in r-spent fuel pools. See. e.a.. letter from NRC to Public Service Electric and Gas Co. (issuing Salem

! license amendments 151 and 131) and enclosed Safety Evaluation at 6-7 (May 4,1994)(PDR Accession ## 9405100311,316)(" Salem Peak Safety Evaluation"); letter from NRC to Tennessee Valley Authority (issuing Sequoyah license amendments 167 and 157) and enclosed Safety Evaluation at 2-4 (Apr. 28,1993)(PDR Accession ## 9504040161,169)("Sequoyah Safety

. Evaluation").

u

3 i

1 storing only " spent fuel in the ' acceptable range."' Lda , Approved License Amendment and 1

) Technical Specification at 5-6 (Tech. Spec. 5.6.1.g) (PDR Accession # 9807140346). As a condition precedent to approving all of these license amendments, the Commission " require [d]

I assurance that [all of the General Design Criteria] have been satisfied." S_ee

_e 36 Fed. Reg. at

) 3255.

Based on its mistaken reading of the text of GDC 62 and its failure to consider the Commission's many determinations approving the use of credit for bumup, BCOC attempts to l require CP&L to meet stricter requirements than those imposed by the Commission's regulations. A contention that advocates stricter requirements than those imposed by the 1

regulations must be rejected as an impermissible collateral attack on the Commission's  !

3 i regulations. Sig II.B.4., sup.ra. BCOC's Basis 1 asserts that the use of credit for bumup is unlawful under the Commission's regulations, thereby requiring the Applicant to use some other 1

technique to prevent criticality in spent fuel storage pools C and D. _See e BCOC Supp. Pet. at Il- l

12. When the Commission's regulations permit the use of a particular analysis or technique, a l contention which asserts a different technique must be used is inadmissible as a collateral attack on the Commission's regulations. Metrooolitan Edison Co.. (Three Mile Island Nuclear Station, Unit. No.1), LBP-83-76,18 NRC 1266,1273 (1983). Basis 1 of Contention 2 must be rejected because it constitutes an impermissible collateral attack on the Commission's regulations, in violation of 10 C.F.R. Q 2.758. ,

Basis 2 - The Use of Credit for Burnup is Proscribed Because Regulatory Guide 1.13 Requires that Criticality Not Occur Withou Wo Independent Failures, and One Failure, Misplacement ora Fuel Assembly, Could Cause Criticality if

") Credit for Burnup is Used I

p In Basis 2, the Petitioner asserts that the use of credit for burnup is proscribed because it is inconsistent with the statement in Reg. Guide 1.13 that

)

l 1

L

O The nuclear criticality safety analysis should demonstrate that  !

criticality could a91 occur without at least two unlikely, I O . independent, and concurring failures or operating limit violations.

BCOC Supp. Pet. at 13, piting Reg. Guide 1.13 at 1.13-9 (emphasis in original). Without I

explaining why, BCOC asserts that the Applicant's proposed use of credit for burnup:

would not satisfy this requirement because only one failure or violation, namely placement in the racks of PWR fuel not within the ' acceptable range' of bumup, could cause criticality. Note that

'misolacement of a scent fuel assembiv' is identified in the Draft O Reg. Guide as one of nine ' credible normal and abnormal operating occurrences.'

1 BCOC Supp. Pet. at 13 (footnote and citation omitted) (emphasis added). Once again, a review O -- of the Commission's determinations approving the use of credit for bumup is instructive. In Waterford, the Commission's evaluation addressed this very issue.28 Waterford Safety Evaluation, sgga, at 3. The Commission's evaluation states:

O Most abnormal storage conditions will not result in an increase in the k n of the racks. However, it is possible to postulate events, such as the inadvertent misloading of an assembly with a burnup and enrichment combination outside of the acceptable areas in TS Figures 5.6-1,5.6-2, or 5.6-3, which could lead to an increase in O reactivity. However, for such events, credit may be taken for the cresence of at least 1720 carts oer million (nom) of soluble boron reauired in the cool whenever a fuel assembiy is moved, since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accirient O (Double Contingency Principle). The reduction in k n caused by the boron more than offsets the reactivity addition caused by credible accidents. In fact, calculations show that for the most severe accident condition, a soluble boron concentration of 700 ppm boron would be adequate to maintain kenless than 0.95.

O 2

Other Commission approvals of credit for burnup also address this issue. See Comanche Peak Safety Evaluation, sp_ra, at 4; Salem Safety Evaluation spa, at 7; Sequoyah Safety Evaluation

-O apa, at 4.

O

E y ,

hl(emphrsis added). Thus, the Commission has determined that, in the event of" misplacement D'

of a spent fuel assembly," credit for burnup does comolv with the requirement that " criticality could aqt occur without at least two unlikely, independent, and concurring failures." Sg BCOC Supp. Pet. at 13, siting Reg. Guide 1.13 at 1.13-9. It is clear from the Conunission's O

' determinations that there is no conflict between the use of credit for burnup and the requirement that " criticality could not occur without at least two, unlikely, independent, and concurring failures." hl Again, BCOC would have the Applicant meet more restrictive requirements than O those imposed by the Commission's regulations.

Just as in Waterford. an analysis was performed to confirm that " misplacement of a spent g fuel assembly" would not cause criticality. The presence of soluble boron in the spent fuel pool l water, which is required in the Harris spent fuel pools at all times, more than offsets the reactivity addition of the most reactive " misplaced" fuel assembly. In its license amendment i

. application, the Applicant stated that "[t]he use of the high-density region 2 racks has been ,

I shown to be acceptable based on the analysis performed by Holtec Intemational." Lic. Amend.

. App., Encl. I at 2. The Harris spent fuel pools maintain a minimum of 2000 parts per million g (" ppm") of soluble boron in the pool water at all times? Holtec International analyzed l misplacement of a spent fuel assembly with the highest possible enrichment into the spent fuel storage racks to be used in spent fuel pools C and D and confirmed that:

g [A] soluble poison concentration of 400 ppm boron would be sufficient to maintain a kint less than 0.95 (including uncertainties) under the maximum postulated accident condition.

l l l i P i l

l The Harris Plant Operating Manual, Chemistry and Radiochemistry, CRC-001 at 33, requires that j 3 spent fuel pool water maintain between 2000 and 2600 ppm soluble boron at all times.

\

i l

i

Holtec Int'l, Study /Scooine Reoort for Fuel Storage in Harris Pools C and D, HI-971703 at 4-21

? to 22 (July 1997).30 The 400 ppm boron required by analysis is far below the 2000 ppm boron that must be maintained in the Harris spent fuel pools at all times. Thus, just as in Waterford, the Applicant here has demonstrated that the reduction in reactivity caused by the soluble boron in 3 the pool water more than offsets the reactivity addition caused by misplacement of the worst case assembly, and therefore has complied with the Commission's requirements for taking credit for j i

bumup to prevent criticality in spent fuel pool storage. BCOC's contention would have the 1 l

) Applicant meet more restrictive requirements than those imposed by the Conunission's  !

regulations. BCOC's Basis 2 must therefore be rejected as an impermissible collateral attack on the Commission's regulations, in violation of 10 C.F.R. 2.758. Sg Q II.B.4., tup.tg

? C. Contention 3: Inadequate Quality Assurance

1. The Contention l

j BCOC's Contention 3 is multifaceted. Applicant proposes the following restatement of

Contention 3 which incorporates the various allegations found both in the statement of j i

Contention 3 and in the Basis.3'

}

CP&L's proposal to provide cooling of pools C & D by relying  !

upon the use of previously completed portions of the Unit 2 Fuel )

l Pool Cooling and Cleanup System and the Unit 2 Component  !

Cooling Water System fails to satisfy the quality assurance criteria of 10 C.F.R. Part 50, Appendix B, specifically Criterion XIII ,

) (failure to show that the piping and equipment have been stored and preserved in a manner that prevents damage or deterioration),

I Criterion XVI (failure to institute measures to correct any damage l or deterioration), and Criterion XVII (faihire to maintain quality 3

)L The relevant pages of HI-971703 are attached as Exhibit 4. The worst case event at Harris is  !

l misplacement of a fresh unirradiated PWR fuel assembly with 5% enrichment. Sg HI-971703 at 4-l 21. Note that even with ng soluble boron in the pool water, misplacement of the worst case ,

l assembly into pools C and D would p_qt cause criticality, as the resulting maximum reactivity of 1 0.990 is still below 1.000. I 3'

) .Sg; BCOC Supp. Pet. at 14-19.

)

p O_ l records to show that all quMity assurance requirements are satisfied).

The Alternative Plan submitted by Applicant fails to satisfy the l . requirements of 10 C.F.R. Q 50.55a for an exception to the quality l assurance criteria because it does not describe any program for maintaining the idle piping in good condition over the intervening O years between construction and implementation of the proposed license amendment, nor does it describe a program for identifying l and remediating potential corrosion and fouling.

The Alternative Plan submitted by Applicant is also deficient because 15 welds for which certain quality assurance records are lO missing are embedded in concrete and inspection of the welds to demonstrate weld quality cannot be adequately accomplished with a remote camera. j The Alternative Plan submitted by Applicant is also deficient O because not all of the welds embedded in concrete will be inspected by the remote camera and the weld quality cannot be demonstrated by circumstantial evidence.32 As its Basis for Contention 3, BCOC first quotes from 10 C.F.R. Part 50, Appendix B, Criteria XIII, IVI and XVII.33 Next BCOC points to Applicant's statement that certain of the piping isometric packages for field installation of the Unit 2 Fuel Pool Cooling and Cleanup System and Component Cooling Water System piping were inadvertently discarded during a l BCOC raises two other issues. Firg, BCOC questions whether the missing Quality Assurance records are limited to the piping and might not apply to equipment as well. Id. at 15. Applicant's 50.55a Altemative Plan addresses field installation of piping. All Code piping (in the form of O. prefabricated pipe spools) and equipment in the scope of the Altemative Plan was supplied by an approved vendor having the requisite NPT authorization. The vendor data package (including the Code Data Report) for each such item is on hand. Any piping or equipment for which this quality documentation is not on hand will be replaced with appropriately qualified and documented replacement items. Second. BCOC presumes that the remote camera inspection of welds and the piping embedded in concrete will not be conducted until after the issuance of the license anun&nent.

O li at 18-19. In fact the remote camera inspection is scheduled to be conducted within the ner, month and the results will be reviewed by the NRC Staff. Neither of these issues forms the basis for a contention.

BCOC's purported Basis for Contention 3 is discussed at BCOC Supp. Pet. at 15-19; BCOC Supp.

Pet., Ex. 2 (Declaration of Dr. Gordon Thompson at 123) (" Thompson Dect.") and Ex. 4 lg (Declaration of David A. Lochbaurn)("Lochbaum Dect.")

I 10

p

)

l l

1 document control records cleanup effort for Unit 2 documents. BCOC also notes that Applicant k .. is " silent" regarding storage and preservation of previously completed piping and equipment.

' BCOC Supp. Pet. at 17. l j As basis for the assertion that the unused piping may be subject to fouling or degradation, BCOC relies on NRC Information Notices to licensees which discuss problems that have occurred during extended storage or lay-up of piping and equipment. E; Lochbaum Dect. at )

) 117-9.- BCOC faults Applicant's Alternative Plan for not describing "a program for identifying and remediating potential corrosion and fouling." BCOC Supp. Pet. at 18.

BCOC asserts without basis that a remote camera inspection can provide only limited

)

i information about weld quality and cannot provide the level of quality that is available from NDE." E j Mr. Lochbaum cites to a number of 1981 NRC Inspection Reports with minor violations relating to construction ac*1vities at the Harris Plant to suggest that quality standards in the Fuel i Pool piping may not have been met during construction up to the time of cancellation of Unit 2 in December 1983. Lochbaum Dect. at 1110-14.

2. Applicant's Response to the Contention Before addressing Contention 3, we note that on March 24,1999, the NRC Staff  ;

h forwarded to CP&L'a " Request for Additional Information Regarding the Alternative Plan for L

Spent Fuel Pool Cooling and Cleanup System Piping." CP&L responded by letter dated

! April 30,1999 ("RAI Response"). The detailed response to the NRC Staff's questions h incorporates 17 enclosures, including isometric drawings and matrixes which elaborate on the Petitioner does not elaborate on what method of NDE-non-destructive ex.unination - it has in

) mind. A camera inspection is one form of NDE.

L

7...

O information available regarding each weld subject to the 50.55a Alternative Plan. Some of the O

NRC Staff questions and Applicant's responses deal with issues raised by BCOC in Contention

3. Applicant hand-delivered a copy of the RAI Response with all enclosures to BCOC's counsel on May 3,1999. The RAI Response with enclosures 1,3,13 and 16 are included with this O Answer as Exhibit 5.

. a. Lack of Adequate Quality Assurance for Piping O It is undisputed that the piping for the Harris Unit 2 Fuel Pool Cooling and Cleanup System ("FPCCS") was not maintained as part of the licensed HNP, and therefore was not subject to 10 C.F.R. Part 50, Appendix B, once construction of Unit 2 was abandoned in O December 1983. The FPCCS piping was not stored or placed in lay-up prenant to Criterion XIII. It was not subject to the HNP Corrective Action Program. A number of piping isometric packages for field installation of the completed portion of the FPCCS were discarded and are not O available. Lic. Amend. App., Encl. 8, at 3. As a result, certain quality records required by the ASME Code,Section III, are no longer available for 37 of the large bore welds in the completed FPCCS piping.11 at 3,5. Accordingly, BCOC's recitation of the facts relating to the

.O inc mplete c nstructi n f the FPCCS, the inapplicability of the HNP Quality Assurance Program to the FPCCS once it was abandoned, and the discarded piping records fails to identify a genuine dispute with Applicant. The Commission's pleading requirements for contentions O. require that the petitioner "show that a genuine dispute exists with the applicant on a material issue oflaw or fact." S.m QII.B.2., apIn Applicant's own statements are not the basis of a contention.

O

. However, once construction on the Harris Unit 2 FPCCS is completed and the system and spent fuel pools C and D are commissioned and placed in service, the FPCCS must meet the requirements of 10 C.F.R. Part 50, Appendix B. The 50.55a Alternative Plan addresses the C

existing situation where HNP is no longer under construction and certain quality documentation

(

was discarded concerning field welds. Under the circumstances,10 C.F.R. 50.55a permits an alternative demonstration of an acceptable level of quality and safety iu construction. However, j the FPCCS will be subject to 10 C.F.R. Part 50, Appendix B, and must in the future comply with, initI alia, Criteria XIII, XVI, and XVII.  !

I J

The licensed and operating portion of the HNP, including spent fuel pools A and B and the Unit 1 FPCCS, has been subject to the HNP Quality Assurance Program since construction.

BCOC does not dispute the efficacy of the present HNP Quality Assurance Program. BCOC offers no basis for the contention that once placed in service the FPCCS will not successfully meet the requirements of the HNP Quality Assurance Program and 10 C.F.R. Part 50, including Criteria XIII, XVI, and XVII. The only facts presented which border on an attack of the HNP 3

Quality Assurance Program are the presentation of four NRC inspections reports from 1981 which found minor deficiencies in construction quality control. 35 Mr. Lochbaum presents these inspection reports to "suggest that CP&L had problems protecting against deterioration before Unit 2 was cancelled." Lochbaum Decl.,1110-14 (emphasis in original). He does not, however, suggest in any way that the HNP Quality Assurance Program is inadequate.36 There is no basis advanced by BCOC for a contention that the HNP Quality Assurance Program is inadequate or, The 1981 inspection reports, which describe relatively minor deficiencies, certainly cannot support a contention that the HNP Quality Assurance Program is inadequate. Examples of past incidents "are not a sufficient basis to support an assertion that . . . operation might be unsafe [in] the future."

3 Geornia Institute of Technolony (Georgia Tech Research Reactor, Atlanta, Georgia), LBP-95-6,41 NRC 281,299-300 (1995).

Evidence of the efficacy of the HNP Quality Assurance Program is the fact that the Commission issued the Operating License for the Harris plant. In its " Safety Evaluation Report related to the Operation of Shearon Harris Nuclear Power Plant, Units 1 and 2," NUREG 1038 (November 1983),

the NRC Staff concluded: " Construction of Shearon Harris Units 1 and 2 has proceeded, and there is 3 reasonable assurance that it will be substantially completed, in conformity with Construction Permits Nos. CPPR-158 and 159, the application as amended, the provisions of the (Atomic Energy] Act, and l the rules and regulations of the Commission." E at 23-1. The Staff further noted that "such l completeness of construction as is required for safe operation at the authorized power levels must be verified by the Commission before the licenses are issued." M These conclusions were reached two years after the 1981 inspection reports.

)

F

)

i that once placed in commission, the FPCCS will not meet all of the Criteria of 10 C.F.R. Part 50,

) Appendix B. Under the amended Rules of Practice a petitioner must set forth "[a] brief explanation of the bases of the contention." S.sg { II.B.l., agg.

Thus, that part of Contention 3 that alleges CP&L's failure to satisfy 10 C.F.R. Part 50, 3

Appendix B, Criteria XIII, XVI and XVII in the past cannot be a contention. There is no dispute regarding whether the HNP Unit 2 FPCCS was maintained subject to 10 C.F.R. Part 50,

). Appendix B, in the past. It was not. There is no issue to litigate. In the future, CP&L must and will maintain the FPCCS in accordance with the criteria set forth in 10 C.F.R. Part 50, Appendix B. BCOC has provided no basis for a contention that CP&L will not comply with 10 C.F.R. Part j 50, Appendix B, in the future in its operation of the FPCCS and spent fuel pools C and D. A petitioner cannot challenge the applicant's future compliance with clear regulatory requirements without a particularized basis. Sag Q II.B.4., aga Nor has BCOC provided any basis for disputing that the HNP Quality Assurance Program will continue to meet the requirements of 10 3

C.F.R. Part 50, Appendix B, and will continue to provide reasonable assurance of the quality of systems, components and equipment at HNP. Without a basis, a contention cannot be admitted.

Sgg II.B.l., agg. The generalized assertions regarding Applicant's " inadequate quality assurance" must be rejected.

Stripped to its essence, Contention 3 is not about " inadequate quality assurance." Rather,

) BCOC's discussion surrounding Contention 3 and the Lochbaum Declaration address what they perceive to be deficiencies in the 50.55a Alternative Plan. Specifically, BCOC faults the 50.55a Alternative Plan for (1) "failing to describe a program for identifying and remediating potential

) corrosion and fouling;" (2) attempting to demonstrate weld quality by use of a remote camera; and (3) in any event, not even looking at all of the embedded welds. We address each one in I

turn.

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b. Potential Corrosion and Fouling O ^

The 50.55a Alternative Plan does not describe a program for identifying and remediating 1

potential corrosion and fouling. That is not the purpose of the 50.55a Alternative Plan, which {

deals only with an alternative means of demonstrating compliance with the ASME Code, Section O

III. Applicant is not taking exception to the requirement to install quality FPCCS piping that meets the HNP's design basis. Applicant has developed an " Equipment Commissioning Plan" which addresses inspections of the piping and acceptance criteria to ascertain whether the O.

extended storage of the piping and equipment without controlled storage conditions and regular maintenance has resulted in any degradation, including corrosion (microbiologically induced or j otherwise).37 However, the Equipment Commissioning Plan is not, and need not be, part of the O license amendment application. Rather, the Equipment Commissioning Plan is a CP&L intemal document that establishes how Applicant will ensure compliance with NRC regulations, license requirements, and Technical Specification requirements.

O The inspections of the piping to determine if degradation has occurred are described briefly in the RAI Response.38 The portions of the piping attached to spent fuel pools C and D have been flooded with water from the spent fuel pools for a number of years.39 RAI Response, O

Encl. I at 8. Before the FPCCS piping is inspected, the water will be drained and sampled for any potentially harmful contaminants or microorganisms. The piping will be inspected by a g remote camera to determine whether corrosion or other degradation has occurred since The FPCCS piping is 304 or 316 stainless steel piping,3/8 inch in thickness, and either 12 or 16 .

inches in diameter. The Equipment Commissioning Plan is found at RAI Response, Encl.16 at 8-10.-

O *

. RAI Response, Encl. I at 5,8,14 and 15.

The water in the FPCCS piping is the same water as found in the spent fuel pools, which has not corroded stainless steel components in the pools. Nor have there been observed any " minor pinhole leaks" in the FPCCS piping after years ofin-place storage with standing water, as occurred in a ,

relatively short time in piping systems in other plants subject to microbiologically induced corrosion. 1 l

O ME Infonnation Notice No. 85-30 (April 19,1985).

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construction. The camera will have sufficient resolution capability to identify and provide a

'O basis for dispositioning discrepancies which could exist as a result ofimproper installation or i

subsequent degradation. The inspection will be conducted by an appropriately trained and i f qualified Level II NDE inspector and the inspection will be videotaped. RAI Response, Encl.1 )

O at 14-15.

Accordingly, BCOC has not presented an issue that raises a genuine and material issue in

g. . dispute. Applicant has a plan and procedure to inspect the FPCCS piping to determine ifit has degraded in any way, including degradation due to corrosion. The NRC will be reviewing the results of the piping inspection.d With respect to the one issue raised by BCOC that was pleaded with specificity - the potential susceptibility of stainless steel piping to O

microbiologically induced corrosion, Applicant will determine whether the microorganisms found to be the cause of the corrosion at its Robinson Unit 2 or any other potentially harmful mi r rganisms are present in the water in the FPCCS piping at HNP.d' The remote camera O'

inspection will look for corrosion or degradation of any kind. Unless BCOC can establish with i basis and specificity that this inspection plan is inadequate to ensure the piping meets its design )

O' The NRC Staff will be reviewing the results of the piping inspections to confirm that the Applicant will meet its design basis. "The NRC staff has the continuing responsibility to assure that all regulatory requirements are met by an applicant and continue to be met throughout the operating life of a nuclear power plant." Southern California Edison Co. (San Onofre Nuclear Generating Station, Units 2 and 3), ALAB-680,16 NRC 127,143 (1982).

.O "

The environment in the piping at Robinson, and at other plants where microbiologically induced corrosion has been observed, was quite different from that in the FPCCS piping at Harris. The water in the Robinson piping that was subject to microbiologically induced corrosion was service water

. (lake water) which has a high propensity for microorganisms. Sg IE Information Notice No. 85 30 (April 19,1985). In contrast, the water in the FPCCS piping is chemically-treated, demineralized O water, which d es n t afford a favorable environment for microorganisms. The IE Information Notice cited by Petitioner as a basis for microbiologically induced corrosion refers to microorganisms in " soils, sediments, natural fresh water (gg, wells, rivers, ickes) brackish and sea water, as well as oil and other natural petroleum products." M. at 2. None of these describe the water in the FPCCS. There is no indication that microorganisms have been found in internal plant, I demineralized water that is chemically treated. Petitioner has not provided adequate basis for a contention that the FPCCS piping could be subject to attack by microorganisms.

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requirements, there is no real dispute susceptible of resolution in an adjudication. This aspect of h Contention 3 must also be dismissed.

l c. Remote Camera Inspection of Weld Quality l

j BCOC's assertions regarding the adequacy of the remote camera inspection are not particularized or pled with any specificity. BCOC asserts that " remote camera inspection can l provide only limited information about weld quality, and cannot provide the level of quality g_ assurance that is available from NDE." BCOC Supp. Pet. at 18. After incorrectly speculating that the inspection will not be performed until after the license amendment is issued,42 BCOC baldly asserts: "The results of the remote camera inspection are not likely to yield clear 'yes' or g-i

'no' answers regarding weld quality or whether significant degradation or fouling has occurred.

The interpretation of these results, and whether they satisfy section 50.55a, must be subject to questioning in this proceeding." hL at 19. There are no facts or assertions regarding the g adequacy of remote camera inspection of weld quality in the Declaration of Mr. Lochbaum or

- elsewhere in BCOC's Supplemental Petition.

Greater detail regarding the remote camera inspection is provided in the RAI Response, O Encl. I at 14-15. A pipe crawler mounted camera will perform a detailed inspection of the interior surfaces of embedded field welds. The camera will be capable of camera resolution to at least 1/32 inch strand of wire. Embedded field welds will be inspected for cracks, lack of fusion, O lack of penetration, oxidation, undercut greater than 1/32 inch, reinforcement greater than 1/16 I

inch, concavity greater than 1/32 inch, porosity greater than 1/16 inch, and inclusions. While this is not the same as NDE inspections performed at the time of the welding, it will provide

?. direct physical evidence of quality of the embedded welds. Lic. Amend. App., Encl. 8, at 10.  ;

1 i

The remote camera inspection of the piping is scheduled for late May or early June of 1999. RAI

) Response, Encl. I at 15.

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)

y Furthermore, hydrotest records are available for piping lines that include 13 of the 15 embedded k welds. The hydrotest records confirm that each weld had successfully completed NDE and a Weld Data Record had been reviewed prior to the hydrotest." Successful reinspection by NDE l of all 22 accessible field welds - with no rejections - on the same piping provides additional

) ~ assurance of weld quality." Lic. Amend. App., Encl.' 8, at 9-10.

l Accordingly, BCOC has not presented an issue relating to the remote camera inspection that raises a genuine and material issue in dispute. Applicant has an inspection plan and acceptance criteria to inspect the FPCCS piping to determine the physical condition of accessible embedded field welds. The NRC will be reviewing the results of the piping inspection. Unless

) BCOC can establish with basis and specificity that this inspection plan is inadequate to ensure the piping welds meet design requirements, there is no real dispute susceptible of resolution in a 1

hearing. This aspect of Contention 3 must also be dismissed. l l

j d. Inspection of Fewer than All Embedded Welds BCOC contends that "CP&L's approach for the two-thirds of the embedded welds that I

will receive no inspection is inadequate to provide the level of quality and safety required by

) section 50.55a." BCOC Supp. Pet. at 19. BCOC asserts that the " circumstantial evidence" that i confums that the welds were actually inspected is "not an adequate substitute for actual documented evidence that inspections were conducted and the welds found to be in acceptable

?--- condition."" E However, neither the Declaration of Mr. Lochbaum nor the Declaration of RAI Response, Encl. 3, lists each field weld and the records available to support the weld quality.

1 "

All accessible field welds in the scope of the 50.55a Alternative Plan have been reinspected using original construction criteria from ASME Section III,1974-winter 1976 Addenda, ND-5000. Se.e RAI Response, Encl. I at 5.

45 Indeed, the hydrotest records for 13 of the 15 welds are " actual documented evidence that inspections were conducted and the welds found to be in acceptable condition." It is only the specific Weld Data Record itself that is missing.

l

)

. i Dr. Thompson offers any basis for the assertion that Applicant's 50.55a Alternative Plan is O inadequate if all of the welds cannot be inspected."

Currently 6 of the 15 embedded field welds are included in the inspection plan for the remote Camera, including the two field welds for which hydrotest records are not available. RAI Response, Encl. I at 17. The quality of those embedded field welds which are not accessible for remote camera inspection is assured by virtue of the HNP Quality Assurance Program and

~ASME Quality Assurance Program that was in effect at the time of the welding of the large bore piping and throughout construction of the HNP. Considerable evidence exists that the welds of the FPCCS piping were conducted in strict adherence to the programmatic requirements of the HNP Quality Assurance Program, including: (1) the quality of the construction of the licensed O

HNP; (2) re-performance of Code required inspections on accessible field welds in the same piping with no rejectable indications identified; and (3) the existence of numerous Quality Assurance records from the time of plant construction which supports this conclusion. & The 50.55a Alternative Plan has been endorsed by CP&L's nuclear insurer, Hartford Steam Boiler Inspection and Insurance Company. The endorsement letter is authored by Dr. Richard Fiegel, Vice President of Hartford and Chairman of the ASME Council on Codes and Standards. E, Encl. I at 18, Encl.13.

There is no basis to support a contention that the 50.55a Alternative Plan is not O' acceptable because all of the welds will not be inspected by remote camera. In light of the detailed discussion of the adequacy of the 50.55a Alternative Plan in Enclosure 8 to Applicant's license amendment application and in the Response to the NRC Staff's RAls, BCOC must  !

O provide something more, with specificity, to support this aspect of Contention 3.

l Dr. Thompson simply asserts that " failure to satisfy ASME code requirements could increase the probability of design-basis or severe accidents at pools C and D." Thompson Decl., Ex. 2, at 5-6.

No basis is provided for this assertion. Dr. Thompson never suggests what about Applicant's 50.55a Alternative Plan is unacceptable.

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e. Failure to Show a Specific, Tangible Link between Alleged Errors in the 50.55a Alternative Plan and lO Health and Safety Impacts

[

Contention 3 must also be rejected in its entirety because BCOC has failed to show a l specific, tangible link between the alleged deficiencies in the 50.55a Alternative Plan and public lO health and safety. .Spe II.B.6, sp_rg. The suction and discharge of the FPCCS piping in the spent fuel pools are located approximately five feet below the surface of the water and well l above the level of the spent fuel to be stored in spent fuel pools C and D." The piping lines are

.O not subject to high pressure because they are open to the pools, which are open to atmospheric pressure. If one of the piping lines developed a leak, the spent fuel pools could not empty and

. the fuel would remain covered with water. If a weld that is embedded in concrete had a defect, O

there is nothing that could happen that would have any impact on public health and safety. There is no significant pressure in the piping to propagate the defect in the weld. If the weld were to develop a crack, there would be nowhere for the water to go with concrete encasing the piping.

O As noted previously, Dr. Thompson offers no factual basis or even a theoretical explanation for his bald assertion that lack of Quality Assurance documents could " increase the probability of design-basis or severe accidents at [ spent fuel] pools C and D." S.es Thompson Decl. at 6.

g.

Mr. Lochbaum's concerns regarding safety are limited to the " failure . . . to provide reasonable assurance against possible deterioration of the installed Unit 2 spent fuel pool cooling system."

Lochbaum Decl. at j 11. But even here, Mr. Lochbaum fails to show a specific, tangible link O i between the purported failure in the plan to provide for the possibility of deterioration of the piping and any health and safety impacts. A contention must be dismissed where the I

O "

i Harris FSAR at 9.13-6a to 66. "The reduction of the normal pool water level by approximately 5 ft.

due to any postulated [FPCCS] pipe failure will have no adverse impact on the capability of the cooling system to maintain the required temperature and it does not affect the required shield water depth for limiting exposures from the spent fuel. The slow heatup rate of the fuel pool would allow sufficient time to take any necessary action to provide adequate cooling using the backup provided p while the cooling capability for the fuel pool is being restored." 1.d. at 9.13-6b.

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" contention, if proven, would be of no consequence . . . because it would not entitle [the]

lO petitioner to relief." 10 C.F.R. @ 2.714(d)(2)(ii); Yankee Atomic, LBP-96-2, smrg,43 NRC at 78, a[d, CLI-96-7, sipig,43 NRC 235. Here BCOC has utterly failed to show that the consequences ofits alleged deficiencies could in any way affect public health and safety. For O this reason alone, Contention 3 must be rejected.

3. Summary of Response to Contention 3 O Contention 3 is broadly worded with basis and specificity offered for only a narrow segment of the Contention advanced. Eirst, Contention 3 must be rejected because there is no genuine issue in dispute regarding the fact that the FPCCS was not subject to 10 C.F.R. Part 50, '

A PPendix B, and the HNP Quality Assurance Plan once construction of Unit 2 was abandoned.

O Second, there is no basis advanced for a Contention regarding " inadequate quality assurance," l either past or future. Petitioner has made'no particularized demonstration that Applicant will not c nf rm to the Commission's explicit requirements for Quality Assurance of the FPCCS and O

pools C and D. Ihird, BCOC has not advanced a basis for a challenge to the adequacy of Applicant's planned inspections to determine if the FPCCS piping has been subject to any rr si n r f uling. Nor has Petitioner provided an adequate basis for its assertion that stainless O

steel piping with chemically-treated, demineralized water could be subject to microbiologically-induced corrosion. Fourth, BCOC has not provided a basis for its generalized contention of the i adequacy of the 50.55a Alternative Plan using remote camera inspection to confirm existing O

quality documentation. Eth BCOC has not explained why inspection ofless than 100% of the I embedded welds does not provide adequate assurance of weld quality, assuming the remote l

camera inspection of 6 of 15 welds confirms what the NDE inspection found in 22 similar field 0-welds in the same piping. Finally, BCOC has not shown any link between the alleged inadequacies in the 50.55a Alternative Plan and public health and safety. For all of these reasons, Contention 3 must be rejected.

