A(~CESSION 'BR:"F*CIL"50-410

AUTH. NAME'-'ANGAN>C. V.

REC IP. NAMEBERNEROi R.

REGULATORY INFORMATION DISTRIBUTION S TEM (RIDS)

8602200235 DOC. DATE: 86/02/18 NOTARIZED: YESNine Mile Point Nuclear Stationi Unit 2i Niagara Moha

AUTHOR AFFILIATIONNiagara Mohawk Power Corp.

RECIPIENT AFFILIATIONDivision oF Boiling Water Reactor (BWR) Licensing

DOCKET 505000410

SUBJECT: Informs of decision to seek schedular exemption to allowoperation during performance of confirmatory analyses of

. containment downcomers. Requirements under 10CFR50. 12Fulfi lied. AfFidavit 5 ap p l ication ~ enc l.

DISTRIBUTION CODE: BOOID COPIEB RECEIVED: LTR .ENCL SIZE:TITLE: Licensing Submittal: PBAR/FSAR Amdts 5 Related orrespondence

NOTES:

RECIPIENTID CODE/NAME

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NIAGARAMQHAwKPowER coRPoRATI0N/300 ERIE BoULEYARDwEsT, sYRAGUsE. N.Y. 13202/TELEPHONE (315) 474-1511

February 18, 1986(NMP2L 0619)

Mr. Robert Bernero, DirectorDivision of BNR LicensingOffice of Nuclear Reactor RegulationU.S. Nuclear Regulatory CommissionHashington, D.C. 20555

Dear Mr. Bernero:

Re: Nine Mile Point Unit 2Docket No. 50-410

As a result of your January 31, 1986 letter to Mr. B. G. Hooten, NiagaraMohawk Power Corporation has decided to seek a schedular exemption to allowfurther analysis of the containment downcomers for the Nine Mile Point NuclearStation Unit 2. The enclosed application demonstrates that all requirementsunder 10 CFR 50.12 for the issuance of the requested exemption have beenfulfilled. The schedular exemption would allow operation of the unit duringthe period that the confirmatory analyses are being performed. Furthermore,the application requests that the Nuclear Regulatory Commission permit anychanges to the facility which may be required as a result of the confirmatoryanalyses to be completed prior to operation following the first refuelingoutage.

The requested exemption presents no undue risk to the health and safety ofthe public. As discussed in the exemption request, the analysis of thedowncomers is sufficiently rigorous and conservative to permit operation whilethe downcomer design adequacy is reviewed. In its January 31, 1986 letter,the staff found that the downcomer design meets licensing criteria for theupset and emergency conditions, but that design adequacy for the faultedcondition has not been fully demonstrated. The faulted condition requiresconsideration of a simultaneous Loss of Coolant Accident and Safe ShutdownEarthquake with other coincident loads. As discussed in detail in theapplication, the simultaneous occurrence of a Loss of Coolant Accident andSafe Shutdown Earthquake is extremely remote and presents no undue risk forthe period from fuel loading to the first refueling outage. Moreover, thedesign of the reactor coolant pressure boundary and in particular the use ofType 316NG steel for the recirculation system piping makes the probability of

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Mr. Robert BerneroPage 2

a Loss of Coolant Accident at Nine Mile Point Unit 2 extremely small. This isparticularly true for service of such piping during the period of therequested schedular exemption. The exemption application shows that theanalyses and the loads used in the analyses are conservative and that designmargins are present.

You can be assured that Niagara Mohawk is moving forward in a timelymanner to resolve this issue. Ne intend to discuss the Company's approach toresolution periodically with you and your staff. In the meantime, should youhave any questions concerning our program or our schedular exemption, pleaselet me know.