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D IV. GROUP II: ENVIRONMENTAL CONTENTIONS At the outset, we note that the NRC Staff and NRC Counsel have advised Applicant that the NRC Staff, in its discretion, will prepare an environmental assessment ("EA") in connection with its consideration of the instant license amendment application. Thus, the NRC Staffhas O

elected to travel a well-worn path where the destination is inevitable. Because of the Department of Energy's delay in implementing the Nuclear Waste Policy Act of 1982 and in developing the permanent repository for spent nuclear fuel, license amendments to expand spent fuel storage O

capacity have been requested and granted at almost every nuclear operating facility - often more

[

than once. In each case an environmental assessment has been prepared. In each case there has '

been a finding of"no significant [] environmental impacts associated with the proposed action."

O

.S_ee, e_g, 64 Fed. Reg. 2,688 (Union Electric Company, Callaway Plant) (1999); 64 Fed. Reg.

23,133 (Florida Power & Light Company, St. Lucie Plant)(1999). Accordingly, the NRC has never prepared an environmental impact statement ("EIS") in connection with the many O

expansions of on-site spent fuel storage in existing spent fuel pools. See, e.e.. Vermont Yankee Nuclear Power Cornoration (Vermont Yankee Nuclear Power Station), ALAB-919,30 NRC 29, 42 n.13 (1989); Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and O

2), LBP-87-24,26 NRC 159,166 (1987).

The Commission has expressly addressed the environmental and radiological effects of

, on-site spent fuel storage generically in the context oflicense renewal. _S_ee " Environmental Review for Renewal of Nuclear Power Plant Operating Licenses," 61 Fed. Reg. 66,537,66,538 (1996). The Commission has found by rule:

3 [I]f necessary, spent fuel generated in any reactor can be stored safely and without significant environmental impacts for at least 30 years beyond the licensed life for operation (which may include the term of a revised or renewed license) of that reactor at its spent fuel storage basin or at either onsite or offsite independent spent

, fuel storage installations.

• l 1

10 C.F.R. s 51.23(a). S_ee e also Oconee, CLI-99-11, suora, slip op. at 20 (1999). This generic D

finding is focused on the storage of spent fuel after cessation of reactor operation. The specific assessment regarding additional spent fuel storage in pools at HNP will inevitably dictate the same finding of"no significant environmental impacts associated with the proposed action."

D Accordingly, Applicant sought to have this license amendment treated as a " categorical exclusion" not requiring an environmental review or environmental assessment, pursuant to 10

, C.F.R. 51.22(c)(9). However, the NRC Staffs decision to prepare an environmental assessment in its discretion simply either moots, or requires rejection as premature, Contentions 4 through 8. In addition, Contention 6 raises an issue outside of the scope of this proceeding.

, Contention 8 asks the Licensing Board to take an action outside the scope ofits authority. We address each environmental contention in turn.

A. Contention 4: Proposed License Amendment Not Exempt from NEPA D

1. The Contention BCOC asserts in Contention 4 that:

D CP&L errs in claiming that the proposed license amendment is exempt from NEPA under 10 C.F.R. 51.22.

BCOC Supp. Pet. at 22. As its bases for Contention 4, BCOC repeats the allegations in D Contentions 1 though 3 (ist at 21-22) and argues that Applicant does not qualify for a

" categorical exclusion" pursuant to 10 C.F.R. s 51.22(c)(9) by repeating its comments on the NRC Staff's preliminary determination of"no significant hazards consideration"(isl aat 22-36).48

)- As will become clear, it is unnecessary to discuss Petitioner's ptuported basis for this contention.

'8 The bulk of the Petitioner's arguments directly and impermissibly would challenge the NRC Staff's "no significant hazards consideration" determination. "No petition or other request for review or

)' hearing on the staff's significant hazards consideration determination will be entertained by the Footnote continued on next page

)

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2. Applicant's Response to the Contention

O Applicant never claimed that it was " exempt from NEPA." Applicant believes that this license amendment application falls within the " categorical exclusion" set forth at 10 C.F.R.

51.22(c)(9). The categorical exclusions are part of the Commission's implementation of the

'O' National Environmental Policy Act ("NEPA"). The Commission has found by rule that a certain

" category of actions does not indivMually or cumulatively have a significant effect on the human environment." 10 C.F.R. { 51.22(a). In any event, this contention is moot because the NRC )

O  !

Staffis preparing an environmental assessment pursuant to its regulations implementing NEPA.  ;

1 The NRC Staffis neither treating the license amendment application "as exempt from NEPA" nor as a " categorical exclusion." Contention 4 does not raise a genuine issue in dispute and must O

be rejected.

1 B. Contention 5: EnvironmentalImpact Statement Required l

1. The Contention

O BCOC's Contention 5 asserts the following:

The proposed license amendment is not supported by an lO Environmental Impact Staternent ("EIS"), in violation of NEPA and NRC's implementing regulations. An EIS should examine the effects of the proposed license amendment on the probability and consequences of accidents at the Harris plant. As required by NEPA and Commission policy, it should also examine the costs O and benefits of the proposed action in comparison to various Footnote continued from previous page Commission. The staff's determination is final, subject only to the Commission's discretion, on its

'O own initiative, to review the determination." 10 C.F.R. { 50.58(b)(6). Yankee Atomic Electric Co.

(Yankee Nuclear Power Station), CLI 98-21,48 NRC 185,204 n.7 (1998); Florida Power & Light CA (St. Lucie Nuclear Power Plant, Unit 1), LBP-88-10A,27 NRC 452,456-457 (1988)(holding that the " Board is barred as a matter of Commission regulation" from granting a hearing on the Staff's significant hazards consideration determination). BCOC is precluded by 10 C.F.R.

{ 50.58(b)(6) from challenging the first prong of the test for " categorical exclusion" under 10 C.F.R.

O_ { 51.22(c)(9).

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5 alternatives, including Severe Accident Mitigation Altematives and the alternative of dry cask storage.

I l BCOC Supp. Pet. at 36. As its basis for this Contention 5, BCOC incorporates by reference the l bases advanced for Contention 4 and legal argument as to the requirements of NEPA. hi m at 36-i p 38. Again, it is unnecessary to discuss Petitioner's purported basis for this contention for the reasons discussed below.

2. Applicant's Response to the Contention D

The Commission's rules at 10 C.F.R. { 51.31 provide that "[u]pon completion of an environmental assessment, the appropriate NRC staff director will determine whether to prepare .I O an environmental impact statement or a finding of no significant impact on the proposed action."

Consequently, it is premature to admit a contention asserting that an environmental impact statement is required until the NRC Staff has issued its environmental assessment. Pursuant to l

C 10 C.F.R. { 51.20, a prerequisite for the instant licensing action to require preparation of an environmental impact statement is an NRC Staff finding that the proposed action is "a major Federal action significantly affecting the quality of the human environment." Absent such a D- finding, an environmental impact statement is not required pursuant to the criteria set forth in  ;

Q 51.20. A contention that an environmental impact statement is required where the criteria in Q 51.20 are not met would impermissibly challenge the Commission's rules. As discussed in D Section II.B.4., synta, a contention may not challenge a Commission rule.

In Diablo Canyon, LBP-93-1, supra,37 NRC 5, this exact issue was addressed by the licensing board in connection with a contention that an environmental impact statement must be prepared in connection with the issuance of a license amendment. At the time of the prehearing conference, the Staff had not yet prepared its environmental assessment. The Diablo Canyon licensing board held that admitting such a contention was premature:

Insofar as this contention seeks an EIS, therefore, it is premature.

We are denying it on that basis. After the Staffissues its EA, and

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assuming that the EA will not call for an EIS, [intervenor] may l submit a late-filed contention calling for an EIS. Such a C. contention, to be accepted, would have to be based on substantial l and significant information indicating why an EIS is called for.

! & At 36. See also Consumers Power Co. (Big Rock Point Nuclear Plant), ALAB-636,13 NRC O 312,350 (1981) (reversing the licensing board's order to the NRC Staff to prepare an EIS on a l

proposed spent fuel pool expansion prior to the Staff's preparation ofits environmental assessment)."

g.

Both the Commission's rules and precedent require that Contention 5 be rejected as premature. Consequently, while Applicant strongly disagrees with the legal and factual arguments advanced in support of Contention 5, it serves no purpose to argue them in the abstract. Certainly if an environmental impact statement is not required and the Conunission does not decide to prepare one in its discretion, what should be considered in such an environmental impact statement is an academic issue. If Petitioner disagrees with the NRC C

Staff's findings in the environmental assessment, the Commission's rules provide an opportunity for late-filed contentions.10 C.F.R. Q 2.714(a)(3).

C. Contention 6: Scope of EIS Should Include Brunswick and Robinson O Storage

1. The Contention BCOC asserts in Contention 6 that:

3 The EIS for the proposed license amendment should include within its scope the storage of spent fuel from the Brunswick and Robinson nuclear power plants.

D " See generally Kellev v. Selin,42 F.3d 1501,1518 (6th Cir.) cert. denied. 515 U.S. I159 (1995)("An agency decision, based on an EA, that no EIS is required, can be overturned only ifit is arbitrary, capricious, or an abuse of' discretion' . . . We will not ' substitute ourjudgment of the environmental impact for thejudgment of the agency, once the agency has adequately studied the issue."') siting Crounse Com. v.1.C.C. 781 F.2d 1176,1193 (6th Cir.) cert. denied. 479 U.S. 890 p- (1986).

3

O BCOC Supp. Pet. at 38. BCOC provides two brief paragraphs as basis for this contention.

O First, BCOC asserts that "there is no independent utility to the racking of a spent fuel pool: the only reason for the application is to permit the expansion of spent fuel storage at the g plant [,] . . . not only . . . scent fuel generated by Harris, but also . . fuel from Brunswick and Robinson." & (emphasis added).

Second, BCOC contends that "CP&L has a global plan for storage of spent fuel from its three North Carolina reactors, includine the ootion of dry cask storace at Brunswick." &

(emphasis added). BCOC bases this allegation on the fact that CP&L submitted an application for an ISFSI at Brunswick 10 years ago, and BCOC's assumption that "the application is still O

pending."5 E BCOC also bases its contention on the fact that a Department of Energy (DOE) report from 1994 states that an ISFSI at Brunswick will be used as a backup for storage if transshipment to Harris "is prohibited." E at 39.

O 2. Applicant's Response to the Contention Just as Contention 5 must be rejected as premature, so should Contention 6. However, Contention 6 should also be rejected with prejudice at this stage because it attempts to raise issues that are beyond the scope of this proceeding, and runs directly counter to Commission precedent. Furthermore, the specific bases asserted by the Petitioner fail to meet the J

Commission's pleading requirements because the first part of the bases fails to identify a genuine dispute with Applicant, and the second part is factually incorrect.

D BCOC states that it was unable to determine if the application was still pending based on its review g of the correspondence index in the NRC Public Document Room. BCOC Supp. Pet. at 38.

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a. BCOC's Basis Fails to Meet the Commission's Pleading Requirements O

BCOC's first specific basis fails to identify a genuine dispute with Applicant. The Commission's pleading requirements for contentions require that the Petitioner "show that a O genuine dispute exists with the applicant on a material issue oflaw or fact." 10 C.F.R. l

@ 2.714(b)(2)(iii). Any contention not meeting the Commission's contention requirements "must be rejected." Palo Verde, CLI-91-12, g.gpra,34 r NRC at 155; e alse { II.A;, suprg. BCOC's O first basis states that the reason for CP&L's amendment application is to permit the expansion of spent fuel storage at Harris for " spent fuel generated by Harris" as well as " fuel from Brunswick and Robinson." BCOC Supp. Pet. at 38. CP&L concurs with this statement and, in fact, O specifically stated in its license amendment application that:

l Activation of these two pools [C and D] will provide storage l capacity for all four CP&L nuclear units (Harris. Brunswick 1 a_nd

2. and Robinson) through the end of their current licenses.

O Lic. Amend. App., Encl. I at 1 (emphasis added). NRC granted CP&L a license for Harris to I receive and possess spent fuel transshipped from Brunswick and Robinson at Harris as part of Harris's initial operating license approval in 1987. & Shearon Harris Nuclear Power Plant,  ;

Unit 1, Facility Operating License, License NPF-63 at 3 (Jan.12,1987)(Section 2.B(8)). The instant license amendment application does not involve the legal ability of Harris to accept and store spent fuel from Harris, Brunswick, and Robinson; the NRC approved this over 12 years ago. See e j_ dad The only issue presented by this amendment is how spent fuel at Harris, including fuel from Harris, Brunswick and Robinson, will be stored. Applicant fully agrees with the Petitioner that the reason for activating Harris spent fuel pools C and D is specifically to store

" spent fuel generated by Harris" as well as " fuel from Brunswick and Robinson."" h BCOC However, even if spent fuel were no longer shipped from Brunswick and Robinson to Harris, there is

" independent utility" in the instant license amendment application to store spent fuel from Harris

Q alone.

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Supp. Pet. at 38. BCOC's first specific basis does not identify a dispute with Applicant, and D

therefore this basis must be rejected because it fails to comply with the Commission's pleading requirements to "show that a genuine dispute exists with the applicant." See II.B.2., suora.

, BCOC's second specific basis asserts that "CP&L has a global plan for storage of spent fuel" which includes "the option of dry cask storage at Brunswick." BCOC Supp. Pet. at 38.

BCOC asserts as its support the fact that CP&L submitted an application for an ISFSI license at

, Brunswick 10 years ago, and BCOC's mistaken belief that "the application is still oending." Isd2 (emphasis added). However, in September 1991, CP&L requested the NRC to delay issuing the license pending further notice. Letter from G. Vaughn (CP&L) to NRC (Sept. 17,1991)(PDR

, Accession No. 9109260069). By letter dated November 2,1994, the NRC informed CP&L that:

Because of your circumstances and inactivity in the licensing process, we have determined that to continue the review is not an effective use of resources and, therefore, are suspendine review of your license aonlication.

D Letter from C. Haughney (NRC) to R. Anderson (CP&L) at 1 (Nov. 2,1994) (PDR Accession No. 9411090152) (emphasis added). Since that time, no further activity has been performed on 3 an ISFSI at Brunswick. BCOC's assertion that the " application is still pending" is mistaken.

The only other support cited by Petitioner for this basis is a DOE report from 1994. This report does not provide a basis for the contention. The statement in the DOE report is predicated D

on use of the Brunswick ISFSI if transshipment to Harris "is prohibited." Sse BCOC Supp. Pet.

at 39. As discussed above, the Harris license explicitly allows receipt of spent fuel transshipped from Brunswick. Therefore, transshipment to Harris is clearly nq1 orohibited and thus the

) predicate for the statement in the DOE report is not correct. A petitioner's mistaken understanding of the facts regarding an application does not provide a basis for a litigable contention. Seg @ II.B.3., suora. BCOC's assertion that the application for an ISFSI at Brunswick "is still pending"is both unsubstantiated and mistaken. The DOE statement cited by

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BCOC for support is reliant on a predicate circumstance (prohibition of transshipment) that does O not exist. Because BCOC's statements supporting this basis are unsubstantiated and mistaken, BCOC's second asserted basis for Contention 6 must be rejected for failing to provide a sufficient basis for an admissible contention, as required by the Commission's regulations.

O b. BCOC's Contention is Outside of the Scope of this Proceeding BCOC's Contention 6 must also be rejected because it raises an issue that is outside of '

O the scope of this proceeding. The alternative of spent fuel storage at Brunswick and Robinson would have been within the scope of the hearings on receipt of Brunswick and Robinson fuel at Harris, as part of the initial operating license for Harris in 1987. However, the alternative of O

spent fuel storage at Brunswick and Robinson is outside the scope of this license amendment I proceeding to expand the capacity of the Harris spent fuel storage pools.

O BCOC's Contention 6 asserts that the EIS for this license amendment should include

" Brunswick and Robinson storage," including "the option of dry cask storage at Brunswick."

BCOC Supp. Pet. at 38. This very issue has previously been addressed in a prior agency O Proceeding. _See Vircinia Electric and Power Co. (North Anna Power Station, Units 1 and 2),

LBP-84-40A,20 NRC 1195,1200, aff'd. ALAB-790,20 NRC 1450,1453-54 (1984). In North Anna, the applicant ("VEPCO") had submitted two separate license amendment requests: (1)

O n li ns am ndment request was to receive spent fuel shipped from VEPCO's Surry plant at North Anna (Case OLA-1); and (2) a second license amendment request was to expand the storage capacity of the North Anna spent fuel pools (Case OLA-2). North Anna, LBP-84-40A,

'""'*' 20 NRC at 1195; see also North Anna, ALAB-790, suora,20 NRC at 1451-52. The O

petitioner in that case attempted to include a contention in the North Anna spent fuel pool expansion proceeding, OLA-2, asserting that the environmental analysis for the North Anna Spent fuel pool expansion amendment must " consider [] the attemative method of constructing a

.O dry cask storage facility at the Surry Station." North Anna, LBP-84-40A, supra,20 NRC at 1199

O

(emphasis added). The petitioner's contention, therefore, asserted that the amendment for spent 3

fuel pool expansion at the North Anna plant must consider the alternative of dry storage at VEPCO's Surry plant. The Board rejected this contention as beyond the scope of the spent fuel pool expansion proceeding (OLA-2), stating that the contention was " directed solely to the 3 transshipment of Surry spent fuel assemblies or to an alternative thereto." E at 1200. The Board concluded that the two proceedings were separate actions and that the contention regarding dry storage the Surry plant lacked basis with respect to the North Anna spent fuel pool

) expansion proceeding.

The Atomic Safety and Licensing Appeal Board affirmed the Licensing Board's decision.

North Anna, ALAB-790, supra,20 NRC at 1454. The Appeal Board agreed that the two J

proceedings were separate, and that the petitioner's " bases . . . were inadequate to allow [the petitioner) to be heard with regard to the proposed modification of the North Anna spent fuel

) pool." E at 1453. The Appeal Board concluded that:

As a matter of both fact and law, a modification to the North Anna spent fuel pool can and will have no bearing upon whether . . .

VEPCO is given the green light to transport Surry assemblies for receipt and storage at North Anna.

E at 1454.

Just as in North Anna OLA-2, in this proceeding the Applicant is seeking approval to

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expand the capacity of the Harris spent fuel pools. Just as the petitioner asserted in North Anna, BCOC has asserted here that the environmental analyses to support the Harris spent fuel pool expansion must consider storage at the Applicant's other licensed plants as an alternative.

)

BCOC Supp. Pet. at 37-38. Again, as in North Anna, the proceeding to address receipt of spent <

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0-1 fuel at Harris from the Applicant's other reactors was a separate licensing proceeding.52 Just as I.O in North Anna the Petitioner's contention asserting that storage at Brunswick and Robinson must be considered in the license amendment for spent fuel pool expansion at Harris is beyond l the scope of this license amendment proceeding.

O' l- Contention 6 must be rejected because it lacks basis, is beyond the scope of this proceeding, and is contrary to NRC case law precedent.

O' D. Contention 7: Environmental Assessment Required

1. The Contention l BCOC asserts in Contention 7 that:

l0-Even if the Licensing Board finds that no EIS is required, it must order the preparation of an EA.

O i

BCOC Supp. Pet. at 39. Again, it is unnecessary to discuss Petitioner's purported basis for this contention as the NRC Staff has stated that it will prepare an environmental assessment.

2. Applicant's Response to the Contention

!O-Contention 7 must be rejected as moot in light of the NRC Staff's decision to prepare an i environmental assessment in connection with its consideration of the license amendment

, application. See Q III.D.2., agtra

'O l

E. Contention 8: Discretionary EIS Warranted

1. The Contention 1O BCOC Contention 8 asserts the following:

_ Indeed, the proceeding to approve receipt of Brunswick and Robinson spent fuel at Harris occurred over 12 years ago. flee ce Carolina Power and Light Co. (Shearon Harris Nuclear Power Plant), ALAB-

O 837,23 NRC 525,542-44 (1986).

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l Even if the Licensing Board determines that an EIS is not required under NEPA and 10 C.F.R. Q 51.20(a), the Board should

'O- nevertheless require an EIS as an exercise ofits discretion, as permitted by 10 C.F.R.' (( 51.20(b)(14) and 51.22(b).

BCOC Supp. Pet. at 40. BCOC provides four pages of text as the basis for its contention. To O. facilitate the determination of admissibility of this contention, the Applicant has summarized the bases asserted by BCOC for Contention 8 as follows:

10 C.F.R. { 51.20(b)(14) and 51.22(b) provide for the preparation of an EIS where, upon its own initiative or request from any party, O' the Commission finds that "special circumstances" exist, and this case presents special circumstances; The NRC should prepare an EIS as an exercise ofits discretion; O An EIS should include storage of spent fuel at Brunswick and Robinson; The NRC should evaluate the apparent conflict between the CP&L proposal and the NRC's Waste Confidence decision.

O BCOC Supp. Pet. at 40-43.

2. Applicant's Response to the Contention

.O 1 As with Contention 5, it would be inappropriate to consider whether an environmental - '

impact statement should be prepared until after the NRC Staff publishes its environmental l assessment. However, Contention 8 should also be rejected with prejudice because the O

Licensing Board has no authority to direct the Commission to perform a discretionary act.

Furthermore, Petitioner has made no showing of"special circumstances" which would warrant such a discretionary environ. mental impact statement. Finally, preparation of an environmental

.O impact statement regarding additional spent fuel storage at HNP in a spent fuel pool would simply be redundant of a number of definitive, generic findings by the Commission regarding the "small" and " insignificant" environmental impacts from many decades of on-site spent fuel

.O

- storage.

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a. Outside the Scope of the Licensing Board's Authority

) This Licensing Board was established to preside over any proceeding pursuant to l

Petitioner's hearing request in connection with CP&L's license amendment application. The delegation of authority from the Commission is set forth in the Federal Register notice D

establishing the Licensing Board. 64 Fed. Reg. 10,165 (1999). The delegation refers specifically to a number of sections in 10 C.F.R. Part 2. Nowhere does the Commission delegate to the Licensing Board the authority to order the preparation of a discretionary environmental impact statement pursuant to 10 C.F.R. { 51.20(b)(14) or 51.22(b).

Indeed, in promulgating its environmental rules, the Commission stated that 3 Section[s] 51.20(a) and (b)(13) [now renumbered as (b)(14)] also provides that the Commission may prepare an [EIS] in connection with other types of proposed actions . . . when the Commission determines, in the exercise ofits discretion, that it is advisable to do so. It is not possible to predict how often or under what 5

circumstances the Commission micht wish to exercise this discretion.

49 Fed. Reg. 9352,9362 (1984)(emphasis added). The Commission goes on to say that:

3 the Commission believes that its responsibilities for protecting the public health and safety and giving appropriate consideration to environmental values will be best served ifit retains the flexibility and authority to direct its staff to prepare environmental assessments or environmental impact statements very early in the decisionmaking process.

E at 9366 (emphasis added). The Commission reserved to itself the discretion to direct the NRC Staff to prepare environmental assessments and environmental impact statements that were not required by its regulations.53 The Commission also notes that:

l In a similar circumstance in St. Lucie. LBP-88-10A, suora,27 NRC at 457, the Licensing Board

) found that in promulgating 10 C.F.R. 50.58(b)(6),"the Commission made it clear that the reference Footnote continued on next page t

p i

l the Commission may wish, as a matter of discretion, to have the I

! benefit of an environmental assessment or an environmental  !

) impact statement in considering the desirability of a proposed

! course of action, even though, as a strict legal matter, neither may l be required. A major purpose of { Sl.22(b) is to preserve this l necessary flexibility. l

( l b E The Commission goes on to say that:

l It is not possible to predict how often or under what circumstances l the Commission might wish to exercise this discretion. However, there are likely to be at least a few occasions on which actions,

) which in normal circumstances might qualify for a categorical 4

l exclusion or only result in a finding of no significant impact following the completion of an environmental assessment, would, because of uniaue. unusual or controversial circumstances, require extensive environmental review. 1 3

E at 9362 (emphasis added). Here there are certainly no unique, unusual or controversial circumstances to the Commission, as amendments to expand spent fuel storage at reactor sites j have become commonplace.54 However, the Commission has reserved to itself the determination of when such a test might be met. Absent a specific delegation of authority, the Licensing Board cannot order the preparation of a discretionary environraental impact statement.

) Contention 8 must be rejected because the Licensing Board could not order the relief requested.

)

3-Footnote continued from previous page to the ' Commission' meant the Commissioners themselves and that this Board had no authority to

. act on the Staff's finding as such."

The standard would be rendered meaningless if a petitioner's intervention and contention itself y rendered the circumstances " controversial."

r O

L b. Petitioner Offers No Basis for "Special Circumstances" That Could Warrant a Discretionary Environmental

,O

Impact Statement l I

l 10 C.F.R. Q 51.20(b)(14) and 51.22(b) provides far the preparation of an environmental a l

assessment or an EIS where, upon its own initiative or request from any party, the Commission j finds that "special circumstances" exist. Sg ahto 10 C.F.R. 51.21. Petitioner refers to CP&L's plan to store spent fuel from Brunswick and Robinson at Harris and Petitioner's preference for dry cask storage at Brunswick and Robinson as part of the "special circumstances" here. BCOC l Supp. Pet. at 40 - 42. For the reasons set forth in response to Contention 6, CP&L's plans for storage of spent fuel at Brunswick and Robinson are outside the scope of this proceeding. The i

HNP is already licensed to accept and store spent fuel from those CP&L-owned units. i O i The second "special circumstance" cited by Petitioner is an " apparent conflict" between  ;

the CP&L license amendment request to expand its spent fuel storage capacity at the Harris  :

O facility and the NRC's Waste Confidence decision. Sg BCOC Supp. Pet. at 42. Supporting its 4 l

assertion, the Petitioner quotes Enclosure 1 of the Applicant's license amendment request," DOE l

spent fuel storage facilities are not available and are not exoected to be available for the

]

O foreseeable future." BCOC Supp. Pet. at 42 (emphasis added by Petitioner). The Petitioner also l provides an excerpt from the Commission's Waste Confidence determination that "[t]here is reasonable assurance that at least one mined geologic repository will be available within the first O. quarter of the twenty-first century . . ." IdJciting 10 C.F.R. { 51.23).

There is no conflict between the Applicant's statement and the Commission's determination. The Applicant states that the DOE repository is not expected to be available for O

over 10 years and that CP&L's storage needs begin in the year 2000, which is long before the DOE expected availability date in 2010. Lic. Amend. App., Encl. I at 1. The Applicant's statement and its need for storage in the year 2000, before the DOE expected availability date of D

the repository in 2010, is certainly not in conflict with the Commission's determination that a 3

O repository would be available "within the first quarter of the twenty-first century," or before the O

year 2025. The Petitioner's mistaken understanding of the Applicant's Matement in the amendment request does not form the basis for any "special circumstance." Indeed, the entire nuclear utility industry is faced with the same delay by the Department of Energy.

O

, c. Preparation of a Discretionary EnvironmentalImpact Statement Would Be Redundant to Generic EnvironmentalImpact Statements Prepared by the Commission

'O The Commission has prepared a number of generic environmental impact statements that have looked at the environmental consequences oflong-term storage of spent nuclear fuel in on-site spent fuel pools and elsewhere. The Commission's findings have been consistent for over two decades and were repeated approvingly just last month, as noted in the introduction to the

' discussion of environmental contentions, agga at 53. In its " Generic Environmental Impact Statement for License Renewal of Nuclear Plants," NUREG-1437, at Vol.1, xlviii (May 1996),

O the Commission found:

[T]here is ample basis to conclude that continued storage of existing spent fuel and storage of spent fuel generated during the  !

O li enSe renewal period can be accomplished safely and without significant environmental impacts. Radiological impacts will be ]

well within regulatory limits; thus radiological impacts of on site storage meet the standard for a conclusion of small imoact. The nonradiological environmental impacts have been shown to be not significant; thus they are classified as small. The overall l 0 conclusion for on-site storage of spent fuel during the term of a renewed license is that the environmental impacts will be gn_gil for each plant.

1 O (Emphasis added.) The Commission defined "small" to mean "not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource." hd2 l at Vol.1, xxxv. See also " Final Generic Environmental Impact Statement on Handling and l O Storage of Spent Light Water Power Reactor Fuel," NUREG-0575, Vol.1 (August 1979). These facts would not change with a new, discretionary environmental impact statement.

O

jD"

I

? l l Contention 8 must be rejected.

to l IV. CONCLUSION l l

For the reasons set forth with respect to each contention, Applicant submits that  !

l LO Contentions I through 8 must be rejected and, consequently, BCOC's Petition to Intervene must ;

l be dismissed.

l Respectfully submit 1 d, l 1 6dd ~ .

\

!O OfCounsel:

Steven Carr John HJO'Neill, Jr.

i((iapt i R. Hollaway V I Legal Department SHAW, PITTMAN, POTTS CAROLINA POWER & LIGHT & TROWBRIDGE COMPANY 2300 N Street, N.W. i 411 Fayetteville Street Mall Washington, D.C. 20037-1128 I Post Omce Box 1551 -CPB 13A2 (202) 663-8148 O Raleigh, North Carolina 27602-1551 Counsel For CAROLINA POWER &

(919) 546-4161 LIGHT COMPANY Dated: May 5,1999

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D UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board D

In the Matter of )

)

CAROLINA POWER & LIGHT ) Docket No. 50-400-LA COMPANY )

D (Shearon Harris Nuclear Power Plant) ) ASLBP No. 99-762-02-LA CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing " Applicant's Answer to Petitioner Board of D Commissioners of Orange County's Contentions," dated May 5,1999, was served on the persons listed below by U.S. mail, first class, postage prepaid, and by electronic mail transmission, this 5th day of May,1999.

G. Paul Bollwerk, III, Esq., Chairman Frederick J. Shon Administrative Judge Administrative Judge D

Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 e-mail: gpb@nrc. gov e-mail: fj's@,nrc. gov B Dr. Peter S. Lam Office of the Secretary Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel Washington, D.C. 20555-0001 U.S. Nuclear Regulatory Commission Attention: Rulemakings and Adjudications Washington, D.C. 20555-0001 Staff p e-mail: psl@nrc. gov e-mail: hearingdocket@nrc. gov (Original and two copies)

Marian Zobler, Esq.

Richard G. Bachmann, Esq.

• Adjudicatory File Office of the General Counsel

) Mail Stop O-15 B18 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555 e-mail: harris @nrc. gov

) *

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[

iO Diane Curran, Esq. James M. Cutchin, V, Esq.

Harmon. Curran, Spielberg & Atomic Safety and Licensing Board Panel Eisenberg, L.L.P. U.S. Nuclear Regulatory Commission 1726 M Street, N.W., Suite 600 Washington, D.C. 20555-0001 Washington, D.C. 20036 e-mail: jmc3@nrc. gov

'O e-mail: dcurran@Moncurran.com  ;

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• by mail only

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John . O'Neill, Jr.

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3 g Title /Acoroval Sheet SYSTEM # 4065 CALC. TYPE Mechanical O

CAROLINA POWER & LIGHT COMPANY m SF-0040 (CALCULATION #)

FOR Scent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis n (TITLE INCLUDING STRUCTURE / SYSTEM / COMPONENT) v FOR SHEARON HARRIS NUCLEAR POWER PLANT X n NUCLEAP, ENGINEERING DEPARTMENT v

QUALITY CLASS XA OB OC OD 0E REV. RESPONSIBLE pj DESIGN VERIFIED BY APPROVED BY NO. ENGINEER O ENGINEERING REVIEW BY RESPONSIBLE SUPERVISOR DATE DATE DATE n AA >n 0 / W D /h'ff T po ().1l A V l- \p- 0 m

~' REASON FOR CHANGE O

REASON FOR CHANGE O

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. Project No.: CALCULATION SHEET pgg D-Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project'Ibermal-Hydraulic Analysis List of Effective Pages PAGE REV PAGE REV PAGE REV l 1 0 28 0 B e 11 0 29 0 C 0 l g lii 0 30 0 D 0 1 0 31 0 E O 2 0 32 0 F 0 3 0 G 0 4 0 H 0 5 0 I 0 i g 6- 0 J 0 l 7 0 K 0 8 0 L 0 9 0 M 0 i 10 0 N 0 11 0 0 l 0 '

g 12 0 P O 13 0 Q 0 l 14 0 R 0 15 0 S 0 16 0 T 0 17 0 U 0 18 0 V 0 19 0 W 0 20 0 X 0 21 0 Y 0 22 0 Z 0 23 0 AA 0 24 0 BB 0 25 0 CC 0 26 0 Attachments DD 0 27 0 A l 0 EE O l

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es gi.c32 **' O Project No.: CALCULATION SHEET p;j,, j

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Title:

Spent Fuel Pools C and D Activation Project \

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Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis

{

i Table of Contents l Section hage LISTOF EFFECITVE PAGES i

TABLE OF CONTENTS 11

.O 1.0 PURPOSE 1

2.0 REFERENCES

1 j

3.0 ENGINEERING ANALYSIS SOITWARE 3

'O 4.0 CALCULATION 3

S.0 CONCLUSIONS 32 Attachments Subject Total Pages O A Calculation SF-0040, Revision 0, PROTO-FLO" Model Modifications for 77 the HNP Component Cooling Water System Rev 2 B Calculation SF-0040, R? vision 0, CCWS Alignment Summary 3 C Calculation SF-0040, Revision 0, Evaluation of Minimum RHR Heat 17 Exchanger CCW Flow Requirements for Design Basis Accident Conditions

.O D Calculation SF-0040, Revision 0, Evaluation of Maximum RHR Heat 34 Exchanger CCW Flow Requirements for Design Basis Accident Conditions E Calculation SF-0040, Revision 0, Evaluation of Minimum SFP Heat 34  !

Exchanger CCW Flow Requirements for Various Operating Conditions F Calculation SF-0040, Revision 0, Rebalance CCW System Flow 14 Distribution For LOCA: Sump Recirculation (RHR Only) Alignment

'O G CalculatiorrSF-0040, Revision 0, Determine Minimum CCW Heat 19 Exchanger Service Water Flow During LOCA: Sump Recirculation (RHR.

Only) Alignment H Calculation SF-0040, Revision 0, Rebalance CCW System Flow 37 j Distribution for Minimum CCW Pump Developed Head  !

I Calculation SF-0040, Revision 0, Evaluation of CCW System Normal 35 O System Alignment Hydraulic Performance J Calculation SF-0040, Revision 0, Evaluation of CCW System Dual Train 40 Hot Shutdown (350F) System Alignment Hydraulic Performance K Calculation SF-0040, Revision 0, Evaluation of CCW System Single Train 35 Hot Shutdown (350F) System Alignment Hydraulic Performance ,

L Calculation SF-0040, Revision 0, Evaluation of CCW System Refueling 33 I

'O- Core ShufYle System Alignment Hydraulic Performance M Calculation SF-0040, Revision 0, Evaluation of CCW System Refueling 39 Normal Full Core Offload System Alignment Hydraulic Performance 30

)' Cr-cted by: Date:

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

re 1 a 32 *'O Project No.: CALCULATION SHEET p;g Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis

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1.0 PURPOSE I ne purpose of this calculation is to document the thermal hydraulic capacity of the Component Cooling Water System (CCWS) to support the activation of Spent Fuel Pools C and D at CP&L's Harris Nuclear Plant (HNP).

His calculation is only valid for Spent Fuel Pool C and D heat loads up to 1.0 MBTU/hr and does not consider

) the effect of potential increases in core thermal power due to the Steam Generator Replacement / Power Uprate Project.

2.0 - REFERENCES (1) Harris Nuclear Plant Calculation CC-0039 Revision 0, Development of Component Cooling Water System PROTV-FLO nermal-Hydraulic Model

) (2) Harris Nuclear Plant Calculation SW-0088 Revision 0, Development of Emergency Service Water System PROTO-FLO nermal-Hydraulic Model (3) Harris Nuclear Plant Calculation HNP-M/ MECH 1011 Revision 2, Pump Degradation Limits for l ESW, CCW & ESCW, dated 5/10/97 (4) Stone & Webster Feasibility Study for Pool Cooling and Clean-Up of Harris Nuclear Plant Spent Fuel Pools C & D. Revision 0, prepared 10/6/97 (5) Prelimmary Harris Nuclear Plant Drawing CAR 2166-G-412 Rev 11 dated 10/6/97

} (6) Preliminary Harris Nuclear Plant Drawing CAR 2165-G-255 Rev 16, dated 4/4/97

(7) Preliminary Harris Nuclear Plant Drawing CAR 2165-G-127 Rev 15, dated 10/4/97 (8) Crane Technical Paper 410, C1988 Crane Company

)

(9) Harris Nuclear Plant Calculation NSSS-38 Revision 2, RHR Heat Exchanger and Pump Cooler Cooling Water Outlet Temperatures, dated 4/30/97 l (10) Harris Nuclear Plant Engineering Service Request 9700536 Rev 0, Emergency Service Water System -

! FSAR Table 9.2.1 5 Supporting Documentation, dated 10/16/97 l

(11) Harris Nuclear Plant Engineering Service Request 9600126 Rev 0, Spent Fuel Pool Cooling System, dated 3/5/97 (12) Harris Nuclear Plant Final Safety Analysis Report Section 9.2.2 Component Cooling System Table 9.2.2-3 Amendment No. 35 (Superseded by RAF 2160)

) (13) Harris Nuclear Plant Design Basis Document, Component Cooling Water System, DBD-131 Revision 6, dated 6/19/97

r j

Cwpted by: Dite:

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date: 1 N iiia32 "* 0 Project No.: CALCULATION SHEET pig Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis Attachments Subject Total Pages N Calculation SF-0040, Revision 0, Evaluation of CCW System Refueling 41 Abnormal Full Core Offload System Alignment Hydraulic Performance i

O Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA-Safety 36 l Injection Phase Alignment Hydraulic Performance j P Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA- 18

] Containment Sump Recirculation (RHR Only) Alignment Hydraulic Performance Q Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA- 25 Containment Sump Recirculation (RHR and SFP) Alignment Hydraulic Performance R Calculation SF-0040, Revision 0, Evaluation of CCW System Normal 32

] System Alignment Thermal Performance l

S Calculation SF-0040, Revision 0, Evaluation of CCW System Dual Train 33 Hot Shutdown (350F) System Alignment Hermal Performance i T Calculation SF-0040, Revision 0, Evaluation of CCW System Single Train 30 Hot Shutdown (350F) System Alignment Thermal Performance U Calculation SF-0040, Revision 0, Evaluatiot of CCW System Refueling 30 j Core ShufGe System Alignment hermal Performance V Calculation SF-0040, Revision 0, Evaluation of CCW System Refueling 31 Normal Full Core Offload System Alignment nermal Performance W Calculation SF-0040, Revision 0, Evaluation of CCW System Refueling 32 Abnormal Full Core Offload System Alignment %ermal Performance X Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA-Safety 33 3 Injection Phase Alignment Thermal Performance Y Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA- 18 Containment Sump Recirculation (RHR Only) Alignment Thermal Performance Z Calculation SF-0040, Revision 0, Evaluation of CCW System LOCA- 27 Containment Sump Recirculation (RHR and SFP) Alignment Thermal j Performance AA Calculation SF-0040, Revision 0, Evaluation of UHS Thermal Margins 4 BB Calculation SF-0040, Revision 0, Evaluation of Short Term Transient Fuel 68 Pool Temperature Response During HNP Cooldown Operations CC Calculation SF-0040, Revision 0, Design Verification Records 28 DD Calculation SF-0040, Revision 0, Evaluation of CCW System Plant Startup 36 Alignment Hydraulic Performance EE Calculation SF-0040, Revision 0, Evaluation of CCW System Plant Startup 33 Alignment Thermal Performance

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n n---- ted by: Date:

jeg[ undy CAROLINA POWER & LIGifT COMPANY Calculation ID: SF-0040

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42 d 32 "* 0 Project No.: CALCULATION SHEET p;g l- Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis

-(14) Harns Nuclear Plant Calculation CC-0038 Revision 0, CCW Heat Exchanger Performance Dunng Post-Accident Recire Alignment, dated 4/21/97 (15) Harris Nuclear Plant Calculation SW-0085 Revision 0, Ultimate Heat Sink Analysis, dated 1/6/96 l

l '(16) Harris Nuclear Plant Calculation CC-0037 Revision 2, CCW Flow Rates for Vanous Valve

) Alignments, dated 4/8/97 ,

(17). Reactor Coolant Pumps, Technical Manual VM-MRF

' (18) Harris Nuclear Plant Design Basis Document, Service Water System - Traveling Screens and Screen Wash System - Waste Processing Building Cooling Water System, DBD-128, Revision 6, dated

) 6/18/97 (19) Harris Nuclear Plant Technical Specification Section 3/4.7.5 Ultimate Heat Sink, Tech Spec laterpretation 95-03

'(20) Harris Nuclear Plant Calculation SW-0078 Revision 4, ESW System Perf(xmance Evaluation, dated 6/11/96 (21) Harris Nuclear Plant Calculation HNP-M/ MECH-1008, Revised Contamment Analysis for an Increase in the laitial Temperature from 120*F to 135'F Revision 1, dated 4/8/97 (22) Harris Nuclear Plant Calculation CC-0020, Revision 1, Component Cooling Water System Performance, dated 9/3/96 (23) Meeting Minutes of 11/25/97 Meeting Between CP&L and Proto-Power Corporation (24) Harris Nuclear Plant Engineering Service Request - Action Item, ESR 9500442 Revision 0 AI#2, 1

dated 8/l1/97

) (25) Harris Nuclear Plant Final Safety Analysis report AmenAmant no. 45 p. 5.4.7-10I, "Boration and Inventory Control" (26) Harris Nuclear Plant Calculation HNP-F/NFSA-0026 Revision 0, Maximum Decay Heat Load for Spent Fuel Pools A, B & C Through the End of Year 2001, dated 4/16/98

)' (27) Not Used.

l (28) CP&L-Harris Nuclear Plant Letter 10003481-Model-00, Estimated Impact of Power Uprate, dated L November 6,1997 (29) Harris Nuclear Plant Calculation SW-0080 Revision 5, ESW Flow Requirements Based on Reservoir Level, dated 5/2/97

)

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JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

'83d 32 "" O 3 Project No.: CALCULATION SHEET p;;,.

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis

) (30) Hanis Nuclear Plant Operating Procedure OP-145, Section 8.9 (31) Harris Nuclear Plant Final Safety Analysis Report Table 9.2.1-7, Amendment 15 (32) Westinghouse letter CQL-290, dated 6/509 b (33) Harris Nuclear Plant Engineenng Service Request 9700272 Revision 0, dated S/6/97 (34) Harris Nuclear Plant Calculation 9-FHB-2 Revision 1, Fuel Handling Building Air Conditioning System, dated 5/24/86 (35) Harris Nuclear Plant Engineering Service request 9700252 Revision 0, Evaluation of EPT-174 Data, O dated 40/97 .

1 3.0 ENGINEERING ANALYSIS SOFTWARE His calculation was performed using PROTO-FLO 3.04 and PROTO-HX" 3.02. He default PROTO-FLO database, CCW2.DBD (dated 10/14/98, Size 800KB) is included in Attachment (A).

O 4.0 CALCULATION Reference (1) was used as a starting point for the analysis of the CCWS system to determine thermal and hydraulic margins. He default benchmarked PROTO-FLO" database, CCW.DBD, was modified to create a new PROTO-FLO* default database, CCW2.DBD, which incorporates the proposed CCWS tie-ins for the fuel g pool C and D heat exchangers as well as other modifications defined in Table 1. Case alignments for: l

- Startup Operations (A CCWS Train Operating)

Normal Operations (A CCWS Train Operating)

- Hot Shutdown at 350*F (A and B CCWS Trains Operating, Split),

- Safe Shutdown at 350*F (A CCWS Train Operating, Single Failure),

O Refueling: Core Shuffle (A CCWS Train Operating, Single Failure),

- Refueling: Full Core Offload (A CCWS Train Operating, Single Failure),

Refueling: Abnormal Full Core Offload (A and B CCWS Trains Operating, Split),

LOCA: Safety Injection Phase (A CCWS Train Operating),

- LOCA: Containment Sump Recirculation with CCWS Nonessential Header Isolated (Recirc(a)]

(A CCWS Train Operating, Single Failure) and j j

b - LOCA: Containment Sump Recirculation with Limited Fuel Pool Cooling [Recirc(b))

(A CCWS Train Operating, Single Failure).

j i

i a

were developed to capture all the major CCWS system operating conditions. All heat exchanger thermal l

models use design fouling factors rather than IST results to ensure that design basis conditions can be met even '

1 with extreme fouling conditions. CCW pump degradation to the 10% IST limit, Reference (3), was included  !

for the flow margin portion of this analysis.

{

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i

r Cw..yoted by: Date:

Je6Lundy CAROLINA POWER & LIGHT COMPANY C:lculation ID: SF-0040 Checked by: Date:

,,4 g Project No.: CALCULATION SHEET pg; Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis j Table 1 Modified CCWS Pipe Sections Pipe Section Serv 6ce Modi 6 cat 6en 64 BRS Supply Replaced MiscK of 1100 with MiscK=18.l t from l

Reference (22)

Adjusted ICC-356 to 24.17% Open O ^dJ"*d 'cC- 5' 8 75 % Pen Adjustad 1CO363 to 20.56% Open 64 8RSEC Added isolation valve to simulate CP&L direction to assume the BRS Skid is abandoned '"re.

85 BRSEC Added isolation valve to simulate CP&L direction to assume the BRS Skid is abandoned !"~

41/53/60/64 LOCAlsolate Added simulation valve for LOCA case shgnrnents, j b Nodc0001 Pressure in-line Node References (12)and (13)

Changed to in-lme pressure node to clinunaic Node 0026 bypass flow through the Surge Tank which is not I consistent with actual CCWS operation 1 FixedI/ Faxed 2 Deleted Nodes Elinunated Surge Tank nodes and lines to properly model CCWS and eliminate recirculating flow through the Surge Tank 105 RHR Pmp B Ctr Corrected Heat Load Tag O i2i ^Hxisoi Added SrP H= ^ isointion vaive FPI/FP2 Added simulated fuel pool coolang purrys DummySFPCPurg Added simulated fuel pool coohng pump curves calibrated to 3750 spm per Reference (11)

BRS Evap Cooler Deleted fixed heat load per Assunption 4.1.12 Pumpt Degn&dPumpt Added 10% TDH Degraded CCW Pump Curve Pump 2 DegradedPump2 Added 10% TDH Degraded CCW Pump Curve G Pump 3 DegradedPump3 Added 10% TDH Degmded CCW Pump Curve 314 TEMP 1 Added TEMPI isolation Yalve to Enhance Computational Stability 300-319 Proposed CCWS Tie-Ins to Additions are denoted by Altxx. See Attachment A FP Hx C and D 900-905 Fuel Pools A/B and C/D Added simulation for fuel pools A/B and C/D to provide fuel pool equilibrium temperature as a g

function of fuelpool heat load.

27/28/29 Plisolate Added Pumpt Isolation valves with Cv=1000000 to allow for Pump 2 (B Train) Operation SFP Hx D Fixed Heat Load Changed SFP Hx D to a fixed heat load to irriprove l computational efficiency at low CCWS flow rates and light FP C/D heat load.

33 DischXTic Added gate valve with Cv=1000000 to sinustate split CCW train ops, Reference (30) 20 SuctXTie Added gate valve with Cv=1000000 to simulate splitCCWtrainops Reference (30) 43 XSLD HX Added Simulation isolation Valve with Cv=100000 to isolate the Excess Letdown Heat Exchanger Only O

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C+ted by: DIte:

Jeff Lundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

'85 d 32 a"0 Project No.: CALCULATION SHEET Fk Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis

)

4.1 Bases and Assumptions 4.1.1 Case alignments which specify a single CCWS train operation assume the use of the "A" train as CCW Pump A delivers slightly less total developed head and therefore is the least hydraulically capable CCW pump.

) 4.1.2 All CCWS cooled heat exchangers use 'fesign fouling factors. His assumption is unconservative when analyzing the perfo mance ofindividual beat exchangers but is conservative and realistic in terms of overall CCWS thermal performance as the CCW beat exchanger fouling factor significantly exceeds the other CCWS cooled heat exchangers and limits the heat rejection capability of the CCWS. He tube plugging for the CCW heat exchanger is also assumed to be 0% u the design CCW heat exchanger tubeside fouling factor of 0.00176 hr-sqft.'F/ BTU significantly (50.4 percent) exceeds the current worst case trended tubeside fouling factor,

) Reference (35), of 0.00117 hr-sqft *F/ BTU thus the assumption of additional CCW heat exchanger degradation from tube failures would be overly conservative, given the excessive design fouling factor.

4.1.3 CVCS flow to the letdown heat exchanger is assumed to be at design Letdown flow conditions of 120 gpm per CP&L direction, Reference (23).

) 4.1.4 Both RHR pumps and oil coolers are assumed to be operating and rejecting heat whenever the RHR system is activated for conservatism except for single CCW train failure cases which include Safe Shutdown (350*F),

Refuel-Core Shuffle, Refuel-Normal Full Core Offload and all LOCA cases.

4.1.5 The minimum ESWS flow to the CCW heat exchangers is 8500 gpm. .

) 4.1.6 A maximum ESWS supply temperature to the CCW heat exchangers is assumed to be 95'F, Reference (13).

4.1.7 For the purposes of this analysis, Spent Fuel Pool heat exchangers A and D are in operation. It is assumed that the hydraulic resistance of CCWS piping to and from Spent Fuel Pool heat exchangers B and C are equivalent to Spent Fuel Pool heat exchanger A and D supply and return piping.

) 4.1.8 A maximum CCWS supply temperature of 105'F is assumed to be applicable during all operating modes except for Hot and Safe Shutdown Cases and LOCA: Containment Sump Recirculation Cases, Reference (IP 4.1.9 A maximum CCWS supply temperature of 120*F is assurned for all CCWS system lineups other than those identified in Assumption 4.1.8. Reference (13) states that the CCWS is designed for a maximum temperature of 120*F (for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) which is based on the maximum permissible temperature to the reactor

)

coolant pumps. A review of the reactor coolant pump technical manual, Reference (17), with the cogmzant plant engineer shows that there is no explicit time limitation on operation of the reactor coolant pumps with thermal barrier cooling in excess of 105'F so long as RCS temperature is less than 400*F. %erefore, it is assumed that the statement of"approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />" is descriptive in that the CCWS supply temperature is

- only expected to be in excess of 105'F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during plant cooldown operations.

) 4.1.10 Le reactor coolant pumps are assumed to be secured during Safe Shutdown, Refueling Operations and LOCA:Recirc cases. He CCWS flow is assumed to be supplied to the RCPs, for the Safe Shutdown and

)

l O-Computed by: Date:

JeffLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date:

'8 6 #32 h'O p, j,,, y CALCULATION SHEET p;;,,

O Project

Title:

Spent Fuel Pools C and D Activation Project )

Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis lQ Refueling operations cases, even though they are not rejecting heat to the CCWS. This assumption is conservative in terms of CCWS flow margins.

l 4.1.11 The heat load from the Gross Fuel Failure Detector (GFFD) and the Prunary Sample Coolers are considered to l be transient relative to the total steady state CCWS heat load and are assumed to be negligible for a steady state system thermal-hydraulic calculation per discussions with HNP System Engineering.

'O 4.1.12 The CCWS alignments assume that the Boron Recovery Skid is abandoned in place and does not requae heat l

removal or CCWS flow, per CP&L System Esigliiser-g ducction, Reference (23).

4.1.13 Analytical thermal uncertainty on overall CCWS heat transfer is assumed to be inherent and included in individual shell and tube heat exchanger models which were developed from manufacturer data sheets.

'O 4.1.14 Letdown heat exchanger operation is NOT required during Safe Shutdown conditions as boration capacity is  !

required to be maintained by the Boric Acid Tank, the Boric Acid Transfer Pumps, the Refueling Water Storage  ;

Tank and the Centrifugal Charging Pumps, Reference (25). '

4.1.15 CCW trains 'A' and 'B' are split whenever both RHR heat exchangers are in service, Reference (30), with the j O n nessential header assumed to be aligned to the 'A' CCW train. i 4.1.16 CCWS flow to the letdown heat exchanger is set to 610 gpm (575 gpm, Reference (12) + 6% hydraulic uncertainty, Reference (1)) for the purposes of establishing a hydraulic design basis for the CCW system.

l 4.1.17 It is assumed that this calculation is only valid for Spent Fuel Pool C and D heat loads less than 1.0 MBTU/hr.

,O 4.1.18 The thermal effect of the HNP Power Uprate project increased core thermal rating is not accounted for in this calculation.

4.1.19 Excess letdown heat exchanger process side parameters are only specified for the plant Startup case alignment when maximum letdown system capacity is required. Excess letdown heat exchanger CCWS flow is maintained O f r all alignments except for the LOCA:Recirc (RHR Only) and LOCA:Recirc (RHR and SFP) alignments during which the excess letdown heat exchanger is isolated by the Phase A containment isolation signal

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O Computed by: Date:

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

], 7 ,, g Project No.: CALNEN SHEET

-O p;ie Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis O

4.2 CCWS Alignments The baseline CCWS alignments were developed based on Reference (12) defined lineups with the exception of the excess letdown heat exchanger. Thermal and hydraulic margins for the CCWS are not compared to the values in Reference (12). Rather, all margin comparisons are to either design data sheet values or to inferred flow and heat load O

values from other design basis documents or calculated values provided herein. All alignments assume the operation of one or two CCW trains, consistent with plant operating requirements. For Safe Shutdown, single failure Refueling operations and LOCA case alignments, the CCW "A" cooling train is considered to be in operation. Table 2 sununanzes each operating CCWS lineup.

Table 2 O Major CCWS Alignments Startup L..e Hot 50 Safe 50 Refusi Refund Redund LOCA LOCA LOCA 1/3 Core Normal Fut Atarmel Fun Safety Sump Rodrc wi1h Sump Radre Load Mode 1 @ 4 hrs (350F) Shufne Core Ofiloed Core CIRoad insedian Esserent Hooder with Limited (Mode 6) (Mode 6) (Mode 6) Phase Only 0 SFP Cootn0 RHR Pmp B Flow Only Flow Ordy x x Fkm OrW RHR Hz B x Fkm Only RHR Pmp A Fkm Onty Fkm Only a x x x x Fh Orty x x RHRHxA x x Flow Orvy Flow Only Fion Only x x BRS: Dist Car BRS: Evap Car BRS: Vent Corxt Q L&k-..Hx XSLD Hx x

x x

Fkm Onty x

FW Only x

FW Ordy Fh Only Fkm Only Flow Only Flow Only RCDT Hx x x x x x x x x Seel Water Hx x x x x x x x x SFP Hx A x x x x x x x x x SFP Hx B RCP A a x x Fh Onty Fum Only Fkm Only Flow Onty a RCPB x x x Flow Ordy Flow Only Flow Ordy Flow Only x RCP c I x x Fkm Orty Flow Only Flow Only Fkm Only x 0 SFe Ha c SFP Hz D z x x x x x x x x GFFD Fkm Only Fkm Onty Flow Only FkmOney Flow Only Flow Only Fkm Orty Sample Cooters Fkm Crdy Fkm Orvy Flow Only Fkm Ordy Flow Oniv FW Only Flow Only C., v CCW Trans 1 1 2 (Sept) 2 (Spid) 1 1 1 1 14 1/0 g tr wry single g Failure. B' CCWSingle CarWasnment Failure. AI O woies . . - 8* rain, *ear v raiDn, v ~A "

- Cooiers -

Fakse XSLD Hz. RCDT CC"* "* Leeds leoluted Hx Secura4 Only Except forSFP

"~

~ """ T All operating lineups use the benchmarked CCW pump curves for the thermal margin analysis and the IST program 10 0 percent degraded pump curves for the flow margin analysis. CCW flow to the cooled components for normal operations is consistent with the benchmarked values of Reference (1).

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} C+q-2ted by: Date: 1 JeffLundy - CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 l Checked by: Date:

es g < 32 "" O Project No.: CALCULATION SHEET Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Eermal-Hydraulic Analysis l

l

} The thermal conditions applied in each CCWS =h-t are the design values included in Reference (1) and

==---.u in Table 3, except where specifically noted.

He RHR heat exchanger flow for all cooldown conditions is based on maintaining the CCWS supply temperature at the design basis limit of 120*F at an RHR heat exchanger inlet temperature consistent with the lineup up to a maxunum RHR system flow rate of 4500 gpm, which is the RHR pump runout limit, Reference (28). %e RHR heat exchanger

}. inlet temperature is specified to be consutent with the corresponding Reactor Coolant System temperature for that condition. De RHR heat exchanger conditions for post LOCA containment sump recirculation operations are those identified in References (9) and (14).

He RHR pump oil cooler heat loads are applied for each lineup in which RHR system operation is indicated. Table 2.

) De thermal-hydraulic conditions of the spent fuel pools are based on the estimated heat load which would occur immediately prior and following the refueling outage in the Year 2000 at a Spent Fuel Pool Cooling (SFPC) system mass flow rate of 1.88E6 lbm/hr, Reference (!!), which conservatively results a specified SFPC volumetric flow rate of 3750 gpm. Table 4 summanzes the assumed heat loads for Spent Fuel Pools A/B and C/D as well as the applicable dates as the limiting heat load for each CCW system alignment does not necessarily correspond to operations at the l completion of the Year 2000 outage.

) Refueling case alignment maximum heat loads are identified in Reference (26) for the Normal Full Core Offload scenario. An estimate of Core Shuffle and Abnormal Full Core Offload scenario heat loads is performed to satisfy the analysis requirements of NUREG-0800.

De base heat load for fuel pool A/B is estunated as follows:

Normal Full Core Offload (RFO7) = 35.06 MBTU/hr [ Reference (11))

FuelPool A/B Base Heat Load (RFO7) = 5.16 MBTU/hr [ Reference (11)]

Calculated Refueling Heat Load (RFO7) = 29.9 MBTU/br l Specified fuel pool A/B and C Heat Load = 44.13 MBTU/hr (Attachment 5 ofReference(26)]

FuelPoolC Heat Load = 0.9957 MBTU/hr (Attachment 8 of Reference (26)]

Refueling Heat Load = 29.9 MBTU/hr Estunated Fuel Pool A/B Base Ht Load = 13.23 MBTU/br, use 13.3 MBTU/br for conservatism.

He Core Shuffle refueling alignment heat load of 25.0 MBTU/hr is estimated as follows:

Fuel Pool A/B Base Heat Load As of 9/26/2001 = 13.3 MBTU/hr Fuel Pool A/B Core Shuffle Heat Imad = 11.68 MBTU/hr = 16.84 -5.16 MBTU/hr [ Reference (1l)]

Fuel Pool A/B Core Shuffle Total Heat Load As of 9/26/2001 = 13.3 + 11.68 = 25 MBTU/hr He maximum Abnormal Full Core Offload alignment heat load of 44.1 MBTU/hr is estimated as follows:

Fuel Pool A/B Base Heat Load As of 9/26/2001 = 13.3 MBTU/hr

) Fuel Pool A/B Abnormal Full Core Offload Heat Load = 30.71 MBTU/hr = 35.87 -5.16 MBTU/hr [ Reference Fuel Pool A/B Abnormal Full Core Offload Total Heat Load As of 9/26/2001 = 13.3 + 30.71 = 44.1 MBTU/hr

, Rese heat loads do not include the effect of any change in HNP core thermal power rating.

c O'

Computed by: Date:

JeffLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date:

r8 9 32 '" O 1

Project No.: CALCULATION SHEET pjj*.

O Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis O l l

Table 3 '

Summary of CCWS Operating Alignment Thermal Boundary Conditions l Alignment '

% Norenal Hot S/D Safe S/D Refuel Reheel Refuel LOCA LOCA IDCA (350 7 ) (350 7 ) (Mode 6) (Mode 6) (Mode 6)

Load Umts Mode i Core A

@ 4 lus @ 4 hrs Full Absormal $1 Recut Recirc Reference Shufne Offload (RHR) (RHR and SFP)

RHR Pump B HeatImad 0 0 70,000 N/A N/A N/A 70,000 0 MA N/A (BTU /hr) Calc NSSS 34 R2 RHR Hz B Flow (gprn)/ N/A MA 650/ N/A N/A N/A 0/ N/A N/A N/A Tin (7) 350 Calc NSSS.38 R2CC-140 0038 R0 RHR Pump A Heat Load 0 0 70000 70000 70000 70000 70000 0 70000 70000 Calc NSSS-38 R2 (BTUh)

ItMR Hz A Flow (spm) / N/A N/A O Tin (T) 650/350 800/350 0/140 0/140 0/140 N/A 3903/244.8 3903/209 Calc NSSS-34 R2/CC-0038 R0 BRS: Dist Cir Heat Load N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Assuuned BRS Skad (BTUk) Ab=da==d hiplace BRS: Evap Cir Heat Load N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Assummed Bits Sidd (BTL W)

Ah,da==d impte=

BRS: VentCond Hea: Lead N/A N/A N/A N/A N/A N/A MA N/A N/A N/A Asmanad 8RS Sidd (BTU &)

Akada=ad laplace Letdows Mx Flow (spm) / 120/ 120/ 120/ Secmed 0 0 0 120/ 0 O Tin (v> 380 380 350 380 0 Demga CYCS Flow at aCS Te ,

XSLD HA Flow (spm)/ 24.8/560 N, A N/A N/A N/A N/A N/A N/A N/A N/A Spec Sheetin VM-T.a (T) MRK RCDTHA Flow (spm)/ 89.42/ 89.12/ 89.12/ 89.12/ 89.12/ 89.12/ 89.12/ 89.I2/ 0 0 Spec Sheet im VM.

Tin (T) 180 180 180 680 ISO 880 180 180 MRK Scal Water Hx Flow (spm)/ 128. t/ 128.1/ R28.8/ 128.1/ 128.1/ 128.1/ 128.t/ 128.1/ 0 0 Spec Sheetin VM.

Tin (T) 138.5 138.5 138.5 1383 138.5 138.5 138.5 138.5 SFP Ha A Hea: Land MRK 15200000 15200000 13500000 13500000 25vvvvuv 38780000 38780000 15200000 0 15200000 Esamsted tora O SFP Ha B

(***)

Heat Lead N/A N/A N/A N/A N/A

• 26)

N/A N/A N/A N/A N/A Secured (BW&)

RCP A Heat Land 367000 367000 367000 0 0 0 0 367000 0 0 (BTU &) to of WEC CQl-53616/5n9 Value RCPB Heat Lead 367000 367000 367000 0 0 0 0 367000 0 0 la of WEC CQl-(BTU &) 53616/$n9 Value RCP C Heat Lead 367000 367000 367000 0 0 0 0 367000 0 0 in of WEC CQL-0 SFPitxC (BM W)

Heat Lmed N/A N/A N/A N/A N/A N/A N/A N/A 53616/509 Value N/A N/A Secured (BTU &)

SFP Ha D Heat ised I R--- - 1000000 10vuvuv 10uvuvu 8000000 1000000 1000000 1000000 0 1000000 Estanated tese (BTU &) N/A OFFD Refersace(26)

Hea: Lead N/A N/A N/A N/A N/A N/A N/A N/A 0 0 Assunted Neghgible.

(BWh) H Load = 0J4 Sanp6e Coolers MBTUk g Heat Lead N/A N/A N/A N/A N/A N/A N/A N/A 0 0 Assumuod Neghgible s/ (B n W ) Due toTrandent Land FuelPoot A/B Heat Land 15200000 15200000 13500000 13500000 25000000 31780000 3t780000 15200000 15200000 15200000 Estamoemd tman (BTUk)

Fuel PoolC/D Reference (26)

Heat Lead 1000000 1000000 1000000 4000000 1000000 1000000 1000000 1000000 1000000 1000000 Estunneed bara (BTU &)

i Ref-(26)

CCW 7:ams No Operating i Split (t/t) i I I Spin (1/1) i Spht(I/0) Split (t/0) Consessent with DBD13t O

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)' Computed by: DIte:

JeffLundy - CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

'810d 32 "" O Project No.: CALCULATION SHEET p 3,;

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis 3 Table 4 Summary of Spent Fuel Pool A/B Heat Loads for Various CCW System Alignments Ahgnment As of Date CCW SFP A/B SFP A/B Temperature Ten , heat Lead Lirnit (*F) Limit (*F) (MBTU/hr)

) Normal 10/22/2001 105 137 15.2 Hot S/D (350) 9/15/2001 120 137 13.5 Safe S/D(350*F) 9/15/2001 120 137 13.5 Refuel-Core Shuffle 9/22/2001 105 137 25.0 Refuel-Normal Full Core OfDoad (1) 9/22/2001 105 137 31.78 Refuel-Abnormal Full Core Offload (1) 9/22/2001 105 137 31.78 g' LOCA Safetyinjection 10/22/2001 105 137 15.2 LOCA-Recirc (RHR Only) 10/22/2001 120 137 15.2 LOCA-Recirc (RHR/SFP) 10/22/2001 120 137 15.2 Notes: 1) Assumes that 265.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> have elapsed since reactor shutdown to

) reduce core decay heat to within the heat removal capacity of the SFP heat exchangers.