Very truly yours,

C. V. ManganSenior Vice President

Enclosure1325G

xc: R. A. Gramm, NRC Resident InspectorProject File (2)

X"

APPLICATION FOR SCHEDULAR EXEMPTIONRELATED TO FURTHER ANALYSIS OF AND POSSIBLE

MODIFICATION TO CONTAINMENT DOWNCOMERS

I. Introduction and Summar

Nine Mile Point Unit 2 ("NMP Unit 2") is a nuclear

power plant employing a General Electric Company singleJ

cycle, forced circulating boiling water reactor ("BWR") with

a plant rated core thermal power level of 3323 MWt

corresponding to a net electrical output, of 1080 MWe. The

containment design utilizes the BWR Mark II concept of

over-under pressure suppression with multiple downcomerst

connecting the drywell to the pressure suppression chamber.

The primary containment is a .steel lined, reinforced

concrete enclosure housing the reactor and suppression pool.

The downcomers consist of 121 24 inch diameter Schedule 20

pipes open to the drywell and submerged approximately 11

feet below the high water level of the suppression pool,

providing a flow path for uncondensed steam into the pool.,E

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'heNMP Unit 2 downcomers are made of stainless steel, withI

a nominal 0.375 inch wall thickness.

Other Mark II BWRs utilize carbon steel downcomers and

use a support structure to brace the downcomers. Other

applicants have utilized this design because of their need

for licensing approval to be consistent with their earlier

schedule for operation. Niagara Mohawk Power Corporation

("Applicant" or "Niagara Mohawk" ), because of the licensing

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and construction schedule of NMP Unit 2, was able to

incorporate in its design knowledge gained from: further

refinements and definitions of the Mark XX loads approved by

the NRC and thus was able to design its downcomers without,

bracing.

Niagara Mohawk tendered an application for an operating

license for NMP Unit 2 on January 31, 1983. The Final

Safety Analysis Report ("FSAR") was docketed on April 12,

1983 subsequent to the completion of the Staff acceptance

review. At a meeting with the NRC in February 1983, the

matter of the design of the downcomers was noted by the

Applicant as unique to NMP Unit 2.

The design of the containment and downcomers was a

significant operating license review issue. A chronology of

this review effort is contained in Appendix A to this

submittal. Both in the draft Safety Evaluation Report

("SER") for NMP Unit 2 issued in May 1984 and in the SER

(NUREG-1047) issued in February 1985, the matter of the

containment systems and, in particular, the downcomer design

was extensively discussed in Section 6.2 and classified as a

confirmatory item (see also SER Section 1.9 and Table 1.4).

The design of the downcomers . was considered by a

'subcommittee of the Advisory Committee on Reactor: Safeguards

("ACRS") on February 20-21, 1985 and by the full committeeil

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at its March 7-8, 1985 meeting as part of its review of the

operating license application for NMP Unit 2. Although the

matter of containment and downcomer design was discussed

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during the subcommittee and full ACRS meetings (see Tr.

254-60 of the February 21, 1985 subcommittee meeting and Tr.

169-82 of the March 7, 1985 full committee meeting), the

ACRS letter of March 11, 1985 to the Commission 'id not

discuss this matter.1

While further, meetings and exchanges of information

were made on this subject as detailed in Appendix A, it was

not until December 20, 1985 that the Staff met with the

Applicant and expressed concerns regarding the downcomer

evaluation. A further meeting was held on January 15, 1986,

with a submittal containing additional analyses being made

on January 20, 1986. As a result of these series of

meetings and the submital of additional information, on

January 31, 1986, the NRC wrote to the Applicant with a

report of its basic review conclusion. The, Staff found:

Based on our review of your mostrecent submittals, the staff and itsconsultants conclude that the unbraceddowncomer design at NMP-2 is marginal.The staff also concludes that the NMP-2design meets the licensing criteria for

you have not adequately demonstrated thedesign adequacy for the faulted condi-tion. Specifically, the downcomer maylose geometrical stability beforereaching the calculated stress levelsfor the faulted condition. (Emphasis inoriginal.)

The Staff's letter concluded:

To repeat then, the staff concludesnow that you have not adequately demon-strated the design adequacy for thefaulted condition, that is, the condi-tion where the loads of bothLoss-of-Coolant-Accident (LOCA) and Safe

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Shutdown Earthquake (SSE) must beconsidered.