4.3 Evaluation of Minimum RHR Heat Exchanger CCWS Flow The po.st-modification CCW flow balance evaluated in this analysis maintains a maximum design CCW

] temperaiure of 120*F, while considering the addition of 1.0 MBTU/hr to the C and D Spent Fuel Pools,6 percent modeling uncertainty per Reference (1), and a RHR heat exchanger UA value which is modeled to change with fluid properties. The licensing basis previous to this calculation is based on an assumed RHR heat exchanpr UA of 1.635E6 BTU /hr *F, derived from the design RHR heat exchanger overall heat transfer coefficient of 382 BTU /hr-sqft *F which is in turn based ou an RHR heat exchanger inlet temperature of 139*F 4 and the overall heat transfer surface area of 4280 sqft. However, during the initial phase of cont == ment sump

] recirculation, the RHR tube side inlet temperature rises to 244.I'F, which increases the calculated overall heat transfer coefficient to 421.2 BTU /ht 'F due to the change in the RHR heat exchanger tube side fluid vtscosity.

4 These conditions would tend to increase heat transfer through the RHR beat exchanger and increase CCW j system supply temperatures above the maxtmum CCW supply temperature of 120*F for the given limiting  !

conditions of minimum CCW best exchanger Service Water flow and maximum Service Water supply temperaiure.

) Two changes are prescribed herein to address the heat loads and conditions above in the post modification CCW flow balance. First, the rmmmum specified CCWS flow to the RHR heat exchanger must be reduced to a level consistent with heat rejection value of 111.1 MBTU/hr, consistent with Reference (9). An analysis of i

RHR heat exchanger thermal performance, Attachment (C), was performed to determine the minimum shell

{ side tiow rate at 120*F shell side inlet temperature, 244.I'F tube side inlet temperature and 1.846E6 lbm/br

} tube side flow rate, consistent with Reference (21). His analysis shows that a mmimum CCWS flow rate of

'4874 grm at 120*F is required at the beginning of the sump recirculation phase. The specified CCWS flow to the Rill >. heat exchanger under these conditions, assuming 6 percent modeling uncertainty consistent with

)

1

g. CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 I l Checked by: Date:

e, g ; .'32 '" O p Project No.: CALNATION SHEET p;ge Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal. Hydraulic Analysis Reference (1), is 5166 gpm or approximately 5200 gpm. As the containment sump temperature decreases, the minimum required CCWS tlow also decreases, as shown in Figure 1 of Attachment (C), based on maintaining a maximum RHR heat exchanger tube side outlet temperature of 180*F, Reference (21). %e CCWS was in rebalanced using the CCWS PROTO-FLO" model in the LOCA:Recire (RHR Only) alignment, Attachment (F), with a 10 percent degraded CCW pump curve, by adjusting ICC-146 to 47.9 percent open. When the nominal CCW pump curve is applied to the previously balanced CCWS, CCWS flow to the RHR heat exchanger increases to approximately 5440 gpm resulting in an increased RHR heat exchanger heat duty of 118 MBIU/hr. He increased RHR heat exchanger heat duty results in an excessive CCWS supply temperature j which cannot be maintained below 120*F, given 8250 gpm ESWS flow to the CCW heat exchanger. Holding the position of ICC-146 (or ICC-166) constant, the specified ESWS flow to the CCW heat exchanger was increased to 8500 gpm which results in a CCW beat exchanger outlet temperature of 120*F, Attachment (G),

O consistent with the original assumption used in setting the minimum CCWS flow to the RHR heat exchanger, documented in Attachment (D).

Herefore, a reduction in the muumum specified RHR heat exchanger CCWS flow to 5200 gpm from the original 5600 gpm specification and an increase in the nummum specified CCW heat exchanger ESWS flow to 8500 gpm from the original 8250 gpm are necessary to meet all the thermal-hydraulic assumptions which are O used in the HNP Containment Analysis, Reference (21). A minimum specified ESWS flow of 8500 gpm to the CCW heat exchangers was verified to be within the capacity of the current ESWS system, Reference (20), even considering the most limiting ESWS single failure of a MCC IB35-SB feeder breaker failure, Reference (29).

4.4 . Evaluation of Maximum RHR Heat Exchanger CCWS Flow '

An evaluation was performed, using the RHR heat exchanger PROTO-HX" model developed in Reference (1),

g to estimate the maximum CCWS flow rate which could be accommodated during the initial phase of containment sump recirculation. This analysis shows that a maximum CCWS flow of 5220 gpm is attainable for a CCW heat exchanger ESWS flow of 8250 gpm and a maximum CCWS flow of 5440 gpm is attainable for an ESWS flow of 8500 gpm in order to maintain a CCWS supply temperature of 120*F. Given that the RHR heat exchanger throttle valves (ICC-146 and ICC-166) are set on the basis of maintaining a mmimum CCWS flow rate under all hydraulic conditions, including modeling uncertainty and CCW pump degradation limits, O when the CCWS is in the LOCa recirculation alignment, there will be excess flow to the RHR heat exchanger, approximately 5440 gpm total, Attachment (D). He thermal effect of the excess RHR heat exchanger flow can be mitigated with an increase in the nunimum ESWS flow to the CCW heat exchanger of 250 gpm.

4.5 Evaluation of Minimum Spent Fuel Pool Heat Exchanger CCWS Flow An evaluation of the minimum thermally required CCWS flow to the Spent Fuel Pool heat exchangers was O. performed by generating heat duty versus CCwS flow for mil combinations of design CCwS suppiy temperatures and SFP temperature limits. His analysis is performed using the PROTO-HX" model developed in Reference (1) and assumes 5 percent tube plugging and design fouling factors. CCWS desip supply

! temperatur.s of 105'F for normal and refueling system alignments and 120*F for cooldown and LOCA: Recirculation alignments are used in the analysis. A maximum SFP temperature limit of 137*F for all fuel pool operations is also assumed. Figure I and Table 5 summarize and Attachment (E) documents the results of this analysis.

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! Coupted by: D:ts:

JeffLundy CAROLINA POWER & LIGHT' COMPANY Calculation ID: SF-0040 Checked by: Date:

'8 12 d 32 '" O PT lect N "

O File:

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis iO -

Figure 1 l

SFP Hx Duty at a Given CCW Flow Rate 35 l 30 ' _M, O if 25 ur E

,w s ,/

O li 15 ' u 7

$ ( A>'

.O i 10 ' >

b s. f/ + FP Terrp = 137F :xlW Tony = 120F '

+ FP Tene e 137F,cCWTene e 105F Oi i i O 1000 2000 3000 4000 5000 6000

,O o =iva'r-aasracia o ooio4 a< a+reeru CCW Flow (gpm) 5% Tube Pbggstg Table 5 Minimum SFP Heat Exchanger CCW Flow Requirements

.O SFP Hx A/B SFP Hx A/B SFP Hx C/D SFP Hx C/D Thermal Flow Minimum Thermal Flow Minimum Requirement Flow (1) Requirement Flow (1)

Alignment As of Date (gpm) (gpm) (gpm) (gpm)

Normal 10/22/2001 1200 1272 60 63.6 Hot S/D (350F) 9/15/2001 2800 2968 125 132.5 O sa,e Slo (350e) s,35/2002 2800 2968 125 132.5 Refuel-Core Shuffle 9/22/2001 2800 2968 60 63.6 Refuel-Normal Futi Core Offload (2) 9/22/2001 5400 5400 (3) 60 63.6 Refuel-Abnormal Full Core Offload (2) ~9/22/2001 5400 5400 (3) 60 63.6 LOCA-Safoty injectum 10/22/2001 1200 1272 60 63.6 LOCA-Rearc (RHR Only) 10/22/2001 0 0 0 0

,g LOCA-Reorc (RHR/SFP) 10/22/2001 3830 4059.8 125 132.5 Note 1: Minimum Heat Exchanger Flow indudes 6% Hydraulic Uncertainty Per CP&L HNP Calculation CC-0039 j Revision 0 j

Note 2: Assumes Sufficient Decay Time to Reach 31.78 MBTU/hr (265.36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after S/D) -

Note 3: SFP Hx A/B Max Flow is 5400 ppm per design data sheet which should not be exceeded to ensure now l induced tube vibraton problems do not occur.

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O Computed by: Date:

JeffLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Onecked by: Date:

c 13 d32 "" O O Project No.: CALCULATION SHEET Fik-Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis

.O 4.6 CCWS Hydraulic Margins in order to accommodate the changes in the CCWS load flow requirements identified 6bove, the CCWS PROTO-FLO" model was rebalanced. Based on previous analysis, it was determira:4 that the most limiting CCWS alignment is the Hot S/D (350F) 'A' CCW Train with the Nonessential header case in wiuch the CCW pump develops the least head due to the maximum CCWS flow requirernents. Herefore, the CCWS was O rebalanced using the Hot S/D (350F) alignment, the 10 percent degraded CCW pump curve and mimmum CCWS flows to each load with the exception of the RHR heat exchangers which were h-i-~*8 in the LOCA:Recire (RHR Only) alignment. He results of this analysis are documented in A**h= rats (F) and (H).

He resulting changes in throttle valve position or miscellaneous loss coefficients are shown in Table 6. It is noted that the SFP heat exchanger C/D throttle valves, AltV-15 and AltV-II, are heavily throttled and will require a suitably sized bypass line with a smaller throttle valve in order to achieve acceptable throttling

-Q characteristics. His modification to the CCWS retum line from SFP heat exchangers is a design detail wiuch will have to be resolved at a later date by the cognizant design organization.

Table 6 Estimated Change in CCWS Throttle Valve Positions and RCP Line Miscellaneous Loss Coefficients O

Service Brottle Valve Old Position / Misc K New Position / Misc K RHR Heat Exchanger A ICC-146 49.24 % Open 48.61 % Open RHR Heat Exchanger B ICC-166 49.24 % Open 47.91 % Open RCDT Heat Exchanger 1CC-187 8.85 % Open 41.98 % Open XSLD Heat Exchanger ICC-197 12.91% Open 80.23% Open O SFP Heat Exchanger A(B) 1CC-382(398) 34.35 % Open 27.94% Open SFP Heat Exchanger D(C) AltV-15(11) Not lastalled 2.03 % Open RCP A Upper Bearing Cooler ICC-258 194.00 14.14 RCP A Lower Bearing Cooler ICC-264 90000 11971 RCP A Hermal Barrier ICC-224 510.00 58.28 RCP B Upper Beanng Cooler ICC-273 211.00 O RCP B 14wer Beanng Cooler 14.14 ICC-279 30584.00 11971 RCP B hermal Barrier ICC-235 320.00 58.28 RCP C Upper Bearing Cooler ICC-284 206.00 14.14 RCP C Lower Beanng Cooler ICC-290 80610.00 11965 RCP C hermal Barrier 1CC-246 404.00 52.87 O

He hydraulic margins for the CCWS were evaluated utilizing the system throttle valve positions documented in Attachments (F) and (H) and degrading the operating CCW pump curves by 10 percent of the total developed head, Reference (3). The effect of the letdown heat exchanger was simulated by changing ICC-TCV-337 to a flow control valve with a setpoint of 610 gpm per Assumption 4.1.16, the specified letdown heat Q' exchanger CCWS flow rate under non-startup conditions with hydraulic uncertainty. For the plant Startup case aliganwnt, ICC-TCV-337 was restored to a temperature control valve with a setpoint of 120*F. ESWS flow to the operating CCW heat exchangers was set to 8500 gpm at 95'F. The resulting CCWS flows were tabulated and reduced by 6 percent to account for modeling and instrument uncertainty as established in Reference (1).

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D y CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date:

'* 14 d32 "" O Project No.: CALCULATION SHEET g pg; Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project hennal-Hydraulic Analysis g his process is repeated for each of the major CCWS system lineups with the results documented in Attachments (I) through (Q) and suuuund in Tables 7a through 7j.

Minimum CCWS flow requirements in Tables 7a through 7j to cooled components are established from design data sheet values, Reference (1), for all components except for the RHR Pump Coolers, Reference (9), the RHR heat exchangers, Section 4.3, the Letdown heat exchanger, Reference (31), and the Spent Fuel Pool heat 3 exchangers, Section 4.5.

The Hot Shutdown (350F) and Safe Shutdown (350F) RHR heat exchanger nummum CCW flow limits were determined, Attachments (J) and (K), by using the RHR heat exchanger PROTO-HX" model, Reference (1), to meet a heat duty of 118.9 MBTU/hr and 177.76 MBIU/hr with the maximum RHR pump flow of 4500 gpm at an RCS temperature of 350*F. The Hot Shutdown case required heat duty of 118.9 MBTU/hr is determined as g follows:

RCS Sensible Heat Removal = 66.96 MBTU/hr [ Table 9.2.1-7 of Reference (31)]

Decay Heat 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after S/D = 110.8 MBTU/hr (Table 9.2.1-7 of Reference (31)]

RCP Heat (3 Pumps Operating) = 60 MBTU/hr [ Reference (32)]

Total Heat Removal Required = 237.76 MBTU/hr with 2 RHR/CCW aains in operation g or = 118.9 MBTU/hr per RHR heat exch ,ger The Safe Shutdown case RHR heat exchanger required heat duty of 157.76 MBTU/hr is taken from Table 9.2.1-7 of Reference (31). The nummum required CCW flow to the RER heat exchanger, assuming design fouling factors and 120*F CCW supply temperature is 1300 gpm and 2560 gpm per operating heat exchanger for the Hot Shutdown and Safe Shutdown cases, respectively.

D The LOCA: Recire (RHR and SFP) case represents maintaining CCWS flow to both the Spent Fuel Pool and RHR heat exchangers. It is assumed that the operators do not adjust flow to the Spent Fuel Pool heat exchanger in order to maintain RHR heat exchanger flow as the estimated CCWS flow to the RHR heat exchangers exceeds the thermally required CCWS flow of 2250 gpm for a containment sump temperature of 209'F, Attachment (C), to maintain a 180*F RHR heat exchanger outlet temperature, Reference (21). When the g containment sump temperature reaches 209'F, the CCWS flow to the RHR heat exchanger is 4450 gpm, Attachment (Z). The worst case CCWS flow to the RHR heat exchanger at the point of bringing the Spent Fuel Pool heat exchangers online is 4430 gpm with a corresponding heat removal of 80.53 MBTU/hr, consistent with Reference (14).

The results of this analysis, Tables 7a through 7j, show that sufficient CCWS flow is available to cooled g components under most major system alignments, given the 10% IST pump degradation ihaits assumed by Reference (3), with the exception of the Spent Fuel Pool heat exchanger A (or B) under the LOCA:

Recirculation (RHR and SFP) alignment and the nonessential header loads under the Refuel-Nonnal (or Abnormal) Full Core Offload case. Evaluation of the system thermal analysis results during the LOCA: Recirculation (RHR and SFP) alignment, Attachment (Z), shows that the steady state equilibrium temperature of fuel pool A/B does not exceed 136*F, even with the assumptions of 10% percent degraded g CCWS flow, minimum ESWS flow of 8500 gpm to the CCW heat exchangers, use of design fouling factors for all heat exchangers and design (maximuin) Ultimate Heat Sink temperature of 95'F. Acceptable fuel pool A/B temperature indicates that the minimum specified CCWS flows to the Spent Fuel Pool heat exchangers are very h

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ute by: Date  !

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

es g3w32 '" O D Project No.: CALCULATION SHEET p;g,;

)

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis O

conservative and that acceptable operation of the Spent Fuel Pool heat exchangers under limiting conditions can be achieved. These results demonstrate that the redistribution of CCWS flow is adequate for the limiting CCW pump developed head.

For the Refuel-Normal and Abnormal Full Core Offload cases in which a single failure of the 'B' CCW train or j

O when the CCW trains are split, insufficient CCW flow is provided to the SFP heat exchanger A (or B), the Seal i

Water heat exchanger and the RCDT heat exchanger for the limiting hydraulic case of 10% degraded CCW l pump curve operation. A separate heat exchanger performance analysis was done for each heat exchanger assuming all other thermal conditions were specified as design values except for the CCW flow and supply tempenture, documented in Attachments (M) and (N). He results of this analysis indicates that the SFP Hx A (or B) can just accommodate an assumed full core offload heat load of 31.7 MBTU/hr at design SFPC thermal i g conditions, therefore the negative CCW flow margin is acceptable under these extreme thermal-hydraulic conditions.

He results of the Seal Water heat exchanger performance analysis show that the estimated heat duty is 3.1 percent less than the design heat duty but this is judged to be acceptable as the reactor coolant pumps are not operating during refueling operations and the seal injection supply temperature only rises from 115.0 to

""' j O

he results of the RCDT heat exchanger perfonnance analysis show that the estimated heat duty is 0.9 percent less than the design heat duty of the heat exchanger, resulting in an increase in RCDT heat exchanger outlet temperature from 130.0 to 130.5*F. It is considered that this small increase in RCDT heat outlet temperature is acceptable as RCS temperature is less than 140*F during this operating mode while the design RCDT heat exchanger inlet temperature is 180*F.

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• JeffLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date: .

4 16"'32 "" O Project No.: CALCULATION SHEET p;3, 1

Project

Title:

Spent Fuel Pools C and D Activation Project J

i Calculation

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Spent Fuel Pools C and D Activation Project'Ibennal-Hydraulic Analysis

'O Table 7a Sununary of CCWS Flow Margins Normal Ops Section Component Calculated Flow With 6% len Flow Flow Marge (%)

(gpm) Uncertainty (gpm)

. 105 RHR Pump B 6.9 6.5 5 30 %

1 108 RHRHxB 36 34.0 0 MA 115 RHR Pump A 7 6.6 5 32%

112 RHRHxA 36.9 34.8 0 NA 66 BRS: Dist Car 0 0.0 0 MA O

80 BRS: Evap Ctr 0 0.0 0 NA 73 BRS: Vent Cond 0 0.0 0 N/A 61 Letdown Hx 1158 1092.5 575 90%

45 XSLD Hx 318 300.0 247 21 %

44 RCDTHx 303 285.8 225 27%

54 Seal Water Hx 308 290.6 230 26% l 98 SFP Hx A 3613 3408.5 1200 184 %

91 SFP Hx B N/A N/A. O MA 204 RCP A Upper 01 Cooler 1942 1832 150 22%

203 RCP A Lower 01 Cooler 6.7 6.3 5 26%

O 205 RCP A Thermal Barrier 51.6 48.7 40 22%

208 RCP B Upper Of Cooler 194.2 1832 150 .22%

207 RCP B Lower Oil Cooler 6.7 6.3 5 26 %

209 RCP B Thermai Barrier 51.6 48.7 40 22%

214 RCP C Upper 01 Coo 6er 1942 1832 150 22%

Q 212 RCP C Lower Of Cooler 6.7 6.3 5 26 %

215 RCP C Thermal Banier 51.6 48.7 40 22%

304 SFP Hx C N/A - 0 N/A 305 SFP Hx D 160.9 151,8 59 157%

Node 0401 GFFD 14 - 14 Specified N oe 28 Sampie Cooiers 160 160 Specmed O -

5 and 28 Total CCWS Flow 6887 64972 3305 97 %

Operating CCW Train A A l

Notes RHR Hx Outlet isolaton Valves are Shut LD HX Flow is set based on maintaming 120F LD Outlet Temp.

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l e l y gr y CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

n g7 g32 "" O Project No.: ON SHEE pig,;

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project 'Ibermal-Hydraulic Analysis O

Table 7b Summary of CCWS Flow Margins Hot S/D (350F) g Section cwt @,ent Calculated With 6% Min Flow (gpm) Flow Margin (%)

Flow (gpm) Uncertainty 105 RHR Pump B 7.2 6.8 5 36 %

108 RHR Hx B 5199 4904.7 1300 2 77 %

115 RHR Pump A 5.6 5.3 5 6%

O 112 RHR Hx A 3983 3757.5 1300 189 %

66 BRS: Dist Car 0 0.0 0 N/A 80 BRS: Evap Car 0 0.0 0 N/A-73 BRS: Vent Cond 0 0.0 0 N/A 61 Letdown Hx 610 575.5 575 0%

g 45 XSLD Hx 262 2472 247 0%

44 RCDT Hx 249.5 235.4 225 5%

$4 Seal Water Hx 253.9 239.5 230 4%

98 SFP Hx A 2980.3 2811.6 2800 0%

91 SFP Hx B -

0 N/A g 204 RCP A Upper Oil Cooler 160 150.9 150 1%

203 RCP A Lower Oil Cooler 5.5 5.2 5 4%

205 RCP A Thermal Bamer 42.5 40.1 40 0%

208 RCP B Upper Oil Cooler 160 150.9 150 1%

207 RCP B Lower Oil Cooler 5.5 5.2 5 4%

209 RCP B Thermal Barrier 42.5 40.1 40 0%

214 RCP C Upper Oil Cooler 160 150.9 150 1%

212 RCP C Lower Oil Cooler 5.5 52 5 4%

215 RCP C Thermal Bamer 42.5 40.1 40 0%

304 SFP Hx C N/A 0 N/A 305 SFP Hx 0 132.7 1252 125 0%

Node 0401 GFFD 14 -

14 Spedfied Node 0028 Sample Coolers 160 160 Speci6ed f 5 and 28 Total CCWS Flow 14529.3 13706.9 7571 81 %

Operating CCW Train A/B Split A/B Split Notes LD Hx Flow Lirnted to a 575 GPM Nominal Value Denned in FSAR D

D

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Computed by: Date:

JeELundy CAROLINA POWER & LIGIIT COMPANY Calculation ID: SF-0040 l Checked by: Date:

'8 18 d 32 "" O Project No.: CALCULATION SHEET g pg i Project

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Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis O

Table 7e Summary of CCWS Flow Margins Safe S/D (350F)

S6Cuon Cuivenait Calculated Flow Wrth 6% Men Flow (gprn) Flow Margen (%) 'l g (gpm) Uncertainty j

105 RHR Pump B 0 0.0 0 N/A 108 RHR Hx B 0 0.0 0 N/A j 115 RHR Pump A 5.8 5.5 5 9% )

112 RHR Hx A 4119 3885.8 2560 52% 1 3 66 BRS: Dist Car 0 0.0 0 N/A 80 BRS: Evap Ctr 0 l

0.0 0 N/A J 73 BRS: Vent Cond 0 0.0 0 N/A 61 Letdown Hx 0 0.0 0 N/A 45 XSLD Hx 271.6 256.2 247 4% 1

] 44 RCOT Hx 258.8 244 2 225 9% j 54 Seal Water Hx 263.5 248.6 230 8%

1 98 SFP Hx A 3096 2920.8 2800 4%

91 SFP Hx B - -

0 N/A I

204 RCP A Upper Oil Cooler 165.9 156.5 150 4% '

20s RCP A Lower Od Cooler 5.7 5.4 5 8%

205 RCP A Thermal Bamer 44.1 41.6 40 4% l 208 RCP B Upper Od Cooler 165.9 156.5 150 4%

207 RCP B Lower Od Cooler 5.7 5.4 5 8% l 209 RCP B Tbiii.; Barrier 44.1 41.6 40 4% I 214 RCP C Upper Od Cooler 165.9 156.5 150 4%

212 RCP C Lower O0 Cooler 5.7 5.4 5 8%

215 RCP C Thermat Bamer 44.1 41.6 40 4%

304 SFP Hx C N/A -

0 N/A 305 SFP Hx D 137.9 130.1 125 4%

Node 0401 GFFD 14 -

14 Specded O Node 0028 Sample Coolers 160 -

160 Specmed 5 and 28 Total CCWS Flow 9003 8493.4 6951 22 %

Operating CCW Train A A a) RCPs are secured but CCWS now is maintained.

g b) Letdown secured g c) *B' CCW Train Single Failure 9

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• JeffLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date:

4 1932 "" O

, Project No.: CALCULATION SHEET pg,.

J Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis 0

Table 7d Summary of CCWS Flow Margins Refuel Core Shufne Sects Component Calculated With 6% Min Flow (gpm) Flow Margin (%) l Q Flow Unwrtainty (gpm) 105 RHR Pump B 0 0.0 0 N/A 108 RHRHxB 0 0.0 0 MA 115 RHR Pump A 5.8 5.5 0 N/A 112 RHRHxA 4125 3891.5 0 MA 1 66 BRS: Dist Cir 0 0.0 0 N/A I l

80 BRS: Evap Ctr 0 0.0 0 N/A j 73 BRS: Vent Cond 0 0.0 0 MA j 61 Letdown Hx 0 0.0 0 N/A 45 XSLD Hx 272 256.6 247 4%

) 44 RCDTHx 259.3 244.6 225 9%

54 Seat Water Hx 264 249.1 230 8%

98 SFP Hx A 3103 2927.4 2900 1%

91 SFP Hx B - -

0 MA 204 RCP A Upper Oil Cooler 166.3 156.9 150 5%

] 203 RCP A Lower Oil Cooler 5.7 5.4 5 8%

205 RCP A Thermal Barrier 44.1 41.6 40 4*4 208 RCP B Upper Oil Coo 6er 166.3 . 156.9 150 5%

207 RCP B Lower Oil Cooler 5.7 5.4 5 8%

209 RCP B Thermal Banier 44.1 41.6 40 4%

214 RCP C Upper Oil Cooier 166.3 156.9 150 5%

212 RCP C Lower Oil Cooler ST 5.4 5 8%

215 RCP C Thermal Bamer 44.1 41.6 40 4%

304 SFP Hx C t#A -

0 MA 305 SFP Hx D 13b 2 130.4 59 121 %

, Node 0401 GFFD 14 - 14 Speofied d Node 0028 Sample Coolers 160 -

160 Sperf.ed 5 and 28 Total CCWS Flow 8997.6 8488.3 4420 9.I 4 Operating CCW Train A A a) RCPs are secured but CCWS Row is maintained.

Notes b)'B' CCW Train Single Failure

$ c) No min now is defined for the RHR hx as RPV is defueled I

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.O c---- -ted by- Date jeg[ undy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

's20 #32 "" O Project No.: CALCULATION SHEET Fik-Project

Title:

Spent Fuel Pools C and D Activation Project C=Wlation

Title:

Spent Fuel Pools C and D Activation Project 'Ibermal-Hydraulic Analysis O Tabie 7 -

Summan of CCWS Flow Margins Refusi- Normai Core OfRoad Seccon Cw . ,~. - a Calculated Flow With 6% Min Flow Flow (gpm) Uncertainty (gpm) Margin (%)

0 15 RHR Pump B 0 0.0 0 WA 108 RHRHxB )

0 0.0 0 WA 115 RHR Pump A 4.9 4.6 0 .MA 112 RHR H:: A 3470 3273.6 0 WA 66 BRS: Dist Cir 0 0.0 0 N/A 80 BRS: Evap Cir 0 0.0 0 MA 73 BRS: Vent Cond 0 0.0 0 NA 61 Letdown Hx 0 0.0 0 MA 45 XSLD Hx 224 211.3 0 N/A 44 RCDTHx 213.8 201.7 225 -10%

54 Seal Water Hx 217,1 204.8 230 -11%

O 98 SFP Hx A 5325.9 5024.4 5400 -7%

91 SFP Hx B -

0 N/A 204 RCP A Upper 00 Cooler 137.1 129.3 0 N/A 203 RCP A Lower ON Cooler 4.7 4.4 0 N/A 205 RCP A Thermel Santer 36.3 34.2 0 NA Q 208 RCP B UpperOR Cooler 137.1 129.3 0 N/A 207 RCP B Lower OR Cooler 4.7 4.4 0 MA 209 RCP B Thommi Banter 36.3 34.2 0 N/A 214 RCP C Upper OR Cooler 137.1 129.3 0 N/A 212 RCP C Lower ON Cooler 4.7 4.4 0 N/A 0 215 RCe C Thermai Bamer 36.3 34.2 O N/A 304 SFP Hx C N/A 0 N/A 305 SFP Hx 0 110.6 104.3 60 74 %

Node 0401 GFFD 14 - 14 Specified Node 0028 S % Coolers 160 150.9 160 Specited 5 and 28 Totai CCWS Flow 10285 9702.8 6089 59 %

Cm us CCW Train A A a) RCPs are secured but CCWS flow is maintained.

b) W CCW Train Single Failure  !

c) No min flow is deRned for the RHR hx as RPV is defueled Notes d) SFP A/B hx CCW set to 5400 gpm j O' e) RCOT, Seal Wtr Hx and SFP AS Hx performance exceeds the design requirements f) No rmn flow is defined for XSLD Hx as LD is secured (

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JeELundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

'821 # 32 "" O I l

Project No.: CALCULATION SHEET pgj )

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis O

Table 7f l Summary of CCWS Flow Margins Refuel Abnormal Core Omoad Sechon CGiwent Calculated Flow With 6% Men Flow (gpm) Flow Margin (%)

(gpm) Uncertainty O

105 RHR Pump B 72 6.8 5 36 %

108 RHR Hx B 5213 4917.9 0 MA 115 RHR Pump A 4.9 4.6 5 -8%

112 RHR Hx A 3470.3 3273.9 0 MA 66 BRS: Dist Car 0 0.0 0 MA O e0 BRS: evao Cir 0 0.0 0 MA 73 BRS: Vent Cond 0 0.0 0 NA 61 Letdown Hx 0 0.0 0 N/A 45- XSLD Hx 224.1 211.4 0 N/A 44 RCOT Hx 213.8 201.7 225 -10%

O 54 Semi water Hx 217.1 204.8 230 -11%

98 SFP Hx A 5326 5024.5 5400 -7%

91 SFPHxB . -

0 MA 204 RCP A Upper Oil Cooler 137.1 129.3 0 N/A 203 RCP A Lower Oil Cooier 4.7 4.4 0 NA g 205 RCP A Ti+We Barrier 36.3 34 2 0 N/A 208 RCP B Upper Oil Cooler 137.1 129.3 0 N/A 207 RCP B Lower Oil Cooier 4.7 4.4 0 MA 209 RCP B Tiewe Barrier 36.3 34.2 0

^

N/A 214 RCP C Upper Oil Cooier 137.1 129.3 0 MA 212 RCP C Lower Oil Cooier 4.7 4.4 0 MA 215 RCP C Thermal Bamer 36.4 34.3 0 N/A 304 SFP Hx C N/A 0 MA 305 SFP Hx D 110.6 104.3 59 77 %

Node 0401 GFFD 14 -

14 Specified Node 0028 S .v;e Cooiers 160 -

160 Specined O 5 and 28 Total CCWS Flow 156052 14627.5 6098 140 %

Operating CCW Train A/B A/B a) RCPs are sea # red but CCWS is maintained.

b) No min flow is defined for the RHR hx as RPV is defueled c) SFP A/B hx CCW set to 5400 gpm Q d) RCOT, Seat Wtr Hx and SFP A/B Hx performance exceeds the design requirements e) No men flow is defined for XSt.D Hx as LD is secured

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c ----ted by- Dite 3,g[ undy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Oecked by: Date:

422 d32 '" O Project No.: CALCULATION SHEET p Project

Title:

Spent Fuel Pools C and D Actwation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project 'Ibennal-Hydraulic Analysis O

Table 7g i 1

Sununary of CCWS Flow Margias LOCA-Safetylagocelon Secton C .w a Nm Flow Witi6% Min Flow (gpm) Flow Margin O (opm) unoeri.ney (%)

105 RHRPwrpB 6.9 6.5 5 30%

106 RHR Hx B 36.3 34.2 0 NA 115 RHR Pwrp A 7 6.6 5 32 %

O -"* """"*^ 7 ** "^

66 BRS: Dist Cir 0 0.0 0 NA 80 BRS: Evap Car 0 0.0 0 NA 73 BRS: Vent Cond 0 0.0 0 NA 61 Letdown Hx 1145 10002 575 88%

45 XSLD Hx 321 302.6 247 23%

O 44 RCDTHx 305.s 2es.5 225 2s%

54 Seal Water Hx 310.5 292.9 230 27%

96 SFP Hx A 3641.6 3435.5 1200 186 %

91 SFP Hx B 0 NA 204 RCP A Upp -08 Cooier 196.1 185.0 150 23 %

O 203 RCe A to.or 08 Coni.r 6.7 6.3 5 26%

205- RCP A Thermal Santer 52.1 49.2 40 23 %

206 RCP B Ui p.i 08 Cooler 196.1 185.0 150 23 %

-207 RCP B Lower 08 Cooler 6.7 6.3 5 26%

209 RCP S Thermal Barrier 52.1 49.2 40 23 %

Q.~ 214 RCP C U,,,, 08 Cooler 196.1 185.0 150 23 %

212 RCP C Lowere Alor 6.7 6.3 5 26% l 215 RCP C Thert ,s "Jarr6er 52.1 492 40 23 %

304 SFP Hx C MA -

0 WA 305 SFP Hx D 162.2 153.0 59 159 %

'O " ' ' "" '

Node 0028 Sample Coolers 0 0.0 0 isoisted 5 and 28 Total CCWS Flow 6746 6364.2 3131 103%

C,, e.v CCW Train A A Notes a) System configuramon is L...- " _ , after 'S' S@ial O ,

O  !

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) Cumy ted by: Date:

JeffLundy CAROLINA POWER & LIGHT COMPANY Cticulation ID: SF-0040 Checked by: Date:

423d32 " O Project No.: CALCULATION SHEET p;;

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Dermal-Hydraulic Analysis l

Table 7h Summany of CCWS FlowMargins I LOCA- Sump Recirc (RHR Ordy)

S.L. Cvi.w it

~

Calculated Flow With 6% Min Flow Flow Margm (%)

(gpm) Uncertainty (gpm) 105 RHR Pump B 0 0.0 0 N/A I 108 RHR Hx B 0 0.0 0 MA 115 RHR Pump A 7.3 6.9 5 37%

112 RHR Hx A 5193 4881.4 4874 0%

66 BRS: Dist Car 0 0.0 0 N/A 80 BRS: Evap Cir 0 0.0 0 N/A 73 BRS: Vent Cond 0 0.0 0 N/A 61 Letdown Hx 0 0.0 0 MA 45 XSLD Hx 0 0.0 0 MA 44 RCDT Hx 0 0.0 0 N/A

! 54 Seal Water Hx 0 0.0 0 MA

) 98 SFP Hx A 0 0.0 0 N/A 91 SFP HxB - -

0 MA 204 RCP A Upper OE Cooler 0 0.0 0 N/A I 203 RCP A Lower OG Cooler 0 0.0 0 N/A 205 RCP A Thermal Barrier 0- 0.0 0 N/A

) 208 RCP B Upper OG Coo 6er 0 0.0 0 MA 207 RCP B Lower Os Cooler 0 0.0 0 N/A 209 RCP B Thermal Bamer 0 0.0 0 N/A 214 RCP C Upper OG Cooler 0 0.0 0 N/A 212 RCP C Lower Oil Cooier 0 0.0 0 N/A 215 RCP C Thermal Bamer 0 0.0 0 N/A 304 SFP Hx C N/A N/A 0 N/A 305- SFP Hx D N/A N/A 0 N/A Node 0401 GFFD 0 -

0 Isolated I Node 0028 Sarnple Coolers 0 - O lsolated 5 and 28 Total CCWS Flow 5238 4923.7 4879 1%

Operating CCW Train A(Spet) A (Split)

Notes a) Only operator acnon is splitting CCW Trams l

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• JefrLundy CAROLINA POWER & LIGHTCOMPANY Calculation ID: SF-0040 .

Checked by: Date: i m 24 # 32 "" O  !

Project No.: CALCULATION SHEET pg l Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project 'Ibermal-Hydraulic Analysis l O Tabie71 Sunuusry of CCWS Mow Margins LOCA Sump Recirc (RHR/SFP)

Sechon Component C* "1 With 6% Mm Flow (gpm) Flow Margin Flow (gpm) Uncertainty (%)

105 RHR Pump B 0 0.0 0 NA 100 RHR Hx B 0 0.0 0 N/A 115 RHR Pump A 6.3 5.9 5 18%

112 RHRHxA 4472 4203.7 2250 87 %

66 BRS: Dist Car 0 0.0 0 N/A l O 80 BRS: Evap Car 0 0.0 0 NA 73 BRS: Vent Cond 0 0.0 0 NA '

61 Letdown Hx 0 0.0 0 NA 45 XSLD Hx 0 0.0 0 NA 44 RCDT Hx 0 0.0 0 NA O 54 Seal Water Hx 0 0.0 0 N/A 96 SFP Hx A 3381.5 3178.6 3830 -17%

91 SFP Hx B - - 0 N/A 204 RCP A Upper Oil Cooler 0 0.0 0 N/A 203 RCP A Lower Oil Cooler 0 0.0 0 N/A 205 O RCP A Thermal Barrier 0 0.0 0 N/A 208 RCP B Upper Oil Cooler 0 0.0 0 N/A 207 RCP B Lower OH Cooler 0 0.0 0 N/A 209 RCP B Thermal Barrier 0 0.0 0 N/A 214 RCP C Upper Oil Cooler 0 0.0 0 N/A 212 RCP C Lower Oil Cocier 0 0.0 0 N/A O 215 RCP C Thermal Barrier 0 0.0 0 N/A 304 SFP Hx C N/A N/A 0 N/A 305 SFP Hx D 150.6 141.6 125 13 %

Node 0401 GFFD 0 - O lsotated Node 0028 Sample Coo 6ers 0 - 0 Isolated O 5 and 28 Totai CCwS Fio. 8038 7555.7 6210 22%

Operating CCW Train A (Spht) A (Split) a) Operators manually bring SFP hxs online by opening Notes upstream isolation valves O

O

O y CAROLINA POWER & LIGHTCOMPANy Calculation ID: SF-0040 Checked by: Date:

• 25 d32 "" O Psoject No.: CALCULATION SHEET p;;,,

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation Tide: Spent Fuel Pools C and D Activation Project Thermal-Hyd.aulic Analysis 1

O Tabie7)

Sununary of CCWS Flow Margins Startup Ops Secton Co.% .t Calculated Flow With 6% Mui Flow (gpm) Flow Margm (%)

(gpm) Uncertainty 105 RHR Pump B 6.9 6.5 5 30 %

108 RHRHxB 35.9 33.9 0 MA 115 RHR Pump A 6.9 6.5 5 30 %

112 RHRHxA 36.7 34.6 0 MA 66 BRS: Dist Car 0 0.0 0 MA

-O 80 BRS: Evap Cir 0 0.0 0 MA 73 BRS: Vent Cond 0 0.0 0 MA 61 Letdown Hx 1250 1179.2 1100 7%

45 XSLD Hx 317.4 299.4- 247 21 %

44 RCOT Hx 301.6 284.5 225 26 %

O 54 Seal Water Hx 306.8 289.4 230 26 %

98 SFP Hx A 3597.4 3393.8 1200 183 %

91 SFP Hx B N/A MA 0 NA 204 RCP A Upper Oil Cooler 193.4 182.5 150 22 %

203 RCP A Lower Od Cooler 6.7 6.3 5 26 %

O 205 RCP A Thermal Barrier 51.4 48.5 40 21 %

208 RCP B Upf-i ON Cooler 193.4 182.5 150 22 %

207 RCP B Lower Oil Cooler 6.7 6.3 5 26 %

209 RCP B Thermal Barr:er 51.4 48.5 40 21%

214 RCP C Uw ON Cooler 193.4 182.5 150 22 %

212 RCP C Lower Os Cooler 6.7 6.3 5 O 26%

215 RCP C Thermal Barrter 51.4 48.5 40 21 %

304 SFP Hx C MA 0 NA 305 SFP Hx D 1602 151.1 59 156 %

Node 0401 GFFD 14 -

14 Spoofied Node 0028 S W cgs 160 160 SpedAed O 5 and 28 TM CCWS Row 6958 6564.2 3830 71 %

Operating CCW Train A A Notes RHR Hx Oucet isolation Valves are Shut O

O

O Computed by: Date:

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 )

Checked by: Date:

4 26d 32 "" O Project No.: CALCULATION SHEET Fk Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Hermal-Hydraulic Analysis O

4.7 Estimate of CCWS Thermal Performance

%e design basis thermal performance of the CCWS was developed by setting the CCWS system thermal boun conditions to the values defined in Table 3. A steady state thermal-hydraulic balance of the CCWS was performed PROTO-FLO 3.04. l For case alignments in which RHR system flow can vary (notably Cooldown and Refueling i alignments), RHR heat exchanger tube side flow is adjusted (up to a maximum of 4500 gpm) to maintain CCW O temperatures at approximately 120*F, consistent with Reference (13). The ESWS flow conditions are assumed to be at the maxunum design temperature of 95'F and the nummum design flow of 8500 gpm, Reference (20), and the CCWS supply temperature is either 105'F or 120*F, depending on the system alignment, Reference (13). Long term stea state spent fuel pool equilibrium temperatures'are estimated from the PROTO-FLO and PROTO-HX results. 'Ibe temperature and heat duty constraints for the CCWS are all satisfied with the current design basis assumptions with the exception of the Startup case alignment in which the CCW supply temperature of 105.I'F slightly exceeds the design O vaue or 105.0*F. The slight increase in CCW supply temperature is considered to be acceptable as the following conditions would not occur simultaneously:

- He CCW heat exchanger model assumes design fouling when trended fouling indicates at least 50 percent margin in the fouling factor.

O -

The CCW heat exchanger Service Water supply conditions of 8500 gpm and 95'F represent the worst case conditions associated with the limiting single active failure of the IMCC-1B35-SB feeder breaker with the ESW system operating on the Main Reservoir at the mmunum design basis level of 205.7 feet.

Maximum letdown flow is assumed on the CVCS side of the Letdown heat exchanger simultaneously with operation of the Excess Letdown heat exchanger at it's design CVCS side conditions.

O Attachments (R) through (Z) and Attachment (EE) document and Tables 8a and 8b summanze the results of this analysis.

4.8 Estimate of Transient Spent Fuel Pool Thermal Performance O

An estimate of the short term transient thermal perfonnance, Attachment (BB), of the spent fuel pools was performed to determine the maximum bulk fuel pool temperature during plant cooldown operations. He transient analysis calculates the bulk fuel pool temperature in 15 minute increments using an estimated fuel pool decay curve correlation, estimated fuel pool heat exchanger thermal performance correlation developed from several PROTO HX" runs, only accounting for the water volume of the fuel pool and neglecting changes in the water thermal properties.

O Fuel pool heatup thermal transients are calculated front p Cp V(=Qpu,jpoof-QSFPHx Equation (l)

• h*'*

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g y CAROLINA POWER de LIGHTCOMPANY Calculation ID: SF-0040 Checked by: Date:

y 427 d32 "" O I l

)

l-Project No.: CALCULATION SHEET Project

Title:

Spent Fuel Pools C and D Activation Project p;j,;

I i

Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis I

j p= Pool Water Density (Ibm / cuft) at temperature Tg l Cp= Pool Water Specific Heat (BTU /lbm / F) at temperature Tg i V= Pool Water Volume (cu.ft)

Tj = Pool Water Bulk Temperature (F) at time tf Tj+; = Pool Water Bulk Temperature (F) at time tj,g QFuelPool=&wyHt(t;)

QSFPHx =f(T;)

Discret1zmg the pool heat up rate term:

dTg+g-Tg

-= T Equation (2) dt tg45 -tg Solving for T at the i+1 time step results in:

0 t -t Tj , g = Tj + Ogy,fp,,7-Qgggg,) Equation (3)

Equation (3) is solved at each time step using the updated decay heat and Spent Fuel Pool heat exchanger correlations described below.

The decay heat correlation for Fuel Pools A/B and C are conservatively estimated from Attachments 5 and 8 of Reference (26) as follows, he Fuel Pool A/B decay heat correlation is calculated by subtracting the values in Attachment 8 for Fuel Pool C from the values in Attachment 5 for Fuel Pools A/B and C. His data is then curve fit, as shown in Figures I and 2 of Attachment (BB), to a generalized decay curve using TableCurve*.

h he Fuel Pool decay heat curves of Reference (26) must be adjusted to represent the decay heat generated from the prevous refueling (RFO9) which would be representative of the fuel pool inventory during the plant cooldown prior to refueling outage 10.~ This calculation assumes that the basic decay heat correlation is conservatively w r esentative of the fuel pool inventory after RFO9 as the decay beat curves from Reference (26) are for the last RPV defueling prior to the Power Uprate outage oflate 2001 (RFO10). %e decay time between RFO9 and RFO10 is calculated to be 519 days (4/15/2000 to 9/22/2001) from Attachment 3 of Reference (26). De adjusted curves are used as input into an Excel b spreadsheet for calculating the transient thermal performance of the spent fuel pools dunng the plant cooldown prior to RFO10.

l The Spent Fuel Pool heat exchanger performance correlation is developed by using the Spent Fuel Pool heat exchanger PROID-HX* model developed in Reference (1) at the minimum CCW flows and maximum CCW supply temperatures identified in Attachment (E). The Fuel Pool Cooling System inlet temperature to the SFP heat exchanger is varied to

). calculate a corresponding heat removal rate for the SFP heat exchanger. These runs, attached, are then curve fit using I

t

r 1

l 7 CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

428 d 32 8" O Project No.: CALCULAT10N SHEET p;3,;

lO Pmject

Title:

Spent Fuel Pools C and D Activation Project l

Calculation

Title:

Spent Fuel Pools C and D Activation Project %ermal-Hydraulic Analysis w

" TableCurve" to develop a correlation of heat removal capacity versus fuel pool outlet (SFP Hx inlet) % ch.

%ese correlations are input into the fuel pool thermal transient spreadsheet.

It is conservatively assumed that the fuel pools are at the maximum temperature limit of 105'F, Reference (33),

the thermal transient. It is also assumed that CCWS supply temperature is a step change to 120*F at the WW o cooldown for an RCS temperature of 350*F. He CCWS supply temperature is maintamed at 120*F throughout the O c d wn transient. Di8 analySI8118 a88um*8 n perator action with respect to the fuel pools durmg the plant cooldown.

He thermal transient for Spent Fuel Pools A/B, su.mu Rud.in Table 1 of Attachment (BB), shows that 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, Reference (13), after the start of the plant cooldown, the fuel pool A/B temperature is 135.7'F which is less than the administrative temperature limit of 137'F, Table 2 of Attachment (BB) shows that fuel pool C will not exceed 113.8'F O which is less than the administrative limit of 137'F and less than the 126*F, assumed for design basis HVAC conditions in Reference (34). nerefore, it is concluded that acceptable spent fuel pool temperatures will be maintained even during a plant cooldown from 350*F to 200*F when elevated CCWS supply temperatures are likely to occur, altho i the fuel pool A/B and C temperatures are bounded by the refueling cases in which the maxunum steady state bulk pI ternperature of 136.3'F and 122.0'F for fuel pools A/B and C, respectively.

l 0- The Fuel Handling Building (FHB) design basis HVAC analysis, Reference (34), shows that four installed air handler cooling coils are sufficient to maintain ambient conditions or 80*F dry bulb temperature and 70 percent Relative Humidity, ne as-built FHB HVAC system only includes three air handler cooling coils, which is justified in Attachment G of Reference (34). A thermal transient analysis of Spent Fuel Pool C was performed to establish the bulk pool temperature at the completion of fuel handling (39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />), Reference (1I), in order to reduce the conservatism in Reference (34). His analysis assumes a step change in CCWS supply temperature to 105'F at the rnmimum CCWS O flow rate defined in Tables 7d through 7f and that Spent Fuel Pool C is at the maximum allowable normal operating temperature of 105'F, Reference (33). Dese thermal conditions are assumed to be maintained throughout the transient even though the CCWS supply temperature will decrease after 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> as the decay heat generated by the recently discharged fuel assemblies in Spent Fuel Pool A/B is decreasing due to longer decay times. %e transient fuel pool C temperature is estimated to be !13.8'F at 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> after commencing fuel handling in the A/B fuel pools which are also assumed to be at the admmistrative temperature limit of 137'F.

O D

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n- -ated by:

Dite:

yeg-Qy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date: .-

4 29 d32 "" O Project No.: CALCULATION SHEET p;g,;

.O  !

Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project Themial-Hydraulic Analysis O Table sa i

Summary of CCWS Steady-State Thermal Capacity Alionment Startup Normal Hot SO (350F) Safe SO (350F)

Lead Uruts Mode 1 @ 4 hrs @ 4 hrs Reference RHR Pump B Heat Load 0 0 0 @Tu*)

70000 0 Calc NSSS 48 R2 RHR Hx 8 Heat Load 0 0 71926000 0 CakxAaled (BTWhr)

RHR Pump A Heat Load 0 0 70000 70000 Calc NSSS-38 R2 (B1Uhr)

RHR Hx A Heat Load 0 0 67817000 81098000 Calculated (BTWhr)

BRS; Dest Car Heat Load 0 0 0 0 Assumed BRS Skad Abandoned O Sw e BRS: Evap Car Heat Load 0 0 0 0 Assumed BRS Sud Abandoned

@TUhr) inplace BRS: Vent Cond Heat Load 0 0 0 0 Assumed BRS SMd Abandoned (BTU /hr)  ;. A -

L+u- -i, Hx Heat Load 15827000 15827000 12536000 0 Ca6culated

@TU/hr)

XSLD Hx Heat Load 5290000 0 0 0 Calculated O *"

RCDTHx Heat Load 2386000 2428000 1871000 1890000 Calculated (BTWhr)

Seat Water Hx Heat Load 1626000 1689000 881000 898000 Cahd ated

@TWhr)

SFP Hx A Heat Load 15345000 15343000 13663000 13680000 Calculated (BTWhr)

SFP Hx 8 Heat Load 0 0 0 0 Secured (BTUh)

O RCP A Heat Load 367000 367000 367000 0 1/3 of WEC COL-536165/79 Value (BTWhr)

RCP8 Heat Load 367000 367000 367000 0 1/3 of WEC COL-53616/479 Value (BTU /hr)

RCPC Heat Load 367000 367000 3G7000 0 1/3 of WEC CQL-536164/79 Value

@TWhr)

SFP Hz C Heat Load 0 0 0 0 Secured (BTWhr)

O SFe Hx o HeatLoad ia** 10 *

• 0 1 *** ia** F-(BTU /hr)

GFFD Heat Load 0 0 0 0 (Bluhr)

Sample Cooiers Heat Load 0 0 0 0 (BTWhr)

CCW Trains No Operstmg 2 (Split) 1 1 1 CorW wfDBD 131 CCW Hx Ht Duty BTU /hr 42.913.000 36.852.000 171,612,748 99.528,000 Calculated O CCw Supper Temp (F) 40s.i 103 8 1i9.6/tio.5 t i9.4 cm- a-d a Hode00ii Doogn CCW Supply (F) 105 105 120 120 Consa. tent w/ 000-131 Temp ESW Flow (C ,.) (gpm) 8500 8500 8500 8500 Desagn Basis ESW inist (F) 95 95 95 95 Temp Fuel Pool AS Temp (F) 122.3 121.0 136.0 135.6 Fuel Pool AS Temp Lirre (F) 137.0 137.0 137.0 137.0 Fuel Pool C/D Temp (F) 117.2 115.8 134.0 133.3 Fuel Pool C/D Temp Limit (F) 137.0 137.0 137.0 137.0 0

1 l

1 Computed by: Date:

JeffLundy CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

'830d 32 "" O Project No.: CALCULATION SHEET File:

O Project

Title:

Spent Fuel Pools C and D Activation Project 1

Calculation

Title:

Spent Fuel Pools C and D Activation Project Thermal-Hydraulic Analysis l

l 3 Table 8b Summary of CCWS Steady-State Thermal Capacity '

Refuel Refuel Refuel LOCA LOCA LOCA Load Units Core Shuffle Futi Omoad AL.w. ; S1 Recire (A) Recuc(B) N;.,-

PHR Pump B Heat Load 0 0 70000 0 0 r

0 Calc NSSS-38 R2 (BTtAtr)

RHR Hx B Heat Load 0 0 0 0 0 0 Calculated

@TUh)

RHR Pump A Heat Load 70000 70000 70000 0 70000 70000 Calc NSSS-38 R2 (BTUh)

RHR Hz A Heat Load 0 0 0 0 118077000 61336000 Calculated (BTU /hr)

BRS: Dist Car Heat Load 0 0 0 0 0 0 Assumed BRS Stud (BTUh) h-aek@

) BRS: Evap Car Heat Load 0 0 0 0 0 0 Assumed BRS Slud (BTUh) h-eM ;,vii..

BRS: Vent Cond Heat Load 0 0 0 0 0 0 Assumed BRS Skid (BTU /hr)

Leidum, Hx Heat Load 0 0 h-AMk@

0 15827000 0 0 Calculated (BTUh)

XSLD Hx Heat Load 0 0 0 0 0 0 Calculated (BTU /hr)

] RCOT Hx Heat Load (BTU /hr) 2394000 2249000 2248000 2437000 0 0 Calmiated Seal Water Hx Heat Load 1673000 1498000 1437000 1701000 0 0 Calculated (BTtMir)

SFP Hx A Heat Load 25271000 32121000 32122000 15341000 0 15399000 Calculated (BTU /hr)

SFP Hx B Heat Load 0 0 0 0 0 0 Secured (BW/hr)

RCP A

] Heat Load (BTU /hr) 0 0 0 367000 0 0 1/3 of WEC COL-53616/5f79 Value RCPB Heat Load 0 0 0 367000 0 0 1/3 of WEC COL 43616/5/79 (BTU /hr) Value RCPC Heat Load 0 0 0 367000 0 0 1/3 of WEC CQL-53616/5/79 (BTU /hr) Value SFP Hx C Heat Load 0 0 0 0 0 0 Secured (BTthtr)

SFP Hx D HealLoad 1000000 1000000 1000000 1000000 0 1000000 Fixed (BTUh)

GFFO Heat toad 0 0 0 0 0 0 (BTU /hr)

Sample C-. Heat Load 0 0 0 0 0 0 (BTU /hr)

CCW Trains No Operatng 1 1 2 (Spht) 1 Spht (1/1) Spkt(1/1) Consistent w/DBO-131 CCW Hx Ht Outy BTU /hr 31.258.000 38.239.000 38.388.629 36,092.000 188,153,000 97,728.000 Calculated q CCW Supp6y Temp (F) 102 8 104.8 104.8/95.0 1036 120.0 1184 Calculated @ W11 s Dessgn CCW Supply (F) 105 105 105 105 120 120 Consrstent w/ DBD 131 Temp ESW Flow (Dessgn) (gpm) 8500 8500 8500 8500 8500 8500 Desgn Basis ESW inlet (F) 95 95 95 95 95 95 l

Temp FuelPool A/B Temo (F) 132.9 136.4 136.4 120 8 isolated 135.9 Fuel Pool A/B Temp (F) 137.0 137.0 137.0 137.0 137.0 137.0

) Fuel Poot C/D Temp (F) 116 9 1215 122.6 115.5 isolated 131.8 Fuel Pool C/D Temp (F) 137.0 137.0 137.0 137.0 137.0 137.0 Umit

V, ,

l 1

Commted by: DIte:

JeffLundy~ CAROLINA POWER & LIGHT COMPANY Calculation ID: SF-0040 Checked by: Date:

r 31 d 32 "" O Proj.ect No.: CALCULATION SHEET File: -

C l Project

Title:

Spent Fuel Pools C and D Activation Project l

t Calculation

Title:

Spent Fuel Pools C and D Activation Project %ermal-Hydraulic Analysis I

4.9 ESWS Hydraulic Margins l Assumption 4.1.5 is critical to this analysis. Table 14 of Reference (20) shows that the nummum available ESWS flow to the CCW heat exchangers is 8797 gpm, including 4 percent ESWS model uncertainty and a single active failure, when operating on the Main Reservoir at the muumum design basis reservoir level. As the worst case dculated single failure flow exceeds the assumed muumum ESWS flow to the CCW heat exchangers, the assumption of a mmimum CCW heat exchanger flow of 8500 gpm is considered to be valid and achievable.

1 4.10 ESWS Ultimate Heat Sink Margins An evaluation of the available thermal and reservoir level margins was performed, Attachment (AA). He l O current UHS analysis of record, Reference (15), evaluated the time dependent effect of a design basis LOCA, i given worst case historical meteorological conditions of 9+1 days. Reference (15) documents a means of l

evaluating the overall energy balance of the HNP main and auxiliary reservoirs. He results from Reference l (15) are that the worst case UHS temperature is 94.2*F which occurs approximately 30 days after a design basis LOCA. The design temperature of the UHS is currently specified as 95'F, Refe ence (19).

O ne thermal margin of the UHS is defined as the difference between the heat rejected from the reservoir at the design temperature and the heat rejection at the maximum estimated water ternperature. Using the UHS heat loss relationship developed in Reference (15) and neglecting the thermal capacitance of the auxiliary reservoir,  !

it was determined, Attachment (AA), that the change in surface heat flux was 6.3 BTU /hr.'F-sq.ft I

( 3.9 BTU /hr-sqft at 95'F and -10.2 BTU /hr-sqft at 94.2*F) due to a change in the reservoir surface temperature from 94.2*F to 95.0*F. The change in heat flux accounts for changes in the convective and evaporative heat D fluxes which are a direct function of the reservoir surface temperature. He change in the surface heat flux I results in a change in the heat rejection capability of 85.17 MBTU/hr, given a reservoir surface area of 1.3519E7 square feet at 249.6 feet, Reference (15).

The activation of Spent Fuel Pools C and D results in an meresse in CCWS and ESWS heat load of approximately 1.0 MBTU/hr, Reference (26). He available thermal mugin of the Ultimate Heat Sink is 85.17 I D MBTU/hr. He change in Ultimate Heat Sink peak temperature is less than 0.01*F, Attachment (AA). It is l concluded that the activation of Spent Fuel Pools C and D are within the current thermal capacity of the Ultimate Heat Sink and have a negligible impact on the design Ultimate Heat Sink temperature.

Reference (15) also evaluated the impact of a design basis LOCA on reservoir levels 30 days after the event l which resulted in the Technical Specification minimum UHS level requirements. He reservoir temperature

] used in the Reference (15) analysis was 95'F for conservatism in order to maxmuze the surface evaporation rate. Based on these considerations, the current UHS level requirements are not impacted so long as UHS thermal margin is available.

)

J L.

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i 10 l .

Computed by: D:ts:

JeffLundy CAROLINA POWER & LIGHTCOMPANY Cciculation ID: SF-0040 Checked by: Date:

'832 d 32 "" O Project No.: CALCULATION SHEET p Project

Title:

Spent Fuel Pools C and D Activation Project Calculation

Title:

Spent Fuel Pools C and D Activation Project 'Ibermal-Hydraulic Analysis l

l0

5.0 CONCLUSION

S l

This analysis docurnents the estimated thc.inal and hydraulic margins in the CCW systein, the ESW system and the UHS. It is concluded that sufficient thermal and hydraulic margins exist in the CCW and ESW systems to support the proposed CCWS tie-in for the Fuel Pool C/D heat exchangers up to a maxunum fuel pool C heat load of 1.0 MBW/hr. It is further concluded that the available thermal margin in the Ultimate Heat Sink is l O sufficient to accommodate the added Fuel Pool C/D heat load of 1.0 MBW/hr which will have a negligible impact on the design Ultimate Heat Sink ternperature or level.

O I

1 l

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CAROLINA POWER & LIGHT COMPANY lO SHEARON HARRIS NUCLEAR POWER PLANT iO PLANT OPERATING MANUAL l

O VOLUME-3-- - - - - - - - - - - - - . - - - - --- - ..-- --

I l

s O PART 4 l

l PROCEDURE TYPE: EMERGENCY OPERATING PROCEDURE (EOP)

END PATH PROCEDURE (EPP) l

'O NUMBER: E0P-EPP-010 t

Q TITLE: TRANSFER TO COLD LEG RECIRCULATION REVISION 11 O

CONTINUOUS USE Continuous Use of Procedure Required.

Read Each Step Before Performing.

O I

l RECEIVED C) M22 91997 e CONTROLLED CM M WP

j. DOCtMENTCONTROL E0P EPP-010 Rev. 11 Page 1 of 22 g

n D

TRANSFER TO COLD LEG RECIRCULATION Instructions Response Not Obtained D

EQIE: The following steps will separate the CCW system into two independent headers and isolate the non-essential header.

(Isolation of the non essential header ensures the design.CCW g flow rate through the RHR heat exchangers.)

8. Verify CCW Alignment To The RHR Heat'Exchangers:

a. Verify both CCW pumps - a. GO TO Step 9.

RUNNING

b. Verify the following valves w-OPEN O ICC-147 1C0-167 l

c. Verify CCW flow to the RHR c. }Dilm CCW flow to the RHR heat exchanger (s). HX(s) verified. TElf do l Step 8d.

Observe !!QIE prior to Step 11 AND continue with ij Step 11.

d. Shut train A CCW O non-essential supply AND return valves:

10C-99 1CC-128 O e. Shut train B CCW non essential supply AND return valves:

1C0 113 1CC 127 D

f. Observe EQIE prior to Step 11 AND GO TO Step 11.

5

] E0P EPP 010 Rev. 11 Page 12 of 22

p 1

. i l: . TRANSFER TO COLD LEG RECIRCULATION l

j FOLDOUT o LOSS OF EMERGENCY COOLANT RECIRCULATION lE emergency coolant recirculation can HQI be established .QB is subsequently lost. IHEN GO TO EPP-012, " LOSS OF EMERGENCY COOLANT RECIRCULATION." Step 1.

o LOSS OF RECIRCULATION SUMP IE recirculation sump level (LI-7160A SA (LI-7160B SB)) drops rapidly to less than 43%. IHEN stop all pumps taking suction on the affected sump. i o LOSS OF SUCTION IE RWST level d'ecreases to 3% (Emp'ty a'larm). THEN secuie all 'pumps' taking suction only from the RWST. -

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TRANSFER TO COLD LEG RECIRCULATION Instructions Response Not Obtained

.9. Verify CCW Alignment To RHR Heat Exchanger A:

a. Verify train A'CCW pump -. a. GO TO Step 10. I RUNNING l

) b. Verify CCW valve from RHR. ,

HX B - SHUT l f

1C0-167 I

c. Verify CCW valve from RHR

) HX A - OPEN:

ICC-147

d. Verify CCW flow to.RHR d. W E H CCW flow to RHR HX A HX A. verified. THEN do Step 9e.

)

Observe HQIE Prior to Step 11 and continue with l Step 11.  :

i

e. Shut train A CCW

) non-essential supply AND l return valves:

4 1CC 99 1CC-128

} f. Observe HQIE prior to Step 11 AND GO TO Step 11.

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'_ E0P-EPP-010 Rev. 11 Page 14 of 22

TRANSFER TO COLD LEG RECIRCULATION FOLDOUT

2) o LOSS OF EMERGENCY COOLANT RECIRCULATION IE emergency coolant recirculation can HQI be established QE is subsequently lost. IHEN GO TO EPP-012. " LOSS OF EMERGENCY COOLANT j

. RECIRCULATION." Step 1. l l

[) o LOSS OF RECIRCULATION SUMP IE recirculation sump level (LI-7160A SA (LI-7160B SB)) drops rapidly to less than 43%. IHEN stop all pumps taking suction on the affected sump.

[) o LOSS OF SUCTION IERWST15 vel ecreases to 3% (Empty alarm). IHEN secure all pumps taking suction only from the RWST. w-

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1 J-E0P EPP-010 Rev. 11 Page 15 of 22

l l TRANSFER TO COLD LEG RECIRCULATION l

I

Instructions Response Not Obtained

10. Verify CCW Flow To RER Heat Exchanger B

l a. Verify train B CCW pump - a. Continue attempts to RUNNING restore CCW.

Observe NOTE prior to j Step 11 AND continue with Step 11.

l

b. Verify CCW valve from RHR l HX'A-.- SHUT 100-147.

c. Verify CCW valve from RHR w HX B - OPEN:

) 1C0-167' i

d. ' Verify CCW flow to RHR d. WHEN CCW flow to RHR HX B HX B. verified. THEN do Step 10e. I Observe t{QIE prior to

) Step 11 AND continue with Step 11.

!q e. Shut train B CCW )

non-essential supply AND

. return valves d l 1CC-113 l 1CC-127

)

i J-r E0P-EPP-010 Rev. 11 Page 16 of 22

a

[ TRANSFER TO COLD LEG RECIRCULATION l

L' l FOLDOUT

l. l

!O o AFW SUPPLY SWITCH 0VER' CRITERIA L E CST level decreases to less than 10%. IHEN switch the AFW water.

i supply to the ESW system using OP-137. " AUXILIARY FEEDWATER SYSTEM".  ;

i Section 8.1. I

Q . o LOSS OF EMERGENCY COOLANT RECIRCULATION H emergency coolant recirculation can HQI be established QH is subsequently lost. THEN GO TO EPP-012. " LOSS OF EMERGENCY COOLANT ,

RECIRCULATION." Step 1. 'I

'O LOSS OF RECIRCULATION SUMP H recirculation sump level (LI-7160A SA (LI 7160B SB)) drops rapidly to less than 43%. IHEN stop all pumps taking suction on the affected ~

sump.

j O L SS OF SUCTION l

H RWST level decreases to 3% (Empty alarm). IHEH secure all pumps taking suction only from the RWST.

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E0P-EPP-010 Rev. 11 Page 17 of 22 0;

O. l TRANSFER TO COLD LEG RECIRCULATION Instructions. Response Not Obtained I

!O-i HQIE: Additional foldout item. "AFW SUPPLY SWITCHOVER CRITERIA".

applies.

l

.O 11'. Implement Function Restoration Procedures As Required..