In our recent meetings you have raisedthe question that you might seek moretime after licensing, perhaps to thefirst refueling outage,, to complete yourwork to resolve this issue. You may beable to do so by requesting a schedularexemption from the regulations for thatpurpose. If you choose to do so youshould consider the likelihood of LOCAand SSE loads in your plant consideringthe design and materials of constructionas well as the design factors discussedabove.

In order to resolve the Staff finding, Niagara Mohawk

must either demonstrate the design adequacy of the installed

downcomers for the faulted condition, or modify the design.

Without the requested relief, either of these courses would

prevent fuel loading 'nd ascension to power in a timely

manner. Consequently, as previously discussed with the

Staff, Niagara Mohawk is seeking an exemption while itpursues the first course, i.e., the performance of

additional confirmatory '.analysis to verify„ the 'ability of

the downcomers to perform their intended function over theII I

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design life of the Station. If such further analysis does

not confirm the appropriateness and conservatism of the

original analysis, modification of the downcomers might be

necessary.

Niagara Mohawk therefore requests a schedular exemption

pursuant to the Commission's regulations under 10 C.F.R.

550.12(a) to allow completion of the analysis and any

resulting requirement for modification of the installed

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downcomers in an orderly manner. Specifically, the

requested exemption is to permit operation during the time

that confirmatory analyses of design margins for the Nine

Mile Point Unit 2 downcomers are being performed.

Furthermore, it is requested that the NRC permit any

hardware changes to the facility required as a result of

this confirmatory evaluation to be completed prior to

operation following the first refueling outage.

, The following discussion sections demonstrate that the

grant of an exemption would be in accordance with the

requirements of 10 C.F.R. $ 50.12(a). In particular, as

discussed in detail below, grant of the exemption would not

present an undue risk to public health and safety. The

Staff has found that the design meets licensing criteria for

the upset and emergency conditions, but that the designt

adequacy for the faulted condition has not been

demonstrated. The faulted condition anticipates the

simultaneous consideration of LOCA and SSE with other

coincident loads. As discussed below, for NMP Unit 2, the

simultaneous occurrence of :a LOCA and SSE is extremely

remote and for the .period in question presents no undue riskto the public health and,, safety. Moreover, because of the

design and materials used in the reactor coolant pressure

boundary, ,the probability of a LOCA at NMP Unit 2,~ l'

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particularly until the first refueling outage, is extremely

small. In addition, there are conservatisms in the

assumptions, analysis and assumed loads which provide margin

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not previously taken into account in determining design

adequacy for the faulted and other conditions.

The requested exemption also meets the other

requirements of ,the regulations.'pecial circumstances

exist which meet the standard, under 10 C.F.R. 550.12(a) (2) .C

The requested exemption is authorized by applicable law,4

including the Atomic Energy Act, 42 U.S.C. 52011 et sec[.,

and National Environmental Policy Act, 42 U.S.C. 54321 et

~se ., and is consistent with the common defense and

security.II. The Requested Exemption Does Not

Present An Undue Risk To ThePublic Health And Safet

As discussed below, the analysis of the unbraced

downcomers is sufficiently rigorous and conservative to

permit operation while the downcomer design margins are

being confirmed. It is the Applicant's position that, as

constructed, the downcomers satisfy all NRC requirements and

conform to all governing codes and standards.

The sufficiency of the downcomers is demonstrated by

the detailed analyses contained in the FSAR (see Appendix

6A, Design Assessment Report for Hydrodynamic Loads), the

references contained therein and the further analyses and

supporting calculations which have been submitted or

referenced in the docket, including those noted in Appendix

A. Those evaluations and analyses are incorporated herein

by reference. In addition, there is margin built into the

design limits and procedures contained in the ASME code used

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to evaluate functional capability such that, no undue risk to

the health and safety of the public exists. Applicant is

confident that the further evaluation will confirm the

margin present in the design. Moreover, there is margin not,

previously recognized in the analysis implicit in the

evaluation because of the requir'ements and limits the NRC

imposes, both in the load combinations and in the individual

types of load definitions specified.