12 . -- Align CNMT Spray For

,O i Recirculation:

a. Any CNMT spray pump - a. GO TO Step 13.

RUNNING w

b. Verify CNMT sump to CNMT-O spray suction. valves - OPEN l

ICT-105 1CT-102

c. Verify RWST to.CNMT spray O pump suction valves - Suur ICT-26 j ICT-71 i iO O 1

o.-

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0. E0P EPP 010 Rev. 11 Page 18 of 22

1 l

TRANSFER TO COLD LEG RECIRCULATION FOLDOUT C- o AFW SUPPLY SWITCHOVER CRITERIA l

IE CST level decreases to less than 10%, IIIgif switch the AFW water i l -

1- supply to the ESW system using OP-137. " AUXILIARY FEEDWATER SYSTEM".

Section 8.1.

O o LOSS OF EMERGENCY COOLANT RECIRCULATION l IE emergency coolant recirculation can HQI be established _QE is I

subsequently lost. THEN GO TO EPP-012. " LOSS OF EMERGENCY COOLANT

! RECIRCULATION," Step 1.

O o LOSS Of_EEGIRCULATION SUMP {

IE recirculation' sump level (LI-7160A SA (LI-7160B SB)) drops rapidly to

'less than 43%. IIIEH stop all pumps taking suction on the affected w sump, a g o LOSS OF SU M l

IE RWST level decreases to 3% (Empty alarm), THEN secure all pumps l taking suction only from the RWST. ,

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E0P-EPP 010 Rev. 11 Page 19 of 22

D

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TRANSFER TO COLD LEG RECIRCULATION l Instructions Response Not Obtained 0 13. Consult Plant Operations Staff To Perform The Following:

o Restore CCW to the spent fuel pool cooling system using OP-145 " COMPONENT-O COOLING WATER".

Section 8.11.

o Sample AND evaluate sump pH.

O o Evaluate locally placing both CSIP auxiliary oil -

pump control switches in AUTO. ..

o Evaluate defeating the 9 automatic start of the CSIPs from the emergency safeguards sequencers.

(Defeating the CSIP automatic start can be S accomplished by de-energizing control power to the sequencers or g lifting the associated leads to the respective CSIP control circuits.)

9

14. Check Time Since LOCA Initiation GO TO EPP-011. " TRANSFER BETWEEN

- LESS TRAN 6.5 HOURS COLD LEG AND HOT LEG

, RECIRCULATION". Step 1.

] HHEH EPP-011. " TRANSFER BETWEEN COLD LEG AND HOT LEG RECIRCULATION" completed. IHEN RETURN TO procedure and step in effect.

D

15. Within 6.5 HOURS Of LOCA Initiation. Complete SI Alignment For Hot Leg Recirculation Using EPP-011.

• TRANSFER BETWEEN COLD LEG AND HOT LEG RECIRCULATION"

)

E0P EPP-010 Rev. 11

) Page 20 of 22

D CAROLINA POWER & LIGHT COMPANY D

SHEARON HARRIS NUCLEAR POWER PLANT PLANT OPERATING MANUAL D

VOLUME 3 l

PART 2 PROCEDURE TYPE: Operating Procedure

) NUMBER: OP-145 TITLE: Component Cooling Water D

D D

2

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OP-145 Rev. 23 Page 1 of 95

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1 8.11 Restorina ccW to Snent Puel Pool Heat Exchancer after Initiation of cold Leo Recirculation i () 8.11.1 Initial condition =

NOTE . The Spent Fuel' pool temperature is predicted to increase to 137 'F in

t. approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 34 minutes after.CCW is isolated to the Spent-( Fuel Pool Heat- Exchanger. assuming an initial temperature of 112 'F and l

maximum normal heat loads subsequent to a LOCA. Actual conditions may l ~1ead to' slower heat-up rates.

l () '

i

1. Time since'CCW was isolated'to the Spent Fuel Pool Heat Exchanger has been greater than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 34 minutes.

2. Spent Fuel Pool temperature is approaching 137 'F.

i

3. Communications have been established between the operators doing

~

j} the valve manipulations and the Hain Control Room.

8.11.2 Procedural Stens

.HQIE .The~following steps will restore CCW to the-Spent Fuel Pool Heat Exchanger while maintaining at least 5600 gpm to the RER Heat Exchanger.

These steps must be completed prior to the temperature of the Spent Fuel

() .. Pool water reaching 150 'F.

1. Verify the following valves are shut:

a. 1CC-207, CCW TO RCPS.

b. 1CC-208, CCW TO RCPS.

..O

c. 1CC-176, CCW TO EXCESS LTDN & RCDT HEAT EXCHANGERS.

d. 1CC-202,.CCW FM EXCESS LTDN & RCDT HEAT EXCHANGERS.

I e. 1CC-304, CCW TO GROSS' FAILED FUEL DETECTOR.

s l () f. 1CC-305, CCW TO GROSS FAILED FUEL DETECTOR.

g. 1CC-114, CCW TO SAMPLE HEAT EXCHANGERS.

h. 1CC-115, CCW TO SAMPLE HEAT EXCHANGERS.

2. Locally shut the following manual valves.

().

a. 1CC-315, CCW Supply Isolation Valve to Seal Water HX.

b. 1CC-328, CCW Supply Isolation Vlv to Letdown HX.

c. 1CC-343, CCW Supply Isolation Valve to BRS Vent Condenser.

! (). d. 1CC-354, CCW Supply Isolation Vlv to BRS Distillate Cooler.

e. 1CC-357, CCW Supply Isolation Vlv to BRS Recycle Evaporator Condenser.

f. '1CC-383, Fuel Pool HX 1&4A CCW Outlet 2nd Isol V1v.

! () ' g. 1CC-411, Fuel Pool HX 1&4B CCW Outlet 2nd Isolation.

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I OP-145 Rev. 23 Page 30 of 95 j

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8.11.2 Procedural Steos (continued)

3. Open ICC-128 (1CC-127), CCW Nonessential Return to Header A(B).

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4. Open 1CC-99 (1CC-113), CCW Hx A(B) to Nonessential Supply.

5. Verify that the SFP cooling train which is to be placed in service is aligned to the B Spent Fuel Pool per OP-116.

6. Restore CCW to the Spent Fuel Pool Heat Exchanger that is aligned

Q to the B Spent Fuel Pool. The following steps are written for aligning the A Train CCW to the A Train SFP HX, the B Train is shown in parentheses.

a. Unlock and shut 1CC-382 (1CC-398), Fuel Pool HX 1&4A (1&4B)

CCW Outlet Isolation Valve.

O

b. Verify oPen 1CC-374 (1CC-384), CCW Supply Isol Vlv to Fuel Pool HX 1&4A (1&4B).

c. Open 1CC-383 (1CC-411),. Fuel Pool HX 1&4A (1&4B) CCW Outlet 2nd Isol Viv.

d. While monitoring CCW flow to the RHR HX on FI-688A1 l O (FI-689A1), slowly throttle open ICC-382 (1CC-398) to '

establish flow to the SFP HX while maintaining between 5600 to 5800 gpm flow to the RHR HX.

7. Honitor Spent Fuel Pool temperature to maintain less than 150*F.

Throttle open 1CC-382 (1CC-398), Fuel Pool HX 1&dA (1&4B) CCW Outlet Isolation Valve as necessary while ensuring

,C FI-688A1(FI-689A1) remains above 5600 gpm.

8. Document component manipulations per OMH-001.

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i OP-145 Rev. 23 Page 31 of 95 j

O-SYSTEM # 5165.5175.5185 CALC. TYPE EACSYS i

O I l

CAROLINA POWER & LIGHT COMPANY O

CALCULATION NO. E-6000 FOR 1

O AUXILIARY SYSTEM LOAD STUDY FOR l

SHEARON HARRIS NUCLEAR POWER PLANT O

NUCLEAR ENGINEERING DEPARTMENT QUALITY CLASS: aA OB OC OD DE O

RESPONSIBLE u DESIGN VERIFIED BY APPROVED BY Rev. ENGINEER O ENGINEERING REVIEW BY RESPONSIBLE SUPERVISOR DATE DATE DATE O

0-5 See original Cover Sheets (following pages) 6 W%% RAH &SM Yf' NOh k V *4 0* N" Y//3/71 4 //6/99 4llM O. REASON FOR CHANGE: See Revision Summary REASON FOR CHANGE:

REASON FOR CHANGE:

O O

4 CALCULATION NO. E-6000 '

PAGE li REV.6 REVISION

SUMMARY

g Rey,# Revision Summary 0-5 See original cover sheet.

6 Incorporated Spent Fuel Pool C & D Activation Project (ESR 95-00425). The specific g

changes are as follows: Added a 150 hp motor each to buses 1&4A33-SA and 1&4B33-SB. Added the motor's heater loads to PP1&4A33-SA and PP1&4B33-SB. Revised the feeder cable length for 1&4A33-SA and 1&4B33-SB. Also revised the limiting loading value for these busses. Added a 60 hp motor each to busses 1-4A1021 and 14B1021.

Added the motor's heater load to PP1-4A10221 and PP1-4B10212. Added a 40 hp motor to Bus 1-4B1021. Added new LCP loads to PP1-4A10221, PP1-4A111 and PP1-4B111.

D Also corrected the existing SFP Cooling Pump and Purification Pump motors to show load required during shutdown / refueling.

Also incorporated the outstanding changes listed on page iii. The specific changes associated with the safety related busses are as follows:

Bus I A-SA - revise the bhp for the ESW pump motor. Bus 1 A21-SA - revised bhp O values for Fan S-2 and S-4, also revised LP-118 and LP-134 loading. Revised the l starting of MOVs 3CZB3,3CZB11 and 3CZB9 to start in any load block. Bus 1 A36-SA

- revised steady state minimum voltage requirement. Revised the starting of MOVs 3CZB5,3CZB7,3CZB1,3CZB13,3CZB17,3CZB25 & R2 to start in any load block. 1 Revised EHC-72 to be on in any load block. Bus 1 A35-SA - revised bhp of MOV O 2CSL523-SA-1, revised the starting of PI A-SA to start in any ly.d bock. Bus 1&4A33-SA - revised LP-502 loading, revised the starting of E-12 to start in any load block. Bus ;

1 A32-SA - revised the starting of E-88 to start in any load block. Bus 1 A31-SA -

revised the starting of Fan S-68, Pumps 2MS-P18 and 2MS-P20 to start in any load block. PPIA211 SA reduced load.

g Bus 1B-SB - revised bhp for ESW pump motor. Bus 1821-SB - revised bhp values for Fan S-2 and S-4, also revised LP-119 and LP-534 loading. Revised the starting of MOVs 3CZB10,3CZB12 & 3CZB4 to start in any load block. Bus IB36 SB - revised steady state minimum voltage requirements. Bus 1B35-SB - revised bhp of MOV 2CSL522-SB-1, revised the starting of Boric Acid Transfer Pump 1B to start in any load block.

Bus 1&4B33-SB - revised the starting of E-13 to start in any load block. Bus 1B31-SB -

O revised the starting of 2MS-P19 to start in any load block. Bus 1B32-SB - revised the starting of E-88 to start in any load block. Bus IB36-SB - revised the starting of MOVs 3CZB14,3CZB18,3CZB2,3CZB20,3CZB22,3CZB24,3CZB26,3CZB6,3CZB8 & R2 to start in any load block. Revised EHC-72 to be on in any load block. PPIB211-SB reduced load.

3 The BOP buses load changes associated with the ESRs listed on page iii, were incorporated into the AFLSC, BFLSC, Q1 Load List, Q2 Load List, AOVERSUT &

BOVERSUT databases. The full load runs and LOCA runs were not revised since the new load changes are enveloped by the existing future load margin /POLDF data.

Incorporated changes due to Calculation E-6003, Rev. 3.

D This revision changed the methodology used for performing the LOCA/ LOOP & LOOP i~ (EDG) Load / Voltage Study, LOCA Voltage Study (including DGVR reset analysis), and Degraded-Grid Voltage Relay Dropout Setting Study. The methodology used for

)

I i

O CALCULATION NO. E-6000 PAGE iii

REV. 6 l determining the transformer and cable losses for the diesel loading was also revised. This l revision also revised the POLDF data and ESW Pump hp values utilized for the LOCA

'""8-

!O The following outstanding ESRs were incorporated:

ESR 94-00013, Rev.0, ESR 94-00067, Rev.0*, ESR 94-00171, Draft Rev.0*, ESR 94-00543, Rev.1, ESR 94-00151, Rev.0, ESR 95-00085, Rev.0*, ESR 95-00153, Rev.0*, ESR 9540215, Rev.1*,

ESR 9540311. Rev.0, ESR 95-00410 Rev.0, ESR 9540425 Rev.0*, ESR 95-00427, Rev.8,9,10,

'O l

ESR 95-00431, Rev.0*, ESR 95-00565. Rev.0, ESR 95-00908, Rev.0, ESR 95-00939 Rev.0*,

l ESR 95 00993, Rev.0, ESR 95-00995 Rev.0*,4, ESR 95-00997, Rev.0*, ESR 95-01001 Rev.0*,

ESR 95-01031, Rev.0, ESR 96-00025 Rev.2,3, ESR 96-00116. Rev.0, ESR 96-00163, Rev.3, j ESR 96-00216, Rev.0, ESR 96-00226, Rev.0, ESR 96-00315, Rev.0, ESR 96-00320, Rev.0, l

O ESR 96-00511 Rev.0, ESR 97-00183, Rev.2, ESR 97-00237. Rev.1&2, ESR 97-00248, Rev.0, '

ESR 97-00368. Rev.0.2, ESR 97-00559, Rev.0, ESR 98-00006, Rev.0*,2*, ESR 98-00058, Rev.0, PCR-5767 Rev.0

• Indicates that the ESR has not been field implemented. These ESRs should be reviewed during the next revision of this calculation to verify that they were not O cancelled, revised, etc.

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Computed by: Dates CAROLINA POWER & LIGHT COMPANY Calculation ID:

10M ~#. h Wo/f1 E-6ooo C

C cke g '

Date:

Allll%

Pg. 1 of 6 Rev. 6 l CALCULATION SHEET TAR /PID No.: N/A File: N/A Project

Title:

!O C*1 "l*ti " Titl*: ^"*ili*rY 8Y" tem L aa Study j

Status: Prelim. Final Void 1.0 PURPOSE O

1.1 E-6000 serves in the following capacities:

Performs load list analyses to tabulate maximum expected bus loadings during various plant conditions  !

t O . Performs voltage drop analyses under full load conditions e Performs overvoltage analyses under shutdown conditions I t

Performs voltage drop analyses during load sequencing via offsite power and via standby

g diesel generators

• 1 Analyzes auxiliary system degraded voltage performance to support undervoltage relay {

settings per Branch Technical Position PSB 1 {

e Performs short circuit analyses to determine available fault currents in order to assure that O installed equipment ratings are not exceeded 2.0 LIST OF REFERENCES 2.1 Calculation E-6001 Rev.2, " Load Factor Study"  !

2.2 AUXSYS 4078 User's Manual for Electrical Auxiliary System Design Rev. 2,12/89 2.3 Calculation E-6003 Rev. 3 " Minimum and Maximum Operating Voltages Required for Class IE l l Buses" '

'O 1 2.4 Design Basis Document #202 Rev. 4, " Plant Electrical Systems, Off Site Power Systems, Generator, l t Exciter, Isolated Phase Bus Duct. Generator & Exciter Mechanical Support Systems" 2.5 Calculation 17-EP Rev. 2, " Diesel Generator Load Sequence Voltage Profile" l

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.O Computed by: Date: CAROLINA POWsR & LIGHT COMPANY Calculation ID:

"Ad47t'T h f//J/ff E-6000 Date:

O Checke1by:

ga y n. I fM 4/N Pg. 2 of 6 Rev. 6 i CALCULATION sHssT TAR /PID No.: N/A File: N/A i

Project

Title:

O Calculation

Title:

Auxiliary system Load study ,

status: Prelim. Final Void 3.0 CALCULATION O

This study is an analysis of the Harris Nuclear Plant (HNP) electrical auxiliary system under various loading conditions. Computer databases are developed utilizing load input from electrical calculation E-6001 for use with Ebasco software programs ESMS4128 and AUXSYS4078. Synopses of these databases are located in Tab D of this calculation. Configuration of the HNP electrical distribution l' system is modeled in accordance with sketch SK-H0183-E-3100 shown in Tab W. Computer analyses O are then performed to calculate load lists, bus voltages, and short circuit currents based on the E-6000 databases.

3.1 Load List Analyses O The loading of the auxiliary system under various plant scenarios is tabulated via a load list on a per  ;

train basis. In order to quantify loading under various plant conditions (full load, shutdown, start-up, IAss-of-Coolant Accident (LOCA) and LOCA/ Loss-of-Offsite Power (LOCA/ LOOP), a Train-A and l !

Train-B load list is generated based upon electrical calculation E-6001 and E-6000 Tab L. These load I lists are located in a separate binder and are further described in Tab Q. Results of the load list analyses are summanzed in Tab B. Details concerning total plant load magnitudes given to the O Transmission Planning group are located in Tab R. 4 l

3.2 Undervoltage Analyses under Full-load Conditions Bounding analyses are performed to envelop minimum expected voltages throughout the plant t

d'stributi n system dunng both winter and summer fun-load conditions with plant loads fed via either i O

the Unit Auxiliary Transformers (UATs) or Start-Up Transformers (SUTs). Details conceming the j databases set-up to bound these conditions are located in Tab D. When analyzing the Train-A  ;

distribution system, all redundant loads are assumed operating from applicable Train A buses. Under minimum possible voltage conditions as discussed with Transmission Planning, the effect on bus  ;

voltages with all expected loads running and with various 6.9kV and 480V large motors starting is O investigated. This process is repeated for Train-B with redundant loads assumed operating from Train-B buses. Results of the full-load voltage analyses are located in Tab B.

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Computed by: Date: Calculation ID:

CAROLINA POWER & LIGHT COMPANY E-6000 Checked by: Date: Pg. 3 of 6 Rev.6 O bM h. M ' 4//N CALCULATION SHEET j TAR /PID No.: N/A File: N/A Project

Title:

Calculation

Title:

Auxiliary System Load Study O Prelim. U Status: Final LXI Void U 3.3 Overvoltage Analyses During Shutdown Conditions  ;

O \

Existing transformer tap settings are verified to provide satisfactory bus voltage levels within maximum allowable voltage criteria under minimum load conditions. For the overvoltage analyses, the plant is modeled at minimum load during shutdown conditions widle connected to the 230kV switchyard via the start-up transformers. The switchyard is set at its maximum expected voltage per discussions with Transmission Planning (see Tab M). A summary of the results of the overvoltage analyses is located in Tab B.

O i 3.4 LOCA/ LOOP & LOOP (EDO) Load / Voltage Study Analyses are performed to determine EDG terminal voltage and worst-case EDG loading during both LOCA/ LOOP sequencing and LOOP sequencing. To determine worst-case EDG loading, the LOCA/ LOOP and LOOP scenarios have been comNued into one conservative, enveloping scenario called LOCA/ LOOP-O LOOP. EDG terminal voltage during sequencing is obtained from data collected during the 18 month surveillance test. Tab L, Appendix C provides a copy of the EDG terminal voltage data collected during OST testing. Tab D, S.uion 3.4 describes the computer database used in the EDG loading analysis. Tab L provides details of the load model and methodology. Results are summarized in Tab B, Section 3.4.

3.5 LOCA Voltage Study (Offsite Power Available)

O Analyses are performed to calculate minimum available voltages throughout the emergency power system during and after LOCA load sequencing with the auxiliary electrical distribution system supplied from offsite power. A detailed study of emergency power system loads is conducted in Tab L to determine worst case loading during a LOCA. The loading on associated BOP buses is modeled with " lump" motor loads as detailed in Tab E.

With the 230kV switchyard set at its minimum expected voltage, LOCA voltage analyses are performed to calculate minimum expected emergency power system voltages during motor starting and at the end of each load block. Additionally, a detailed study is provided to demonstrate the ability to " reset" the degraded grid voltage relay (DGVR) during LOCA sequencing. Tab D, Sections 3.6 & 3.7 provide details of the databases 3 used. Results of the LOCA voltage studies are shown in Tab B, Section 3.5.

O O

i LO Computed by: Date: Calculation ID:

CAROLINA POWER & LIGHT COMPANY E-6000 Checked by: Date: Pg. A of 6 Rev.6 lh*54 M.#d A//'/99 CALCULATION SHEET TAR /PID No.: N/A File: N/A ProjectTitle:

O Caiculation Titie: Auxiliary System Load Study i

Status: Prelim. U Final lIl Void U l

.O 3.6 Degraded <rrid Voltage Relay Dropout Setting Study )

. Analyses are performed to calculate minimum available voltages throughout the emergency power system during switchyard degraded voltage conditions. Loads fed from emergency power system buses are modeled i in a steady-state operating condition in accordance with the LOCA sequencing study in Tab L. Operation of l motor operated valves is enveloped by modeling the largest motor operated valve on each MCC as running. l 0 The loading on associated BOP buses is modeled with " lump" motor loads as detailed in Tab E.

To verify adequacy of the degraded grid voltage relay (DGVR) dropout setting, the switchyard voltage is

'incrementally reduced until the voltage at the 6.9kV emergency bus reaches the lowest allowed "as-left" .

DGVR. dropout setting with tolerance. The resultant emergency power system voltages are then compared l with their steady-state criteria voltage (as determined in Calculation E-6003). Tab B, Section 3.6 contains a O sumrnary of the voltages present throughout the emergency power system at the DGVR dropout setting.

3.7 Plant Operating Load Demand Factors i

In the normal full-load, LOCA, and degraded-grid voltage analyses, plant operating load demand factors O (POLDFs) are utilized for various non-safety related buses. By using these factors, more realistic normal full-load voltages are calculated. Also, for LOCA and degraded-grid voltage analyses, the factors enable the use of lump motor loads to simplify databases while still accounting for the impact of non-safety related loads on the Class-lE portion of the distribution system. POLDFs are developed for the non-safety related buses shown below.

O- Train A Buses Train B Buses IDI lEl ID2 lE2 O ID3 IE3 1-4AA 1-4AB Table 1: Non-Safety Related Buses Modeled per POLDF Data Tabs D and E of this calculation contain further details conceming the use of POLDF data for the above buse in voltage analyses.

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9 C uted by: Dates

(#

CAROLINA POWER & LIGHT COMPANY Calculation ID:

/ AM /o//ch E-6ooo j /

g j  : , Date: Pg. 5 of 6 Rev. 5

- b CALCUIATION SHEET TAR /PID N N/A Files N/A Project Titles g Calculation

Title:

Auxiliary System Load study Status: Prelim. O Final [i] Void C 3.8 Short Circuit Analysis O

In order to determine worst-case short circuit currents available throughout the auxiliary system, the following analyses are performed on a per train basis:

1) Annivsis with marimum motor contribution and a high prefault voltage of LO pu.

O Maximum motor contribution is modeled by running all Class-lE motors shown as starting or running in load block #9 of the LOCA sequencing study described in Tab L as well as Non<: lass IE motors specified as running during winter or summer full-load conditions per calculation E-6001. Also, the effects of the main generator, security diesel generator, uld emergency diesel generators are included in the analysis. The Train-A and Train-B fu'l-g load short circuit databases are utilized for generation of these mialyses with maximtun motor contribution and a 1.0 pu high prefault voltage.

2) Analysis with minimum motor contribution and a maximum prefault voltage.

0 Minimum motor contribution is modeled by running all motors specified as normally operating during shutdown conditions as noted in calculation E-6001. Also, in order to envelop motor testing and other intermittent motor operation during shutdown, the largest motor other than those already noted as normally running during shutdown conditions is modeled as running in the study. The Train-A and Train-B shutdown databases are utilized for generation of these short circuit analyses. Maximum prefault voltages are 9 applied to the system as described in section 3.3 of Tab B.

A companson of the results from the above analyses is conducted in Tab B. Worst-case available short circuit currents are identified for comparison with installed equipment capabilities.

Maximum acceptable short circuit currents are based upon installed bus ratings and circuit breaker intermpting capabilities.

4.0 CONCLUSION

S Overall, the electrical auxiliary system meets design requirements for proper operation of 3 equipment via both onsite (emergency diesel generators) and offsite (230kV switchyard) power sources. Acceptable switchyard voltages have been discussed and agreed-upon between HNP and Transmission Planning personnel (see Tab M). Details regarding the results of the various l ,, analyses are located in Tab B.

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TAB L CALCULATION E-6000 i LOCA AND LOCA/ LOOP LOOP SEOUENCING PAGE 19, R6 Table 6: Train-A LOCA/ LOOP and LOOP Loading O

LOAD STARTING RUNNING BLOCK kVA kW kVAR PF kVA kW' kVAR PF 1 6987.1 2075.4 6671.8 0.30 1305.8 1190.5 536.5 0.91 0 2 6115.2 2640.3 5515.8 0.43 1976.3 1788.2 841.4 0.90 3 9419.2 3667.4 8675.9 0.39 3083.2 2761.4 1371.4 0.90 4 9542.2 4390.4 8472.2 0.46 3728.4 3364.2 1607.1 0.90 5 8574.8 4555.6 7264.6 0.53 4366.1 3930.4 1901.2 0.90 0

6 6777.9 4527.2 5044.2 0.67 4703.6 4246.0 2023.8 0.90 7 6086.9 4497.8 4101.2 0.74 4676.0 4221.3 2011.4 0.90 8 9842.2 5524.8 8145.3 0.56 5408.9 4887.8 2316.4 0.90 0 9 7994.5 6317.9 4898.6 0.79 6582.8 6030.6 2639.2 0.92 Table 7: Train-B LOCA/ LOOP and LOOP Loading O

LOAD STARTING RUNNING BLOCK kVA kW kVAR PF kVA kW' kVAR PF 1 7002.3 2043.9 6697.4 0.29 1286.5 1161.7 552.8 0.90 0

2 6051.2 2628.2 5450.7 0.43 1960.1 1762.0 858.6 0.90 3 9410.2 3641.4 8677.1 0.39 3067.3 2735.0 1388.5 0.89 4 9514.8 4366.4 8453.8 0.46 3711.5 3337.3 1624.2 0.90 0 5 8561.7 4537.0 7260.7 0.53 4354.7 3908.3 1920.6 0.90 6 6717.0 4463.0 5020.0 0.66 4653.6 4183.6 2038.0 0.90 7 6009.9 4429.0 4062.4 0.74 4620.0 4153.9 2022.3 0.90 8 9786.3 5460.4 8121.3 0.56 5353.1 4820.8 2327.2 0.90 O

9 7915.9 6245.5 4863.6 0.79 6515.5 5957.0 2639.3 0.91 NOTES FOR TABLES 6 & 7 O 1. Running kW load data is used in FSAR TABLES 8.3.1-2a and 8.3.1-2b. Revisions to this data may require an FSAR revision. See Section 4.2.4. Cable & transformer losses of 40kW must be added to the load block 9 " running load" in order to determine final EDG steady-state loading after sequencing.

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Holtec Center,555 Lincoln Drive West, Marlton, NJ 08053 HOLTEC INTERNATIONAL Telephone (609) 797-0900 O

STUDY / SCOPING REPORT O for the FUEL STORAGE PROJECT IN POOLS C AND D O *'

HARRIS NUCLEAR PLANT O II CP&L Holtec Report No: HI-971703 O

Holtec Project No: 70324 Report Category: A O Report Class: Safety Related 0

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0-Executive Summary

O This report presents a feasibility and scoping study of the racking project for pools C and D of the Harris Nuclear Plant. The project will provide increased storage capacity to accommodate spent fuel from Harris, Robinson, and Brunswick plants. Since the O

fuel discharge rates of the plants contain a number of unquantifiable variables, maximum flexibility in storage configurations is sought. This flexibility require.s a

]

development of various rack layouts to accommodate a range in the number of storage O locations for both PWR and BWR assemblies. This report evaluates the criticality, thermal-hydraulic, and structural aspects of the PWR and BWR storage racks proposed by Holtec Intemational.

O The report has been similarly arranged to the format of the licensing amendment j

. application report expected by the NRC and used successfully for numerous previous re-racking projects. Chapters 1 through 3 provide general descriptions of the project O and rack hardware. Chapter 4 includes criticality evaluations extracted from other plants with similar rack designs and fuel types. Chapter 5 develops an overall pool heat load and discusses other pertinent thermal hydraulic issues and associated analysis g techniques. Chapter 6 discusses rack dynamic analysis methodology and the results j from preliminary rs'/x stability evaluations. Chapter 7 presents the pool structure evaluation methomalogy.  !

O The final licensing amendment application report will, of course, be much more extensive and will cover the results of all analyses required to be performed for licensing of the new storage racks.  ;

O I All information in the final report will be prepared in unison with the ongoing CP&L '

project to complete the spent fuel pool cooling and cleanup system for pools C and D, which was abandoned during construction. The pool heat load and thermal hydraulic O

Holtec Report Hi-971703 E-1 Project H-70324

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methods discussed herein provide a starting point for design input development.

D The conclusion which may be developed from this scoping report and supporting documentation is that the proposed rack designs are suitable for the intended application at Harris Nuclear Plant. The racks will perform in an acceptable manner O under all postulated conditions and provide for maximum flexibility and storage capacity increases on an as needed basis.

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Holtec Report Hi-971703 E-2 Project H-70324 !

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4.7.1 are pertinent to the higher-reactivity unborated case.

O-4.7.2 Dronoed Fuel Assembly i For a drop on top of the rack, the fuel assembly will come to rest horizontally on top of

O the rack with a minimum separation distance from the fuel in the rack of more than 12 l _

inches, including an estimated allowance for deformation under seismic or accident I

conditions. At this separation distance, the effect on reactivity is insignificant (<0.0001

,O 5k). Furthermore, soluble boron in the pool water would substantially reduce the reac-tivity and assure that the true reactivity is always less than the limiting value for any probable dropped fuel accident. Consequently, fuel assembly drop accident will not result in a significant increase in reactivity due to the separation distance.

iO 4.7.3 Lateral Rack Movement O' Lateral motion of the rack modules under seismic conditions could potentially alter the spacing between rack modules. Region 2 storage cells do not use a flux-trap and the reactivity is therefore insensitive to the spacing between modules. The spacing o between modules is sufficiently large to preclude adverse interaction even with the maximum seismically-induced reduction in spacing. Furthermore, soluble poison would  !

assure that a reactivity less than the design limitation is maintained under all accident l or abnormal conditions.

i 4.7.4 Abnormal Location of a PWR Fuel Assembly O The abnormal location of a fresh unirradiated fuel assembly of 5.0 w/o enrichment could, in the absence of soluble poison, result in exceeding the design reactivity limitation (k of 0.95). This could occur if a fresh fuel assembly of the highest permissible enrichment were to be either positioned outside and adjacent to a storage i

Holtec Report Hi-971703 Page 4-21 Project H-70324 j

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rack module or inadvertently loaded into a storage cell. The largest reactivity increase g results if a fresh unburned assembly (5.0 w/o enrichment) were to be accidentally installed in a cell with the surrounding cells filled with fuel of the highest permissible reactivity. Calculations for this condition resulted in a maximum reactivity of 0.990, including uncertainties. Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. It is estimated that a soluble poison concentration of 400 ppm boron would be sufficient to maintain a k. less than 0 0.95 (including uncertainties) under the maximum postulated accident condition.

4.7.5 Abnormal Location of a BWR Fuel Assembiv O

This accident is not a postulated event, since the worst case reactivity is considered for BWR fuel assemblies.

O 4.7.6 Eccentric Fuel Assembly Positionina The fuel assembly is normally located in the center of the storage racf. cell with g bottom fittings and spacers that mechanically restrict lateral movement of the fuel assemblies. Nevertheless, calculations with the fuel assembly moved into the comer of the storage rack cell (four-assembly cluster at closest approach),

resulted in a small negative reactivity effect. Thus, the nominal case, with the fuel 8

assembly positioned in the center of the storage cell, yields the higher reactivity.

4.7.7 BWR Fuel Zirconium Channel Distortion O

Consequences of bulging of the zirconium fuel channel are treated as a mechanical deviation in Section 4.6.2.6. Bowing of the zirconium channel (including fuel rods) results in a local negative reactivity effect analogous to that Holtec Report Hi-971703 Page 4-22 Project H-70324 0

.O CP&L O c.,.lio. Pow.r a ught company Harris Nuclear Plant P.O. Box 165 New Hill NC 27562

!/N. ids SERIAL: HNP-99-069 O APR 3 01999 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington,DC 20555 O

SHEARON HARRIS NUCLEAR POWER PLAST DOCKET NO. 50-400/ LICENSE NO. NPF-63 RESPONSETO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ALTERNATIVE PLAN FOR SPENT FUEL POOL O

COOLING AND CLEANUP SYSTEM PIPING j

Dear Sir or Madam:

By letter dated March 24,1999, the NRC requested additional information regarding the Harris O Nuclear Plant (HNP) license amendment request to place spent fuel pools 'C' and 'D' in service.