The NRC Piping Review Committee, which was established

to carry out a comprehensive review of NRC regulatory

requirements for nuclear power plant piping, states at page

2 of NUREG-1061, Vol. 5, "Report of the U.S. Nuclear

Regulatory Commission Piping Review Committee, Summary

Piping Review Committee Conclusions and Recommendations,"

that "certain load combinations, particularly the

loss-of-coolant accident (LOCA) plus safe shutdown

earthquake (SSE) load combination, represent, severe design

requirements . . . No analytical or physical evidence

supports a causal relationship b'etween', pipe break and

earthquake . . . ." Later, in Section 7.2.1 Event,

Combinations, the authors state, "ft]here has never been a

well-developed rational basis for considering concurrent

earthquake and large loss-of-coolant accidents (LOCA) loads

in the design basis."

Decoupling the seismic loads from the LOCA loads forthe faulted condition significantly increases the margin

between the resulting total combined stress and the

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functional capability allowable stress. If the SSE load

were eliminated from consideration in the faulted condition

for NMP Unit 2, any concern as to the adequacy of design*

margins would be eliminated.

Moreover, the probability of a large break LOCA is now

considered to be significantly lower than previously

believed. This is particularly true for NMP Unit 2 because

of the materials used for construction of the recirculation

system. The leak before break (LBB) concept has led the

Staff to initiate rulemaking to modify GDC-4, excluding the

double ended guillotine break (DEGB) from the set of design

basis accidents. It is Applicant's understanding that a new

rule for pressurized water reactors ("PWRs") is expected

shortly, and for BWRs a similar rule is scheduled for

issuance thereafter, perhaps as early as this year.

Inasmuch as BWRs potentially have greater susceptibility to

intergranular stress corrosion cracking (IGSCC), the Piping

Review Committee recommended in NUREG-1061 (recommendation

A-4, p. xi, Volume 5) that the recirculation piping in BWRs

be replaced with alloys resistant to IGSCC, for example,

with Type 316NG to reduce the probability of a DEGB.

However, recirculation piping for NMP Unit 2 isV',constructed of Type 316NG.

already

In this regard,, ongoing studies at the . Lawrence

Livermore, National Laboratory ("LLNL") should be noted.

LLNL has conducted independent confirmatory research under

its Load Combination Program to provide the NRC with a

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technical basis for reevaluating the DEGB design require-

ment. Summary reports have been written on the probability

of pipe failure in the reactor coolant loops of Westinghouse

(NUREG/CR-3660, Vol. 1) and Combustion - Engineering

(NUREG/CR-3663, Vol. 1) PWR plants, and similar work is in

progress for BWRs. Although not directly applicable to

BWRs, the PWR results are indicative of trends to be

expected for BWRs.

* For Westinghouse plants, the median probability of a

direct DEGB in reactor coolant loop piping is about 4.4

x 10 events per plant year; the corresponding median~ 7probability of leak (through-wall crack) is 1.1 x 10

(NUREG/CR-3660, Vol 1, p. 67)

* For Combustion Engineering plants, the best estimate

probability of a direct DEGB in reactor coolant loop-14 -13

piping ranges from 5.5 x 10 to 4.5 x 10 events

per plant year; the corresponding probability of a leak-8(through-wall crack) ranges from 1.5 x 10 to 2.3 x

10 . (NUREG/CR-3663, Vol 1, p. 51)

* For both types of plants, the results indicate that,"direct DEGB and a safe shutdown earthquake can be

considered independent random events whose probability

of simultaneous occurrence during plant life is negli-

gibly low." (NUREG/CR-3660, Vol 1, p. 74)

A paper entitled "Pipe Ruptures in BWR Plants" by the

authors of the two PWR reports cited above was presented at

the 13th NRC Water Reactor Safety Research Information

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meeting in October 1985. It supports the conclusion that

the results discussed above apply to NMP Unit 2. As stated

in its abstract, the report:

[D]escribes a probabilistic evaluationof recirculation, main steam andfeedwater piping in BWR plants. As inthe earlier PWR evaluations, two causesof pipe break are considered: pipefracture due to the growth of cracks atwelded joints ("direct" DEGB) and piperupture caused by theseismically-induced failure of heavycomponent supports ("indirect" DEGB).The probability of direct DEGB wasestimated using a probabilistic fracturemechanics model. The probability ofindirect DEGB was estimated byconvolving seismic hazard and heavycomponent support fragility. Twoadditional factors not applicable to PWRreactor coolant loop pipingintergranular stress corrosion crackingand pipe support fragility — wereconsidered in the BWR study. Theresults of this study indicate that theprobability of DEGB is very low for allthree piping systems, except forrecirculation piping when IGSCC is afactor,, in which case IGSCC dominatesthe probability, of

failure.'pecifically,it was'eported that for a BWR piping

system (recirculation, main', steam and, feedwater)'n the

absence of IGSCC", the probability of a direct DEGB, at the-13 -12

50% confidence limit, ranges from 1.8 x 10 to 3.8 x 10

per year. The corresponding probability of a leak ranges

from 8.8 x 10 to 1 x 10 . As was the case for PWR

piping, it was found that "earthquakes contribute only

negligibly to the probability of direct DEGB."

The final sentence of the report should be emphasized:

"This result [the low probability of a DEGB] implies that if

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IGSCC can be satisfactorily mitigated (for example through

the use of XGSCC-resistant materials), then reactor coolant

piping DEGB could be eliminated as a design basis for BWR

plants as is currently being done through rulemaking action

related to PWR reactor coolant loop piping."

As stated earlier, the NMP Unit 2 recirculation piping

is fabricated from XGSCC-resistant Type 316NG. Additional

steps have been taken, as stated in FSAR Section 5.2.3.4.1,

to avoid stress corrosion cracking. With minor exceptions,

all other wrought austenitic stainless steel in the reactor

coolant pressure boundary is IGSCC resistant low carbon Type

304L or Type 316L. All balance of plant ASME IXI Class 1, 2

and 3 piping and components comply with NUREG-0313, Rev. 1,

with exceptions for service below 200'F, or above 200'F for,

insignificant periods. The exemption is being sought solely

for the first refueling cycle. Experience at BWRs indicates

that. IGSCC is a time-dependent phenomenon and the

probability of failure of reactor coolant pressure boundary

piping during the first refueling cycle is negligible.

Because of the XGSCC mitigating measures taken at, NMP

Unit 2, the occurrence of a LOCA in combination with an SSE

may be ruled., out as exceedingly, low in probability. Based

on this information, at least for the 'nterim, the

coincidence 'of these two

the faulted condition.

events need not be considered forI'oreover,the probability of

'ccurrenceof a large LOCA at NMP Unit 2 is itselfexceedingly small.

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There are three basic types of loads considered in the

analysis of the downcomers; those due to earthquakes (seis-

mic) which have already been considered above, those due toPl

actuation of the safety relief valves (SRV), and those due

to LOCAs (condensation'oscillation or chugging). AddressedI

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below are the margins implicit in the NRC requirements for

the SRV and LOCA loads.

The load requirements for SRV actuation are stated in

NUREG-0802, "Safety/Relief Valve Quencher Loads: Evaluation

for BWR Mark II and III Containments." Two methodologies

are presented there. One is based on tests performed at

Kraftwerk Union's Karlstein testing facility, and an

alternate, used for NMP Unit 2, is based on a series of 200

tests run at a Kraftwerk Union BWR plant. The pressure

traces from three of these 200 tests were chosen for use in

the load specification for NMP Unit 2 because of theirrelatively strong pressure amplitude of 0.5 to 0.8 bars and

because of their differences in frequency content and

damping characteristics. The 0.8 bar peak pressure found in

one of the selected traces was the highest pressure

amplitude ever measured by KWU during in-plant tests of,

relief systems equipped with KWU quenchers. The NMP Unit 2

SRV load specification is based on a conservative amplitude

multiplier of 1.5 and a conservative frequency multiplier

(up to 1.8) to accentuate lower frequencies.

These two factors are the major contributors to the

conservatism built into the NMP Unit 2 load specification.