Enclosure 8 of the HNP license amendment request (ref. SERIAL: HNP-98-188, dated December )

23,1998) provided a detailed description of the proposed alternatives to demonstrate compliance j with ASME B&PV Code requirements for spent fuel pool cooling and cleanup system piping in i a e rdance with 10 CFR 50.55a(a)(3)(i). The NRC has determined that additionalinformation is )

O required to complete the review of the proposed alternative piping plan. Enclosed is the HNP response to the NRC request for additional information. The enclosed information is provided as a supplement to our December 23,1998 submittal and does not change our initial determination j that the proposed license amendment represents a no significant hazards consideration.

O ' Please refer any questions regarding the enclosed information to Mr. Steven Edwards at (919) 362-2498.

Sincerely,

%. . Wawb Donna B. Alexander l

Manager, Regulatory Affairs Harris Nuclear Plant O

KWS/kws Enclosures O 5413 Sheoron Harris Road New HiH NC l

l O. Document Control Desk -

SERIAL: HNP-99-069 Page 2 O'

c: j Mr. J. B. Brady, NRC Senior Resident Inspector (w/ Enclosure 1)

Mr. Mel Fry, N.C. DEHNR (w/ Enclosure 1)

Mr. R. J. Laufer, NRC Project Manager (w/ all Enclosures)

O Mr. L. A. Reyes, NRC Regional Administrator (w/ Enclosure 1)

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O Document Control Desk Serirl: HNP-99-069 Page 3 O

bc: (w/o enclosures)

Mr. K. B. Altman ' Ms. L. N. Hartz 1 Mr. G. E. Attarian Mr. W J. Hindman O Mr. R. H. Bazemore Mr. C. S. Hinnant Mr. S. R. Carr Mr. W. D. Johnson  ;

Mr. J. R. Caves Mr. G. J. Kline i Mr. H. K. Chernoff (RNP) -- Ms. W. C. Langston (PE& RAS File) l Mr. B. H. Clark Mr. R. D. Martin I O - Mr. W. F. Conway Mr. T. C. Morton Mr. G. W. Davis Mr. J. H. O'Neill, Jr.

Mr. W. J. Dorman (BNP) Mr. J. M. Taylor Mr. R. S. Edwards Nuclear Records Mr. R. J. Field Harris Licensing File O Mr. K. N. Harris Files: H-X-0511 H-X-0642 O

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o O- Enclosure I to Serial: HNP-99-069 Page 1 of 19 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/ LICENSE NO. NPF-63

.O RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ALTERNATIVE PLAN FOR SPENT FUEL POOL COOLING AND CLEANUP SYSTEM PIPING O

I. Existine Pioinn System A. Detailed description of the proposed change:

O Reauested Item I.A.1 Provide isometric drawings (isometrics) showing all piping and piping systems within the scope of the proposed alternatives; i.e., for fuel pool cooling and cleanup system O (FPCCS) and component cooling water system (CCWS) piping. Provide Isometric drawings to be used for continuance of design and construction without an N-Stamp.

Response to Reauested Item I.A.1 O Copies of the original construction isometrics are provided in Enclosure 2 and have been marked up to show:

E installed piping (in scope of the Alternative Plan)

E embedded piping O a class boundaries, including safety vs. non-safety related E location and identification of field welds In addition, please note that these isometrics include the following information:

O a material requirements for piping and fittings 5 pipe spool numbers (traceable to vendor data packages) e location of hanger attachment lug welds These markups were based upon detailed field walk downs of the current system O configuration. Documented verification of these details will b: provided by the system turnover / certification process used to implement this activity (ref. responses to RAI items II.2 & 3). Piping outside of Code boundaries is identified on these isometrics only for the purpose of depicting continuity.

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O Enclosure 1 to Serial: HNP-99-069 Page 2 of 19 Reauested Item I.A.2 Provide weld matrixes that list all the welds (each weld should be uniquely identified and traceable to I.A.1 above) within the scope of the alternatives.

Response to Recuested Item I.A.2 O

A matrix is provided in Enclosure 3 for each of the field welds in the scope of the Code related piping discussed in I.A.I. For clarity,in-scope field welds are defined herein as that set of field welds which meet all of the following criteria:

(1) is installed in the ASME Section III Class 3 boundaries of the Component Cooling Water or Spent Fuel Pool Cooling Systems (2) was installed during original plant construction, (3) Code required field installation records are no longer available (4) is consistent with the design of the system as it will be completed (5) is in the "large-bore" piping on the main system flow path. Instmment lines, vents O and drains, branch connections to other systems, etc., are not included.

Reauested Item I.A.3 O (i) In the matrixes or isometrics, identify the piping material (ASME / ASTM

- Specification), weld material (ASME / ASTM Specification), the existence of all required material documentation, and any specific missing docurrentation. (ii) Identify each missing document for each weld. (iii) Identify the method (s) used for reconciliation of each type of missing document. (e.g., missing Certified Material Test Report O reconstructed with complete chemical analysis run on shavings taken from the material).

(iv) For the sampling and testing methods used for reconciliation, identify references used for guidance. (i.e., NRC DG-1070, ASME, or EPRI). Explain any differences between the sampling / testing methods and the selected referenced guidance. (v) For ,

chemical analysis, identify sample size and chemical analysis (mean and standard I O deviation for each element) for each analyzing technique.

Response to Reauested Item I.A.3 l (i) The weld matrix (Enclosure 3) includes a listing of weld material based on a review O of applicable Weld Procedure Specifications (WPS) and Weld Data Reports (WDR) for comparable piping. Note that piping material requirements are included in the isometrics provided in response to requested item I.A.I. All Code piping in the scope of the Alternative Plan has been supplied by an NFT Stamp holder and vendor documentation for this material is on hand. This accounts for material certification O- for all of the piping within the scope of the Alternative Plan and the large majority of the welds in that piping. The outsmnding material certification issue to be addressed j herein is that associated with welding materials for a relatively small group of field installed welds on the large bore (12" and up) Code piping. During construction, O- i t

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b Enclosure I to Serial: HNP-99-069 Page 3 of 19 filler metal traceability was accomplished by recording the material heat number on the WDR. The WDR was incorporated into the piping installation package, and 3 typically became the only source of this information to be forwarded to document control. Since the WDRs for these field welds are not on hand, the traceability of filler metal cannot be established. ,

(ii) The WDR was used to provide the installation record for field welds. Generally, g these reports are no longer on hand for the subject welds.

1 (iii) The WDR contained information pertaining to weld attributes, including identification of the items being welded, specification of the WPS to be used, welder identification, filler metal material identification, NDE requirements, and signature g documeatation (including that of the ANI) that all required attributes were satisfactorily performed and verified as complete. Reconciliation of missing information is presented in the weld matrix discussed in response to requested item I.B.4.

g (iv) The sample size chosen for verifying filler metal composition of accessible (i.e., non-embedded) field welds is 100% All of the accessible field welds (including welds for hanger lugs) in the large bore stainless steel Spent Fuel Pool Cooling System piping subject to the Alternative Plan have been evaluated for material composition using a Metorex X-Met Alloy Analyzer. Additionally, three of these stainless steel welds g have been subject to laboratory analysis of chip samples to verify chemical composition. All three of the large bore carbon steel field welds in the CCW System subject to the Alternative Plan will be evaluated by laboratory analysis of chip samples since the alloy analyzer does not lend itself to reliable evaluation of this material. The use of these specific methods for determination of base metalis

, provided in the Corporate Welding Manual, Procedure NW-16. Chemical analysis was and will continue to be performed by a reputable and recognized laboratory (NSL Analytical Services, Inc of Cleveland, Ohio for completed analyses) to traceable standards. Since some blending of filler metal and base metal may have occurred with the field welds in question, the results of the filler metal analysis is being evaluated by CP&L's Materials Services Section - Metallurgy Unit (See Enclosure 4 for analysis of SFP field welds).

(v) Relative to physical sample size, Corporate Welding Manual Procedure NW-16 calls for the removal of about 5 grams of material for this type of analysis. The precise weight of the sample taken was not recorded, but was sufficient to facilitate the testing for which results are provided herein. Relative to the number of welds subject to chemical analysis, three of the field welds in the stainless steel Spent Fuel Pool Cooling piping were subject to composition analysis by both the alloy analyzer and chemical analysis of chip samples. Note that the purpose of subjecting these three welds to chemical analysis was not to provide inference to the entire population, but

• rather to demonstrate consistency with the alloy analyzer. Since the alloy analyzer does not lend itself to reliable composition analysis with carbon steels, all three CCW field welds will also be subject to laboratory analysis for material composition. The accuracy of the chemical analysis method for each element is listed in the laboratory 9

) Enclosure I to Serial: HNP-99-069 Page 4 of 19.

test report. The laboratory analysis report from the three stainless steel samples already completed is included in Enclosure 4.

3 Reauested Item 1 A.4 In the matrixes or on the isometrics, identify inaccessible non-embedded welds and 3 embedded welds (all other welds should be accessible).

Resnonse to Reauested Item I.A.4 The isometrics are marked up to show which field welds are embedded and thereby 3 inaccessible (Enclosure 2). All field welds which are not embedded are extemally accessible.

Reauested Item I.A.5 D

On the isometrics, indicate the specific location of each weld listed in I.A.2 and identify the boundaries of the systems that are considered safety related. Identify all non-safety related items that appear on the isometrics.

Response to Reauested Item I.A.5 3

The isometrics are marked up accordingly (Enclosure 2).

Requested Item I,A,6 3

(i) Identify in the matrixes, or on the isometrics, the welds that will be or have been inspected or re-inspected that have Code documentation, welds that have been inspected that do not have Code documentation, and welds that will be or have been inspected or re-inspected not to Code. (ii) For the welds that will be or have been 3

inspected or re-inspected but not to Code, describe the inspection technique, mptsce criteria, and documentation. (iii) Identify the edition and addenda of ASME Code that will be or has been used for the above inspections and re-inspections.

] Response to Reauested item I.A.6 (i) Code documentation for welds performed by the piping vendor are included in the vendor data packages. As noted in the Alternative Plan (Enclosure 8 to HNP-98-188, y dated 12/23/98), this accounts for approximately 160 of the roughly 200 welds in the large bore Spent Fuel Pool Cooling piping. Based on available evidence, all of the 40 piping field welds and the 12 hanger attachment pad welds were inspected to Code requirements, but generally do not have the Code required documentation available.

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Enclosure 1 to S! rial: HNP-99-069 Page 5 of 19 Documentation which is on hand for these field welds is listed on the matrix prepared in response to requested item I.A.2. (Enclosure 3).

1 (ii & iii) The accessible field welds within the scope of the Alternative Plan have been re-inspected using original surface examination criteria from ASME Section III,1974

- winter 1976 Addenda, ND-5000. A portion of the inaccessible (embedded) field welds will be subjected to intemal inspections using a high resolution, remotely operated video camera mounted on a pipe crawler. Details of these camera

) inspections, including inspection technique and acceptance criteria, are provided in response to requested items III.3 & III.4.

Reauested Item I.A,7

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Identify any non safety related items installed during the original construction which will be upgraded to safety related status by this amendment; e.g., will any of the non-safety-related ANSI B31.1 piping (Enclosure 8, page 7 of the submittal) be upgraded?

) Resnonse to Reauested Item I.A.7 No such items installed during original construction will be upgraded for use in a Code application in support of this activity. No B31.1 piping will be upgraded for use in a Code or safety-related application. The tumover of piping and equipment within the

) scope of this activity will include a review of all Code items and documentation by the ANI to ensure that each item has the appropriate certification.

Reauested Item I.A.8

)

Identify any commercial grade items requiring dedication installed during the original construction. For these items, is documentation of the dedication program available for review? Are the dedication packages for items available for review?

3- Response to Reauested Item I.A.8 No commercial grade items were installed during the original construction which will now be used inside Code boundaries. The turnover of piping and equipment within the scope of this activity will include a review of all Code items and documentation by the ANI to

) ensure that each item has the appropriate certification.

Reauested Item I.A.9 h' ' Identify any commercial grade items requiring dedication that will be used to complete construction.

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l J Enclosure I to Serial: HNP-99-069 Page 6 of 19 l Response to Reonested Item I.A.9 No commercial grade items will be dedicated for use in a Code application by this activity.

3 The turnover of piping and equipment within the scope of this activity will include a review of all Code items and documentation by the ANI to ensure that each item has the appropriate certification.

O Recuested Item I.A.10 l (i) Was the piping system constructed in accordance with a 10CFR50 Appendix B Program? (ii) Is the construction Appendix B program documentation available for review? (iii) If construction was performed under a different program, identify the O program. Is this program documentation available for review?

Response to Recuested Item I.A.10 C (i) The overall quality assurance program used by Carolina Power & Light Company for the design and construction of the Harris Nuclear Power Plant is described in the Shearon Harris PS AR. PSAR Section 1.8 states that "The Carolina Power & Light Company Quality Assurance Program for the engineering and construction of the Shearon Harris Nuclear Power Plant (SHNPP), which includes the quality assurance O programs for both Ebasco and Westinghouse by reference, is structured with regard to safety-related equipment in accordance with the eighteen criteria of Appendix B to 10CFR50. In addition, the subject Program is structured in accordance with ANSI N45.2 and thereby Regulatory Guide 1.28 . . . ". The PSAR further states that the "Shearon Harris Nuclear Power Plant Quality Assurance Plan" was mplaced by the

1. "CP&L Corporate Quality Assurance Program" on April 1,1974, and provides a cross reference on how the subject plan met the criteria of 10 CFR50 Appendix B.

(ii & iii) Certain aspects of Shearon Harris Nuclear Power Plant construction were subject to QA requirements beyond those outlined in the CP&L Corporate QA 9 Manual. Since CP&L was not only the Owner, but also the constructor, installer, and a fabricator for Code items in the Sheamn Harris Nuclear Power Plant, a separate QA Program was developed, reviewed, approved and implemented specifically to obtain the required AShE N, NA, and NPT Certificates of Authorization. ASME Code Section III, Subsection NA-4133.2 requires that an

1. applicant for a Certificate of Authorization develop a QA program and implementing procedum specific to the proposed scope of work, and that "the applicant shall request the Society to myiew this procedure and Program prior to the issuance of a Certificate of Authorization." For construction of SHNPP, CP&L met this requirement by the formalization of its "ASME Quality Assurance S Manual", intended to meet the criteria in Section III, Subsection NA-4100 of the e

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Enclosure 1 to Serial: HNP-99-069 Page 7 of 19 Code. All Code work by CP&L during the Construction of the Shearon Harris Nuclear Power Plant was performed to the requirements of this QA program manual. A copy of the ASME Quality Assurance Manualis provided in Enclosure

) 5.

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Reauested item I.A.11

) (i) Are the work control procedures and hold point sign-off documents from the original construction available for review? (ii) If these documents are required by Code, what documents are missing?

Response to Reauested Item I.A.11

) l (i) Work control procedures and hold point sign . documents from the construction era are available for review, i

(ii) With the exception of the aforementioned WDRs and associated weld process control

) issues (including NDE) discussed in response to item I.B.4, CP&L has not identified any missing documents requiring consideration under the Alternative Plan.

l Reauested Item I.A.12 (i) Provide a list of qualified weld procedure specifications (WPS) used, and their procedure qualification records (PQRs) (ii) For welds missing welder identification, how will weld integrity be established.

) Response to Reauested Item I.A.12 ,

(i)The welding procedures available for welding during the original constmetion of the piping in question were identified based on a review of available WPS in the welding manual at that tirne. A copy of these WPS and their PQRs are provided in Enclosure

)- 6.

(ii) CP&L has located welder identification markings at each accessible field weld in the scope of the Altemative Plan. These Code required welder symbols can be traced back to the welder responsible for each such weld, and from there, qualification

) records on file can be used to establish that each welder was appropriately qualified.

These markings are not accessible on embedded welds. However, attemate QC records have been located which identify the welders for three of these fifteen welds, and numerous programmatic and procedural assurances existed to ensure that welds

) were made using qualified welders and weld procedures. For embedded welds, internal camera inspections (as described in response to RAI Items III.2,3 & 4 ) will be used to augment programmatic and procedural assurances relative to the quality of these welds.

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D Enclosure 1 to Serial: HNP-99-069 )

Page 8 of 19 l l

l In addition, since the Spent Fuel Pool Cooling piping nozzles exit into the pools l l

below the water level, the portions of the Spent Fuel Pool Cooling System piping O attached to the spent fuel pools (including the embedded piping) are flooded as well.

Beyond intemal camera inspections, water chemistry in these legs of piping will be analyzed to ensure that Microbiologically Induced Corrosion or other corrosion ,

mechanisms have not resulted in degradation of the integrity of field welds or piping.

3 B. Applicable Regulations for Welds and Piping Systems Within the Scope of the Proposed Alternatives Requested Item I.B.1 J

1. Identify the edition and addenda of Code and any Code cases that were used for original construction of the welds and piping systems. If not the same for all the welds, identify the Code requirements for each weld or group of welds.

O Response to Reauested Item I.B.1 Piping was installed to ASME Section m,1974 Edition, Winter 1976 Addenda. The PSA.R and current FS AR provide the CP&L position on conformance to the requirements of Reg. Guides 1.84 and 1.85 relative to use of Code cases. A review of the N-5 Code J Data Report associated with turnover of Unit 1 SFP piping identifies two Code cases used at some point in its construction; it is reasonable to assume that these same Code cases may have been used on the corresponding Unit 2 piping and equipment. These Code cases are:

3 N-240 " Hydrostatic Testing of Open Ended Piping, Section m, Division 1" -

N-275 " Repair of Welds, Section m, Division 1" Likewise, a review of the Unit 1 CCW N-5 Code Data Report shows these Code cases in association with its construction:

9 N-275 " Repair of Welds, Section M, Division 1" N-224 "Use of ASTM A500 Gr. B and ASTM A501 Structural Tubing for Section m, Class 2,3 and MC" N-224-1 "Use of ASTM A500 Gr. B and ASTM A501 Stmetural Tubing for S Section m, Class 2,3 and MC" N-282 " Nameplates for Valves, Section E, Division 1, Class 1,2 and 3 Construction" N-127 " Alternative Rules for Examination of Welds in Piping, Section m, Class 1 and 2 Constmetion" O

6

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O Enclosure 1 to Serial: HNP.99-069 Page 9 of 19 Reauested Item Q2 Identify the edition and addenda of Code and code cases that will be used to complete O

construction of the piping systems. Identify any exceptions to Code requirements and justifications for these exceptions.

Response to Reauested Item I.D.2 i

O Construction will be completed to ASME Section III,1974 Ed, Winter 1976 Addenda.

Code Case N-240 will be used to exempt formal requirements for hydro testing of the l embedded piping connected to the atmospheric spent fuel pools due to the lack of accessibility. The need to invoke other specific Code cases has not been identified. Use l of any such Code case would be consistent with CP&L's position regarding conformance j O with Reg. Guides 1.84 and 1.85. Relative to exceptions to Code requirements, CP&L does not take any such exceptions beyond those specifically i&ntified and addressed by I this Alternative Plan.

l l

O Reauested Item I.B.3 Identify the edition and addenda of Code and code cases that were or will be used for l repair and replacement of welds and piping.

O Response to Reauested Item I.B.3 No repair or replacement activities have been performed on the Code piping subject to the Alternative Plan. Future repair and replacement activities (after completion of construction and turnover) will be governed by the site Section XI Repair and O Replacement program.

Reauested Item I.BA O Provide a matrix (See I.A.2) that identifies the specific paragraph in Code that is applicable to missing weld documents. Identify documentation deficiencies for each weld. Identify any exceptions to Code requirements. Provide altematives and justifications for these exceptions.

O- Response to Reauested Item I.BA A matrix has been provided in Enclosure 7 for Code requirements pertaining to missing weld documents. Additional information relative to specific welds is provided in Enclosure 3. Alternatives andjustifications are identified in Enclosure 2 and discussed O- elsewhere in the Altemative Plan and this RAI response.

O

D' Enclosure 1 to Serial: HNP-99-069 Page 10 of 19 Keouested item I.B.5 g Identify the ASME requirements, including administrative requirements, that were completed prior to stoppage of the original construction of the piping systems. Is documentation of these completed requirements available for review? What ASME data reports were filed and what were their filing dates?

Response to Reauested Item I.B.5 3

None of the piping or equipment in question had completed the system certification process and received an N-Stamp. Generally, requirements which were met are consistent with the status of construction at the time work was halted. For instance, embedded piping had been installed, inspected and tested prior to pouring concrete, but O accessible piping immediately adjacent was still under constmetion. The availability of records for the construction varies. Generally, records generated by site construction i during the installation of the subject piping is not on hand. However, records generated i as a result of QC oversight (NCRs, DDRs, audits, etc) are on hand and retrievable. J Notably, hydro test records are also generally available for that portion of construction i O that proceeded to the extent of hydro testing, including embedded Spent Fuel Pool l Cooling System piping. Hydro test documentation, including verification of weld documentation, is available for all but 2 of the 15 embedded field welds. The remaining 2 are included in the liner leak test boundary and would have been procedurally required to be verified as complete, but were not specifically included in the leak test as inspection O items. (See Enclosure 3 for identification of records available, and Enclosure 8 for the hydro test records specifically discussed herein.) No partial data reports were filed on the subject piping systems. Manufacturer's Code data reports from NIrr suppliers are available in document control for the subject piping, as are warehouse receipt inspection records. These records will be subject to review by the ANI as part of the system l 0 turnover process.

Reauested Item 1.B 6 O Identify ASME survey inspections conducted prior to stoppage of the original construction of the piping systems. Provide documentation for representative internal /

external audits conducted during the peak construction periods for the welds in question (1978 - 1979), particularly in the areas of work control, welding, material traceability and records.

O Response to Reauested Item I.B.6 There are no documented ASME survey inspections on hand specific to the construction of the piping systems in question. There were, of course, ASME surveys associated with O CP&L obtaining and maintaining its N, NA and NPT Certificates of Authorization. This was originally accomplished by an interim letter of authorization in July,1978 allowing CP&L to commence Code work. A follow up survey on the effectiveness of the program O

D- Enclosure I to Serial: HNP-99-069 Page 11 of 19 was conducted in July of the following year, with additional audits occurring in 1982 and 1985, in accordance with Code requirements.

O Information pertaining to audits and inspections performed by parties other than the ASME is provided in response to requested item I.B.7, below. Also, note that the majority of constmetion for the welds in question occurred during the '81 '83 time frame, as attested to by QC records and other documents associated with this l

construction.

1 Reauested Item I.B.7 Identify third party inspections conducted prior to stoppage of the original construction of O the piping systems. Provide a representative sample of documentation for these inspections.

Response to Reauested Item I,B,7 O A number of ANIinspections specifically associated with the construction of the Unit 2 1

& 3 SFP Cooling piping are documented in the form of QA surveillance records, hydro test records and other types of records which would have been subject to ANI review.

Generally, the ANI inspection records which cannot be retrieved are those associated with WDRs and pipe spool packages. Records for which ANIinspections / reviews are O documented are identified in Enclosure 3.

In addition, Corporate QA / QC, which operated independently of the site construction program, provided both quality inspections of work activities and audits on construction activities. Records for which QC inspections are documented are identified in Enclosure O 3, and representative samples of QA audits of the construction program are provided in Enclosure 9. Finally, the NRC performed regular inspections of construction activities, with follow-up activities being initiated as needed for issues identified and tracked to satisfactory closure.

O Reauested hem I B.8 With regard to piping system components / services performed by others, provide documented validations of these vendors services. Provide the documentation of the D audits of the supplier of prefabricated piping.

Response to Reauested Item I,B.8 A review has been conducted which identifies that Code data reports are on hand for pipe 9 spools and components inside Code boundaries. The tumover process for completion and activation of this portion of the plant will include a review of these documents by the ANI. CP&L intends to replace any piping or equipment provided by an outside supplier for which appmpriate Code records cannot be located. Audit records of the supplier of 9

7 1 Enclosure 1 to Serial: HNP-99-069 Page 12 of 19 prefabricated piping and a representative sample of a piping vendor data package are included in Enclosure 10.

y II. Comoletion of Pinine System (Generah Reauested Item II,1

) (i) Identify the differences between HNP's proposed construction program to complete the SFP C and D and the original construction program under HNT's N certificate. (ii)

How will these differences be reconciled?

Response to Reauested Item 11.1

)

(i) CP&L proposes to complete construction per the design requirements of the original construction Code. CP&L is requesting that exception be allowed under 10CFR50.55a.(a)(3)(i) to certain QA requirements generally found in Section III,

) Subsection NA and associated with having certificates of authorization for construction and installation of Code items, and to requirements regarding N-Stamping of the completed systems.

(ii) CP&L proposes to reconcile the differences between the original program and the program to be used for completion by providing comparable assurances, tests,

) inspections and reviews as needed to assure an acceptable level of quality and safety in accordance with 10CFR50.55a.(a)(3)(i). It is CP&L's intention to complete construction using the current Corporate Appendix B QA Program, augmented by supplemental QA requirements to ensure that the intent of Code requirements are adequately addressed. (See response to requested items III.14,15 & 16).

)

Reauested Item II.2 l

Will data packages be prepared?

Response to Reauested Item II.2 Yes. CP&L is implementing a turnover plan which closely emulates that associated with

. the N-Stamping process, including preparation of Section III style data packages and third

) party.(ANI) review.

Reauested Item II.3

) What third party verification is planned? ,

1 1

i

} Enclosure I to Serial: HNP-99-069 Page 13 of 19 Response to Reauested Item II.3

) The Hartford Steam Boiler Insurance and Inspection Co. has been in discussions with CP&L throughout the development of the Alternative Plan. The role that Hanford will play in the cenification / turnover process is very similar to that which would be followed in an N-stamping process. It is intended that the ANI will review work packages, panicipate in field inspections, participate in resolution of field discrepancies and non-

) conformances, and conduct a final review and certification process much like that done for the preparation of an N-5 data repon for each affected system within Code boundaries. Details of this process are contained in a set of" Supplemental QA Requirements" developed for this activity (See response to III.14). A copy of the generic data report to be used for installation of Code items is provided in Enclosure 11.

J III. Specific Comments on Submitted Information ,

Reauested Item III,1

) (i) What was the basis for selecting the four extemally accessible field welds for internal  ;

examination? (ii) Identify these welds in the matrix provided in response to I.A.2 above. l Response to Reauested Item III1

)

(i) Field welds were generally used to join long sections of prefabricated piping, and so were (are) not typically accessible for internal examination with the naked eye. The four field welds in question join the strainer nozzles to the piping, and were identified by a field walk down as being those field welds which could be accessed without specialized pipe crawling / camera equipment. One of these welds is only a

) few feet away from an open pipe end, lending itself well to visual examination with I

the assistance of an examination mirror. The other three field welds were subject to a more limited inspection by inserting a boroscope through nearby pressure taps. Note that a more detailed internal examination of these welds will be performed and )

formally documented when the strainers are disassembled, using the same internal

) inspection criteria as developed for the remote camera inspection discussed in III.2, I

{

3,4 & 5 below.  !

(ii) These welds are identified on the matrix (Enclosure 2 ) as 2SF-37-FW-441,2SF I FW-449,2-SF 36-FW-450 & 2-SF-38-FW-451.

)  ;

Reauested Item III.2 I With reference to the " substantial portion of the embedded piping and field welds", j identify these welds in the matrix provided in response I.A.2

)-  ;

Response to Reauested Item III.2 i i

These welds have been identified on Enclosure 3 as requested. '

)

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). . Enclosure 1 to Serial: HNP-99-069 Page 14 of 19 y Reauested Item III.3 Provide a summary of the inspection procedure used for remote inspection of embedded I welds.

Response to Reauested Item III 3  ;

3 ..s  ;

The procedure will use a pipe crawler mounted camera to perform a detailed inspection of I the interior surfaces of embedded field welds. The procedure will include demonstration j of camera resolution capability to at least 1/32" wire, and performance demonstration of l inspector's ability to discern and disposition flaws of the nature which might be expected g

to be encountered. The inspection procedure will be developed and approved by a Level 1 III inspector under the Corporate NDE Program. Inspections will be performed by an appropriately qualified level II inspector.

) Reauested Item III.4 l

With reference to the remote inspection of the embedded welds, identify the critical characteristics that will be verified and the acceptance criteria to be used.

3' Response to Reauested Item III.4 The inspection will specifically include examination of field welds for the following:

No cracks

) No lack of Fusion (LOF)

No lack of Penetration (LOP)

No oxidation (" Sugaring")

No undercut greater than 1/32 inch No reinforcement (" Push Through") greater than 1/16 inch 3 No Concavity (" Suck Back") greater than 1/32 inch No porosity greater than 1/16 inch i No inclusions Generalized inspections will be performed on the piping interior for indications of arc

3. strikes, foreign material, high / low, mishandling indications, etc.. Any such indications shall be noted and characterized during the inspection and evaluated by Engineering if necessary.

In addition, since the Spent Fuel Pool Cooling piping nozzles exit into the pools below 1 the water level, the portions of the Spent Fuel Pool Cooling System piping attached to the spent fuel pools (including the embedded piping) are flooded as well. The inspection procedure will also include criteria and instructions to conclusively ascertain if i

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Of Enclosure 1 to Serial: HNP-99-069 Page 15 of 19 Microbiologically Induced Corrosion or other corrosion mechanisms have resulted in l degradation of this piping. I 3

Data Recording - The following information will be recorded for each inspection:

1, The inspection will be recorded on videotape in a manner which will facilitate future review and evaluation. I O 2. Indication location ( circumferential, side of weld, etc.), length, and depth (where ,

I applicable) shall be documented and recorded on tape.

References - The following references were used to establish this criteria:

3- ASME Section III, ND-4424 Winter 76 Addenda ANSI B31.1 Paragraph 136.4.2,1980 Edition Corporate Welding Manual NGGM-PM-0003, NW-02, NW-06 O Reauested Item III.5 Provide results of remote inspection with any identified discrepancies Response to Reauested Item III.5 j D

Camera inspections are currently planned for late May or early June of 1999. Results will l be provided upon completion of this activity.

O Reauested Item III.6 l Provide a completed weld data report, representative of those that were discarded.

Identify the critical characteristics and explain how, in lieu of records, each will be validated.

3 Response to Reauested Item III.6 A sample WDR is provided in Enclosure 12. Note that this is a WDR for one of the 15 embedded field welds, extracted from'a DDR (Deficiency Disposition Report) in which a >

0 QA inspector questioned the identity of the adjacent pipe spool. Code required attributes recorded on the WDR are identified and reconciled in Enclosure 6.

Reauested Item III.7 O

With reference to the procurement specification (SS-021, Purchasing Welding Materials for Permanent Plant Construction), did other specifications for other filler materials exist?

LO i 1

) Enclosure 1 to Seiial: HNP-99-069 Page 16 of 19 What assurances are provided that these other filler materials were not used for the embedded piping.

) Response to Reauested Item IH.7 SS-021 is the site spec for procurement of filler material used in the SHNPP Construction Program and referenced in the Work Procedures which implemented this program. SS-021 is the specification for filler material specifically invoked by Code work procedures;

) no substitutes were identified or allowed. Research has not identified any other specification for this purpose in association with construction of SHNPP. Being a fairly new plant, CP&L still employs many of the weld engineers and craft personnel associated with the original constmetion effort. Numerous interviews of these personnel consistently provide the same conclusion; that filler material purchased by CP&L for use

) in Code work in construction of SHNPP was procured to this specification.

Reauested Item HI.8

) Provide any updates / supplements to the Alternative Plan as they become available.

Response to Reauested Item IH.8 These will be provided as requested.

Reauested Item HI.9

) With reference to the "large percentage of embedded field welds" that will be inspected, identify these welds on the matrix provided. Provide technicaljustification for not inspecting the remaining welds.

Response to Reauested Item HI.9

)

The matrix has been marked up as requested. The "large percentage of embedded field welds" referred to are those which CP&L has a high level of confidence can be accessed with available pipe crawling equipment based on a walk down with the vendor for pipe crawler / camera services. The enclosed weld matrix (Enclosure 3) specifically identifies the base scope of field welds which are targeted for inspection. Currently,6 of the 15 embedded field welds are included, which notably includes both of the field welds for which hydro test records are not available.