The 1.5 pressure amplitude multiplier has been shown to be

bounding in the Karlstein tests. The 1.8 frequency factor

artificially introduces substantial power at the low

frequency end of the spectrum that was not observed in the

Karlstein tests. Other conservatisms exist as the result of

the greater pool surface per quencher as compared to the

Karlstein tests, the use of the 1.5 factor on the

differential pressure across the downcomer, and the

application of the peak differential pressure across the

downcomer to the downcomer projected area.

The load requirements for LOCA loads are presented in

NUREG-0808. The loads were largely based on the GE 4TCO

tests, with confirmation from JAERI tests. Based on a

reading of NUREG-0808, sev'eral areas of substantial

conservatism exist in the condensation oscillation (CO)

loads definition which are identified as follows:

"The staff 'inds that those tests provide anI H

appropriate data base for a conservative load speci-

fication for,CO.'" (p. 2-14)

"The results (regarding the 4TCO pressure distribution]are conservative because each pressure-time history is

applied independently and directly to the structure.

It should be noted that the envelope does not represent

a PSD [power spectral density] of a time series of a

hypothetical LOCA; rather, for any LOCA which isdetermined by some initial conditions, the PSD of its

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pressure-history will be bounded by this envelope."

(p. 2-18)

* "The comparison [between the 4TCO and JAERI tests]

shows that the 4TCO PSDs are significantly higher than

the PSDs for the comparable JAERI test at allfrequencies . . . . Thus, additional conservatism is

available with the present load definition when

compared to JAERI." (p. 2-18)

* "Preliminary observations of the JAERI full-scale Mark

II tests [discussing the conservatism due to

experimental data from a single vent facility as

compared with the multivent Mark II plant] indicate a

multivent factor of approximately two." (p. 2-19)

* "In the current methodology, the bounding time series

is selected at each frequency, and the load is applied

directly to the structural models. No credit is taken

for either amplitude averaging or phasing; hence the

conservatism results." (p. 2-19)

, NUREG-0808 also comments on the conservatism inherent

in the chugging load requirement,:

* "Past observations have indicated that the highest

amplitude chugs'ccur when the vent-mass flow is

relatively high and the steam n'early air 'free. Because

these are the, conditions,„toward which the 4TCO tests

are biased,'he staff concludes that the chugs observed

in this test series provide a conservative data base."

(p. 2-24)

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"The staff agrees with the owners that the seven key

chugs selected for design source derivation represent a

conservative subset of chugs from the 4TCO data. The

staff has concluded that the way sources are inferred

from the, key chugs is conservative with respect to

amplitude and power because the msp [mean square power]

of the source parameters is made to match or exceed then

msp of the measured chug pressures for all cases except

for the higher of the two sources used for design

source 807, which achieved 95%." (p. 2-25)

"The staff concludes that the owners'ethod of

proceeding (averaging amplitudes with the largest chug

adjacent to the key chug) is acceptable and provides a

conservative method of accounting for this amplitude

variation in deriving the design sources." (p. 2-26)

"In other words, confining the chug-start times from

100 vents to lie within a 50-ms time window which was

observed from data of only five vents is conservative.

An additional conservatism is the use of 1000 Monte

Carlo trials to select the set of start times with the

smallest variance for plant application. In light of

these various conservatisms, the staff finds acceptable

the methodology for vent desynchronization as detailed

. in Reference 26." (p. 2-31)

"While some arguments could be entered into whether the

above method of comparison [with JAERI data] is the

best, one, the staff feels it is an acceptable one and.

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is important in confirming the conservative nature of

the Mark II owners'eneric improved chugging

specification." (p. 2-32)

To summarize, the probability of the simultaneous

occurrence of the SSE and LOCA is so low as not to require

consideration for the faulted condition. Moreover, the

probability of a LOCA is itself exceedingly small. Finally,

there are conservatisms inherent in the design analysis and

load specifications. Thus, the requested exemption presents

no undue risk to the public health and safety.