Assurance of quality for any embedded field welds which are not subject to remote

) camera inspection is provided by conformance to the requirements of QA Program (s) and implementation procedures which existed at the time of construction along with the body of evidence which directly support adherence to those requirements. This evidence includes: uniform application of QA requirements for the entire site construction 1

e b Enclosure 1 to Serial: HNP-99-069 Page 17 of 19 program,(including the completed and licensed Unit I facility), surveys, inspections, and audits verifying the effectiveness of QA program requirements, construction records which are on hand that attest to quality of construction, and re-performance of Code 4 P required inspections on accessible field welds in these same lines with no rejectable l

l indications identified.

I Reauested Item III.10 (i) Explain what is meant by the statement that internal examination of the embedded welds provides a measure of quality assurance beyond Code requirements. (ii) What additional physical or material attributes will be verified? ,

O Response to Reauested Item III.10 l

(i) This statement is simply intended to identify that many of these welds would have been inaccessible for routine internal inspection at the time of construction (due to I distance from an open pipe end), and since no Code requirements existed to do so, O would not have been subject to an intemal visual examination. Given this, internal camera inspections represent an activity above and beyond that which would have been ,

required under the original construction program. I (ii) See response to requested items III.3 & 4.

'O Reauested Item III,11 l

The submittal refers to opinions by Bechtel and Hartford concerning the benefits in accordance with an N certificate program. Are these opinions documented and available

O for review?

I Response to Reauested Item III.11 1

Hartford's endorsement of the Alternative Plan is provided in Enclosure 13. Note that i O this letter is authored by Dr. Richard E. Feigel, Vice President of Hartford Steam Boiler Inspection and Insurance Co. and Chairman of the ASME Council on Codes and Standards. Bechtel's endorsement of this plan is implicit in that they, as the design A/E, have fully reviewed and incorporated the Alternative Plan into the design change packages for this activity.

O {

Reauested Item III,12 i Provide a copy of the site ASME Section III QA program used during original O construction.

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O Enclosure 1 to Serial: HNP-99-069 l Page 18 of 19 L Response to Reauested Item III.12 ,

l O. A copy of the ASME Section III QA Program manual is provided in Enclosure 5.

I Reauested Item III.13 (i) Provide a copy of the Corporate QA program that will be used to complete construction. (ii) (Provide a list of implementing quality control procedures for welder qualification, weld procedures, inspections, documentation, etc). j Response to Reauested Item III.13 lO (i) A copy of the current Corporate QA Program Manualis provided in Enclosure 14.

l Note that this program manual is used with FSAR Section 17 to define the overall corporate QA program.

(ii) All welding will be accomplished in accordance with the Corporate Welding Manual, which conforms to the requirements of Section IX with regard to welder qualification, lO

l weld procedures and process control. NDE will be performed in accordance with the Corporate NDE Manual. The site Mechanical Modification Procedures (MMPs) are those procedures which will primarily be used to control work control processes. The list of MMPs most applicable to this activity and the index from the Corporate

. Welding and NDE Manuals are provided in Enclosure 15.

O j l

Reauested Item III 14 Provide a copy of the supplemental quality assurance requirements developed to augment O

the Corporate QA Program, which was based on a review of the approved Construction QA Program at the time of construction versus the existing Corporate QA Program.

Response to Reauested Item III,14 O Supplemental QA Requirements are provided in Enclosure 16.

1 Reauested Item III.15 O Provide documentation of the referenced comparison of approved ASME Section III

) Construction QA Program Manual with the effective Corporate 10CFR50 Appendix B QA Program.

Response to Reauested Item III.15 O

Documentation of the referenced comparison is provided in Enclosure 17.

O L

r-e Enclosure 1 to Serial: HNP-99-069 Page 19 of 19 Reauested item III.16

• Provide documentation of the supplemental quality assurance requirements that have been developed specifically for the purpose of addressing differences between ASME Section m quality assurance requirements and the Corporate 10CFR50 Appendix B QA Program.

• Response to Reauested Item III.16 The ASME Section m QA Manual discussed in response to requested items E.14 and m.15 above is the document which was reviewed by the ASME and singularly credited for assuring compliance with Section E requirements in order to authorize CP&L to 8 perform N, NA and NPT stamp activities. The overall corporate QA program may have shared procedures, facilities, etc. with this program, but was not directly relied upon to assure compliance with Section m during the construction effort. Given this, the Supplemental QA Requirements provided in response to requested item E.14 and the QA manual comparison provided in response to item requested item E.15 provide the O documentation requested in this item as well.

3 O

O 3

O

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Enclosure 3 to Serial: HNP 99-069 Page 1 of 13 O '.-

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D Enclosure 13 to Serial: HNP-99-069 D

D Alternative Plan Letter of Endorsement from D The Hartford Steam Boiler Inspection and Insurance Co.

(1 Page)

D D

D D

)

?

c ..

i 1

) MAR- 8-99 MON 17:19 HSB AT ENG FAX N0. 7703926252 P.02 l NAR 08 1999 16817 FR HTFD STEAR 1 DG!tEER 660 722 5530 TO HSBT-AT P.0141 l

, Dr. Riche <d E. Fdtd Y** W O*

Poet 4t'Fex Note 7671  %> /

) ZiA A f-O bico -

'""GGAW. f9cfA_ -

c. = f c j e., eu ,  ;

The briterd Sea.a Route l

) i - u ,. c mume -

I i

o s s., 1 M*d" m m.ssu m REGSVED i

'"'**"5"* 8 1999, WAR.

March 8,1999 ,

i Mr. Steve Edwards i Manager,SFP Activation Project Caro 11aa Power A LightCompany Harris NuclearPlant  ;

) P.O. Box 165 New HillNC 27562 i l

Subjea:10CFR50.55: Ahernadve Plan HH/99-001 j HNP-98-188 o.

) l

Dear Mr. Edwards:

I have reviewed your letter to Mr. Bobo and the referenced anschments addressing various spent fuel pool piping systems. ! have discussed the subject at length with Mr. Bobo, who is in responsible direct che s ,f j Hartford Steatn Boilefs (HSB) ASME Section !!I and XI inspection activities. Subject to detailed {

) verification ofcomptedon ofim&i<hed tasks and the*r compliaace with cornmhments described, we believe that the plan proposed provides an acceptable ahernadve to code compliance in accordance with 10CFR50.55(aX3). Our concurrence extends to both dispositioning issues related to the as-built condition of the sy* ns and fdww 6edvides undar '10. Alternadve Plan for Continuance of Design and i construedon.*

Our posItton is based poncipaDy oc the following:

I

1. She wide use of au lategrated QA progrun at the site with evidence of adequacy provided by liceoslog of Unit 1.

2. C~ W reference by final acceptance documentadon, e.g. hydrostatic test reports, of first tier l WW reports whk.h establish review ofrecords of welder qualificadon and similar code 6 requirecuents.

) 3. Plan provisions to verify code compliance or establish techn! cal equivalency .e4 deposited weld metal analysis.

Very truly yours, b

hf&

IUchard E. Psigel, Ph.D.

Vice President, Engineering Cc: B. Bobo R. Howard

) LOGGED

O Enclosure 16 to Serial: HNP-99-069 Page 1 of 15

,O O

Supplemental Quality Assurance Requirements for the Design Change' Packages Associated with Completion of the Units 2 & 3 Spent Fuel Pool Cooling System

'O O

O O

O O

O-O

1 l

) Enclosure 16 to Serial: HNP-99-069 Page 2 of 15 SUPPLEMENTAL QA REQUIREMENTS l

) The following is a set of supplemental QA requirements developed for the implementation and turnover of Code items associated with the completion and activation of the Unit 2 & 3 j l Spent Fuel Pools at Harris Nuclear Plant. This document will be incorporated directly into l the " Design Requirements" section of the design change packages for the pertinent modifications, and then by specific instructions in the appropriate sections (installation, l

) testing, turnover, etc) as necessary to ensure that its requirements are implemented.

1.0 GENERAL  !

1.1 Scope This document defines the set of QA requirements which will be used to govem the engineering, constmetion and startup of the Section E, Class 3 portions of the Spent Fuel Pool Facilities originally intended to service HNP Units 2 & 3. This portion of the plant was partially installed during original plant construction, but was suspended subsequent

) to cancellation of these units. The development of a supplement specific to this scope is necessitated by the following concerns:

a The original N-certificate associated with this program has long since been

, discontinued, and no partial tumover was conducted.for the partially installed piping 3 and equipment.

a The field construction documentation packages for partially installed piping have been discarded and are no longer available As a result of the above, it is not possible to complete these systems in full compliance

) with Section III utilizing the previously installed piping and equipment. Since the N l stamping process is the prescribed method for demonstrating quality assurance in l constmetion activities, it is necessary to define a suitable alternate program which will ensure that the requisite level of quality exists upon completion and tumover. Generally,

! the corporate Nuclear Generation Group's Quality Assurance Manualis of suitable rigor j to accomplish this. However, the program defined in the corporate QA manual was l developed to comply with 10CFR50 Appendix B as it concerns operating plants, and was not intended to specifically conform to the requirements of Section m. For example, the corporate QA program outlines condition reporting requirements which govern field activities and meets the requirements of Appendix B in this regard. However, this D program does not integrate involvement of the ANI in documenting adverse conditions, nor does it require the ANI to participate in the closeout of adverse condition reports. In addition, the current site procedures pertaining to field activities are generally oriented towards meeting the requirements of Section XI for inservice inspection, rather than Section E.

3 To address this issue, a set of QA requirements have been developed and are presented herein to supplement the corporate Appendix B QA Program. Generally, these requirements were the result of a review of the current corporate Appendix B Quality Assurance Program against the requirements of the approved ASME Section m QA 3

L'

o q l

Enclosure 16 to Serial: HNP-99-069 '

Page 3 of 15 Manual utilized for construction of the Harris Nuclear Plant. These requirements are not intended to delete or revise any requirements in the corporate QA manual, but rather are j

) to provide additional criteria in supplement of the existing program. These criteria will be implemented in one of the following manners:

Revision of site procedures: Since this supplement is not intended to contradict approved site procedures, this might be necessary to reconcile conflicts between the

) Supplemental QA Requirements and that of existing site procedures.

1 Incorooration through the work control orocess: When criteria are stipulated that are not ,

already reflected in site procedures, it may be more suitable to add these through work l planning and specific instmctions in the work package. The requirements for additional l

) involvement of the ANI would be an example of this.

Procedure revisions will be reflected by markups and inclusion on the Document Update Form (DUF), while work package implementation will be accomplished by specific instruction in the appropriate section of the modification package (implementation,

) testing, etc.).

1.2 Responsibilities

)

General- Programmatic responsibilities for implementation of the Corporate Appendix B program, including the site'sSection XI Repair and Replacement Program, are as defined in the Corporate Quality Assurance Program Manual and supporting documents, including site procedures. The involvement of site organizations as pertains to the

) implementation of these supplemental requirements will be subject to their review and approval during the modification approval process.

AIA (ANI) - The Authorized Inspection Agency is responsible for providing' the support necessary for implementation of the supplemental requirements described in this ESR.

j Acceptance of these requirements will be based upon NRC review and approval of the 10CFR50.55a Altemative Plan. Formal AIA endorsement of these supplemental l requirements from a programmatic perspective will accomplished by review and approval of the modification packages which incorporate them.

1 j Modification Engineer - The Modification Engineer for the affected ESR is responsible l for implementing the requirements found herein in the most appropriate manner. This would include either revision of site procedures or through direct incorporation into the modification package, as described above.

) Modification Responsible Engineer - This supplement pertains only to modification activities completing constmetion of the spent fuel cooling systems originally intended to service Units 2 & 3. As such, the ultimate responsibility for adherence for this rests with the RE for these modifications. Since this supplement will be incorporated into the

)

f

[

)

Enclosure 16 to Serial: HNP-99-069 Page 4 of 15 modification packages, the RE is responsible for ensuring that the modification package contains sufficient instructions and guidance to implement it as written.

2.0 DESIGN AND DOCUMENT CONTROL 2.1 Design Control

)' Design Control over the modification design is directed and coordinated by CP&L in accordance with corporate and site procedures goveming the modification process and design activities by outside organizations. This process results in rigorous design review process (including independent design verification) by the A/E and detailed owner's reviews by CP&L engineering personnel.

This supplement pertains only to modification activities completing construction of the spent fuel cooling systems originally intended to service Units 2 & 3. Generally,it is intended that completion of this portion of the plant will be govemed by the same revisions of the Code that were utilized for original design and construction. To that end,

), the applicable version of the Code associated with a particular aspect of construction, and the boundaries of that applicability shall be clearly defined as design inputs in the modification packages. Later versions of the Code may be used only with reconciliation of any differences between it and the Code that was utilized for original design and construction.

)

2.2 Design Specifications 2.2.1 Design specifications will be prepared for all Code stamped items, in accordance with corporate and site procedures, and will be subject to the following

) requirements:

E The specification shall clearly delineate Code classification and boundaries and the pertinent code revision associated with the item.

5 The specification shall address Code requirements for data reports, including

) any that may pertain to transmittal to enforcement authorities.

5 The specification shall fully conform to Section III design requirements.

5 The design specification shall be certified to be correct, complete, and in compliance with the code by one or more Registered Professional Engineers competent in the applicable field of design of components and related nuclear

) power plant requirements. It is noted that some of site's existing design specifications date back to the construction era, but may have been revised since the plant began operation. In these instances, it is acceptable to use previous certified revisions of des.ign specifications, so long as a reconciliation of any subsequent revisions is performed to assess design impact and

) integration into the cutrent the Appendix B Program.

)

p h Enclosure 16 to Serial: HNP 99-069 Page 5 of 15 l

l 2.3 Design Control

)-

l 2.3.1 Design control shall be as directed in the corporate QA program as itnplemented by corporate and site procedures.

I 2.3.2 Design of Code stamped items shall conform to the version of the Code which D would have been utilized during original plant construction. Later versions can be l utilized only with documented reconciliation. Design criteria of Section E, I

Subsection ND shall apply to all Class m piping, equipment and components. I 1

l 2.3.3 Subsequent revision to the affected modification packages shall also be subject to D the supplemental requirements defined herein through completion of construction and the turnover process. l 2.3.4 This supplement is " frozen" as it is incorporated into the 10CFR50.55a

, Attemative Plan and approved by the NRC. Design changes and modification-b revision packages shall not delete or revise the content or applicability'of these supplemental requirements, in whole or part, without NRC approval.

2.4 Applicability of existing site procedures b 2.4.1 It is appropriate to use the site Section XI Repair and Replacement as a guide for j integration of site procedures with the construction of Code related items.

l Generally, existing site procedures shall apply as if the Code portions of construction were being performed as a Section XI Repair and Replacement activity. However, where this supplement contradicts existing procedure or j program requirements, the requirements in this supplement shall take precedent and the affected procedure or program be revised as appropriate.

2.4.2 Welding, including weld procedures, welder qualification, weld material control, use and control of welder ID symbols and preparation of Weld Data Reports, will 3 be done using the Corporate Welding Manual as invoked and implemented -

through site procedures.

2.4.3 The ANI shall have the opportunity to review procedures, including those for welding and QC, which will be utilized for Code related construction activities 9 during the review of work packages prior to field issuance. Ukewise, any revisions to these procedures which is intended to be utilized in the work package i- subsequent to the initial ANI review shall also be identified to the ANI for his review prior to its use.

2.5 Document Control 3

~2.5.1 Document Control will be as currently defined in the corporate Appendix B QA l

program for quality related activities and implemented through site procedures.

f

Enclosure 16 to Serial: HNP-99-069 Page 6 of 15 2.6 Identification of ASME code Documents 3 2.6.1 Purchase requisitions, purchase orders, procedures and other documents generated and / or used at the site for fabrication and installation of Code items shall be identified as "ASME Section III".

3.0 PROCUREMENT D

3.1 General

-3.1.1 The A/E may provide input into the procurement process, however, all procurement will be performed by CP&L under its existing Appendix B Quality 3 Assurance Program and implemented by corporate and site procedures.

3.1.2 Procurement of all code stamped items will be accomplished using approved design specifications certified by a Registered Professional Engineer competent in nuclear power plant design.

D 3.2 Service Contracts 3.2.1 Service Contracts intended to obtain services associated with the engineering or construction of piping and equipment affected by this supplement shall be subject 3 to all the rules and requirements of this supplement.

3.3 Code Stamped Items 3.3.1 It is intended to complete construction to the version of the Code to which the 3 system was originally designed and specified, which governed construction of the existing portion of piping and equipment installed during initial plant constmetion. The applicable version of the code associated with a particular aspect of procurement or construction and the boundaries of that applicability shall be clearly defined in the modification package. Code stamped items shall be 3_ clearly identified as such in the modification BOM or the Equipment Commissioning List. Code stamped items shall be specified and procured so as to fully comply with Code requirements, including the use of qualified suppliers with appropriate Code certification, and shall be stamped in accordance with code requirements.

9 The BOM or the Equipment Commissioning List shall, as a minimum, contain the following information regarding Code stamped items:

Commercial information which sets forth items, quantities, terms, conditions, etc.

, as appropriate, as well as the approved Design Specification (s) which defines the engineering and quality requirements.

3.3.2 Any exceptions to the Design Specifications taken by the supplier with regard to a Code stamped item shall be reconciled by revision to the affected Design S

FV k Enclosure 16 to Serial: HNP-99-069 Page 7 of 15 Specification prior to proceeding with procurement. Any such revision to the l Design Specifications would be prepared, reviewed and approved as set forth for

!O the original specification.

l 3.4 Qualification of Suppliers 3.4.1 Qualification of Suppliers of materials and services shall be accomplished in

,O accordance with the existing CP&L Appendix B Program in accordance with l approved plant procedures. All suppliers must be verified as being on the l approved supplier's list for the scope of supply and holding active certification l from the ASME for any Code items being procured.

lO 4.0 RECEIVING INSPECTION l

4.1 Code stamped items

! O- Inspection, examination and acceptance of Code items shall be accomplished in accordance with corporate and site procedures. Receipt activities shall be documented in the form of a Receipt Inspection Report (RIR). Items accepted shall be appropriately tagged / labeled.

lO Nonconformances noted during receipt inspection shall be reported via Condition Report (nonconformance) initiation, and the affected items placed on hold or rejected. When the vendor's data package is missing or deficient, the item will be placed on hold pending the delivery of the missing information or resolution of the deficiency.

O' When conditions warrant, Conditional Release requests may be granted to permit progression of work involving a nonconforming item awaiting resolution. When this occurs, it will be processed and approved in accordance with existing site procedures. The ANI will be provided with the closure documentation for any

!O. conditional releases affecting Code stamped items or Code related construction.

l l l 5.0 STORAGE AND PROCESS CONTROL g 5.1 Storage Storage requirements for Code stamped items will be clearly identified in the Design Specification. Storage control through manufacture and shipment will be governed by the procurement process.

O l

O

I 0 1 Enclosure 16 to Serial: HNP-99-069 Page 8 of 15 {

1 5.2 Equipment Commissioning Plan 5.2.1. General 1

This section prescribes the methodology which will be followed in commissioning previously installed equipment in support of completing and activating the C & D O Spent Fuel Pools. The subject equipment was installed during the original site construction effort for Unit 2 & 3 fuel storage and handling activities, and was spared in place when these units wem cancelled. This equipment was never incorporated into the operating unit nor has it been formally maintained under controlled storage conditions since that time. Note that the equipment in question O (including Code related equipment) was procumd to applicable design and quality l

l assurance requirements, and this plan does not take exception to any of these requirements. Rather this plan prescribes a set of criteria which will ensure that the equipment in question will meet the applicable requirements of Appendix B l and is capable of performing its intended function in the completed design.

O 5.2.2 Field Walkdown / Scope Development Scope development is accomplished by performing a detailed field walkdown and O comparing the modification design to the field condition. The entire list of previously installed equipment (both Code and non-Code related) which is anticipated to be used in the completed design will be compiled to comprise the scope of the Equipment Commissioning Plan. Note that this plan is not limited to mechanical equipment, and will include civil (pipe supports, penetrations), I&C O (instrument racks, instrumentation, tubing) and electrical (cables, conduit, cable trays, equipment ground connections) as well. Each item in scope will be identified and individually dispositioned in the modification package.

O 5.2.3 Document Review / Retrieval A document retrieval and myiew process will be included in the matrix of commissioning requimments to ensure that required quality assurance information is on hand. Generally, equipment commissioning matrix documentation O requirements will be consistent with that of the original procurement effort. In particular, all Code documentation requirements (including Code data reports) must be satisfied for Code items. Records required for commissioning fall into one of two categories, which are discussed as follows:

O (a) Procurement Documentation This documentation pertains to the information which was originally used to procure the equipment in question and the vendor quality packages which were supplied with the item in response. These records are required to establish traceability and verify that required vendor quality assurance documentation and 0:

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Enclosure 16 to Serial: HNP-99-069 Page 9 of 15 quality releases are on file. Generally, this information is available in the Receipt Inspection Report (RIR) generated at the time the item was received. It is Ho.t

,O acceotable to assume that the necessary information must have been received and is in order by virtue of its being installed in the Celd under control of the construction program, as it would have been possible to have issued the item to l the field with a conditional release with outstanding quality related issues I pending. All Code equipment must have traceability to the Code Data Report (s)

O for its construction.

(b) Field generated records Construction records must be reviewed to ascertain to what extent the existing field condition was documented as being complete and satisfactory. Generally, O this information exists in the equipment installation packages and has been maintained in document control for the major pieces of equipment in question.

Once the equipment installation records have been retrieved, these must be compared against the field condition to verify that the installation as accepted has not been subsequently altered. Previous construction activities can be accepted )

O for use in the modification implementation effort to the extent that required I

installation documentation exists and is verified to conform to the field condition.

In the event that records are found to be missing or deficient, an assessment is performed to detemiine what installation can be accepted by virtue of retest or re- I O inspection, or by use of alternate methods of verification. Alternately, the j implications of the documentation deficiency can be evaluated to determine the I potential impact to quality. Any such evaluation used to accept field conditions in the absence of required information must be formally documented and subject to design review as appropriate. Except as specifically provided in the 10CFR50.55a O Alternative Plan for records of field installation of piping, this equipment commissioning plan is not intended to take exception to Code requirements pertaining to equipment installation or documentation requirements. Given this i

single exception, an evaluation of a deficiency is not allowed to stand in lieu of installation records which are deemed to be specifically required by Section III of O the ASME B&PV Code.

5.2.4 Development of examinations, tests and acceptance criteria O The Equipment Commissioning Matrix shall specify any additional activities necessary to ensure the requisite level of quality assurance in light of the lack of formal controls on storage and handling since this equipment was initially installed. Development of these actiyities willinclude the following:

O E Field verification of equipment identification against procurement documentation.

In the case of Code related equipment, traceability will be established to the Code Data Report (s) and National Board Registration.

E Physical inspections, testing, etc., as required to verify that lack of controlled storage conditions and regular maintenance has not caused any condition affecting 03

O Enclosure 16 to Serial: HNP-99-%9 Page 10 of 15 quality. Commissioning criteria shallinclude consideration of corrosion, fouling, aging, radiation exposure, etc. For Code requirements, any degradation identified O would be assessed in terms of Code requirements, with acceptability based on demonstrated compliance with those requirements.

5 Physical inspections and considerations necessary to ensure that plant activities since constmetion have not resulted in any condition potentially adverse to quality

. (scavenging of parts, introduction of foreign material, damage from personnel and O equipment traffic, etc). For Code equipment and piping, these criteria will specifically consider Code required attributes, with acceptability based on full Code compliance.

O 5.2.5 Repair of Deficiencies ,

Repair of any deficiencies shall be done in accordance with approved procedures. I Since Code items in the scope of this equipment commissioning plan are supplied  !

as completed Section III components from the vendor under that vendor's NFr O Stamp Program, repairs to these items meet the definition of " Repairs" in ASME Section XI and shall be accomplished under the site'sSection XI Repair and l Replacement Program 5.2.6 ANIInvolvement O

Code stamped equipment and related commissioning requirements will be specifically identified as such in the modification package in order to facilitate the system certification process. Provisions shall be made to ensure that any work )

packages generated to commission Code equipment are made available for ANI O review subsequent to work completion.

5.2.7 Revising or Altering the Equipment Commissioning Plan  !

Generally, this equipment commissioning plan does not take exception tc Code or O quality requirements, but rather prescribes a dedication process which will ensure that all such requirements are met in light of the lack of storage control for the equipment it addresses. The sole exception is with regard to field installation records for Code related piping, which are no longer available and are the subj et of a 10CFR50.55a Alternative Plan currently under review by the NRC.

O Acceptance of the field installation of this piping is contingent upon approval of this Altemative Plan by the NRC, and revising the Equipment Commissioning Plan with regard to piping acceptability may require prior notification of the NRC.

Otherwise, this plan does not take exception from design or quality requirements (including ASME Code requirements), and authorization for its use and any O revisions to it are provided under 10CFR50.59.

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Enciosure 16 to Sedal: HNP-99-069 Page 11 of 15 l 5.3 Process Control

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Process control sheets are utilized to establish measures to ensure that processes, I including welding and heat treating, are controlled in accordance with the Code and are accomplished by qualified personnel using qualified procedures.

Generally, process control sheets for Code related construction activities will be i

as provided for under the site's procedures. Additional process control sheets are p fo'ind in the Corporate Welding Manual and Corporate NDE Manual, as invoked j and implemented by site and corporate procedures.

L The ANI will review process control sheets for code related construction activities before they are issued to the field for construction. The ANI will have the O opportunity to add any inspection hold points deemed necessary at this time. All process control sheets for Code related coastruction activities will be reviewed and accepted by the ANI subsequent to completion of field activities.

The hydrostatic test pressure used for pressure testing shall be required to meet O Section III requirements, as opposed to those specified in Section XI. The process control sheets for hydrostatic testing shall reflect the more stringent test criteria.

Nonconforming field conditions will be controlled by site work process control and condition reporting procedures. The ANI will be notified of any condition b reports initiated against code related construction activities, and will verify any such items are resolved prior to signing off the process control sheets for final acceptance.

Identification tags or markings shall be retained on each code item. When it is O necessary to cut or transfer an item during code related construction, material identification shall be transferred to the affected piece prior to cutting. This activity shall be witnessed by QC and appropriately documented in the work package.

D 5.4 Modification Implementation Procedures 5.4.1 Modification procedures are being utilized for code construction (in the context of l this ESR) will be those presently existing for use with the site'sSection XI Repair and Replacement Program, subject to the supplemental requirements prescribed O herein.

5.5 Start-up Procedures 5.5.1 Detailed start-up procedures will be developed and included in the affected modification package. Review of start-up procedures, including QC review, will be documented by review and signature approval as part of the modification approval process.

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Enclosure 16 to Serial: HNP-99-069 Page 12 of 15 6.0 WELDING CONTROL

): 6.1 General Welding activities associated with Code constmetion, including welding procedure qualification, weld materials procurement and control. welding equipment control,

. qualification of welders, weld process control and post weld heat treatment activities shall

)' be controlled in accordance with the Corporate Welding Manual by the Plant Welding Engineer and the Plant Operating Manual. Welding may be performed by Contractors provided that the contractor is fully qualified to CP&L's welding program for the specific welding or welding related activity being performed.

J Contractor's not qualified to and working under CP&L's Corporate Welding Program may only be used for Code welding activities for win:h they maintain their own program

' having the appropriate ASME certification. In this case, a service contract must be provided which authorizes the Contractor to invoke his program for the subject scope of work.

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Work packages involving welding activities associated with Code constmction will be reviewed by QC and the ANI prior to field issuance to ensure that appropriate hold points are included. Weld Data Reports shall be generated for any such welds per the Corporate Welding Program, and hold point inspections shall be accepted by QC and the ANI by

) signature and date on the WDR.

7.0 CONTROL OF EQUIPMENT, TOOLS, GAUGES AND INSTRUMENTS 2

7.1 General Equipment, tools, gauges and instruments specified for calibration control shall be identified, stored, calibrated, and maintained in accordance with site procedures.

3- Calibrations and adjustments shall be accomplished at prescribed intervals and against certified standards having known valid relationships to national standards. If no national standard exists, the equipment manufacturer's recommended standard shall be used.

Recalibration shall be performed any time the accuracy of an instmment is suspect.

3 Traceability shall be maintained between the instrument and equipment or item being tested. The instrument identification number shall be recorded on the appropriate process control documentation. In the event an instrument is found to be out of calibration, a Condition Report must be initiated and an evaluation shall be performed to identify and disposition any suspect inspections, examinations, and test results.

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Enclosure 16 to Serial: HNP-99-069 Page 13 of 15 8.0 INSPECTION, TESTS and NONDESTRUCTIVE EXAMINATION (NDE)

O 8.1 General NDE activities associated with Code construction, including NDE procedures, qualification of personnel and control of inspection and test equipment shall be accomplished as provided in the Corporate NDE Manual. NDE procedures and O acceptance criteria are provided in the Corporate NDE Manual for both original construction code and Section XI requirements. NDE shall be performed on all Code related construction activities in these modifications consistent with Section III requirements, and all such NDE shall utilize Section III acceptance criteria.

O 8.1.1 Process Control Inspection, test and examination requirements shall be defined in the work packages and documented on appropriate process control sheets. These packages will be reviewed by the QC and ANI prior to field issuance. Work will not O Progress past established QC and ANI hold points until the hold point is accepted by signature and date by the QC inspector or ANI.

8.1.2 ANI Review and Approval of NDE Documentation o Records of inspections, tests and examinations containing QC and ANI hold points will not be considered completed until all such hold points are satisfied and the ANI has completed his inspection and signed and dated the process control sheets.

O 9.0 CODE DATA REPORT AND CERTIFICATION 9.1 General O

The piping systems completed under these modifications will not be eligible for N stamping due to issues pertaining to the discontinuance of the original construction program and missing documentation. However, these systems will undergo a certification process similar to N stamping. Installation of Code piping, equipment and O components will be documented on an ASME Section III data report " equivalent form".

This form will be comparable to an NIS-2 form associated with Section XI repair /

replacement activities, and PLP-605 can be used as a guideline for its completion. All work packages for installation of Code equipment shall be clearly identified as such, and provided to the ANI for review prior to field issuance and again upon completion of work 8 tivities C mP etedl and approved documentation pertaining to Code related O

construction, including field generated records and vendor data packages, shall be compiled in packages pending the review of the ANI for system turnover.

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Enclosure 16 to Serial: HNP-99-%9

. Page 14 of 15 i

l The ANI will review the documentation and certify completeness and conformance with

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the requirements of the corporate Appendix B Manual and these supplemental requirements prior to system turnover. Since these supplemental requirements will be implemented either by procedure revision or modification instruction, this certification will be accomplished by verifying that all Code mlated activities were conducted and documented in accordance with site procedures and the requirements of the modification ,

package. The specific list ofitems reviewed to determine completeness and conformance will be provided as an attachment to this certification. Similar to the N-5, this listing will constitute the boundaries of the completed construction which would have normally been N-stamped. '

The completed certification of the affected piping, equipment and components will be included in the modification documentation package as a permanent QA record.

10.0 NONCONFORMANCE AND CORRECTIVE ACTION 10.1 Nonconformance and corrective actions will be addressed within corporate and site procedures, including those associated with procurement, work control and

• condition reporting. Satisfactory resolution of any non-conformances or adverse conditions associated with code stamped items or code related construction activities will be verifiable by the ANI and all other responsible parties prior to turnover.

11.0 RECORDS CONTROL AND RETENTION 11.1 Records control and retention will be as directed by site work control and document control procedures, except as related to the ANI's role in certification as described herein.

> 12.0 AUTHORT7FD NUCLEAR INSPECTOR 12.1 The services of an AIA shrJ1 be used as described herein. It is noted that a qualified ANI would be necessary for Section III construction activities, while an ANII is involved when performing repair and replacement activities under Section i XI. Since elements of both are associated with this modification, dual qualification will be required for the AIA's site representative involved with this modifica: ion. Signoffs for this individual will reflect this dual qualification (ANI

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O Enclosure 16 to Snial: HNP-99-069 Page 15 of 15 13.0 REVIEW, CONTROL AND REVISION OF SUPPLEMENTAL QA REQUIREMENTS O

13.1 These supplemental requirements as incorporated into the modification design and l approved therein will become part of a 10CFR50.55a Altemative Plan and therein subject to NRC review and acceptance. Since NRC acceptance for the attemative plan represents the authorization for these supplemental QA requirements, O revision to these requirements can only be accomplished by submittal and review ,

of the NRC as e revision to the Alternative Plan. Exceptions would be allowed l only for revision to items which comply with all Code and Regulatory requirements and are provided for completeness and clarity (see Equipment Commissioning Plan), or administrative or clerical changes which do not affect O technical requirements.

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