III. Special Circumstances Exist WhichWarrant Issuance Of The RequestedExem tion.

Special circumstances exist under the categories

contained in 10 C.F.R. 550.12(a)(2) any of which would

warrant issuance of the requested exemption. Undue hardship

and costs would otherwise result that are significantly in

excess of those incurred by other licensees. Further, the

exemption is temporary and Niagara Mohawk has made good

faith efforts to comply with licensing requirements. These

special circumstances are discussed in accordance with the

classification contained in the rule.A. Undue Hardshi

(iii) Compliance would result inundue hardship or other costs that aresignificantly in excess of those contem-plated when the regulation was adopted,or that are significantly in excess ofthose incurred by others similarlysituated

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Completion of the additional analyses required by the

NRC, as well as any necessary hardware changes, prior to

operation of the facility would result in undue hardship and

costs. Any physical changes to the facility prior to

operation would significantly increase costs in terms of

redesign, procurement, fabrication, installation and testing

and would result in additional costs associated with delay

in startup of the facility.If bracing systems comparable to those installed at

plants having similar, but not identical, downcomers were

required for NMP Unit 2, the cost of redesign, procurementf

fabrication, installation and testing on an accelerated

basis would be an estimated $ 10 million.A bracing system similar to those on previously li-

censed BWRs would have to be installed with the water

removed from the wet well. Inasmuch as this would essen-

tially preclude parallel integrated system testing which

could not proceed within containment, the delay would add

6-10 months to the construction schedule'. Considering allf

factors, this would result in' monthly delay cost of

$ 60,000,000. The $ 60 million/month has two components.

Fifteen million dollars is estimated as additional overhead

construction costs, that is, the overhead involved in

maintaining the construction status at the site. The

remaining forty-five million dollars constitutes financing

costs. This $ 60 million cost does not include the cost of

replacement power. Postponing modifications until the first

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refueling outage would permit schedule optimization and

allow some of the'equired work to proceed in parallel with

other work to be performed at that time. Thus, to require

modifications prior to fuel loading would result in costs of

$ 370 to $ 610 million attributable to the installation and

delay.

These costs would be significantly in excess of those

incurred by others similarly situated, i.e., by other Mark

II plants as a whole, inasmuch as the timing of the Staff

review and construction allowed modifications proposed by

the utilities to be phased in during the construction phase

so as not to interfere with preoperational testing and so as

not to delay licensing.

Thus, special circumstances exist which warrant grant-

ing of the exemption.

B. A licant,'s Good Faith Efforts

(v) The exemption would provide onlytemporary relief from the applicableregulation and the licensee or applicanthas made good faith efforts to complywith the regulation

As noted above, the exemption is being requested to

provide temporary relief until, at the latest, the end of

the first refueling outage. As further discussed, Applicant

has made good faith efforts to comply with all regulatory

requirements as set forth in applicable Staff guidance.

Applicant identified this design feature in the Final Safety

Analysis Report and in early meetings with the Staff as

unique to this facility. Information requested by the Staff

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was provided by the Applicant in a timely manner. At the

time of the issuance of the SER, the NRC classified the

information to be supplied regarding the downcomer as

"confirmatory," meaning that it had been "essentially been

resolved to the staff's satisfaction, but for which certain

confirmatory information has not been provided by the appli-

cant" (SER at 1-9).

After subsequent meetings to define the information

necessary for the Staff to close out the issue, the

additional information was promptly submitted. As

demonstrated above, the present design presents no undue

risk to the public health and safety in the interim. Thus,

under this criterion, good cause has been shown for granting

the requested exemption.

C. Other Material Circumstances

(vi) There is present any othermaterial circumstance not consideredwhen the regulation was adopted forwhich it would be in the public interestto grant an exemption. If such condi-tion " is relied on exclusively forsatisfying paragraph (a)(2) of thissection, the exemption may not begranted until the Executive Director forOperations has consulted with theCommission.

While not being relied upon exclusively (and thus not

requiring consultation with the Commission), there are other

material circumstances which support issuance of the ex-

emption. As discussed above, ongoing developments in NRC

research and regulatory policy may significantly affect

present downcomer design requirements. The application of

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this research to BWR technology may well occur prior to the

first refueling outage for NMP Unit 2 and could obviate the

need to redesign downcomer bracing for a hypothetical

double-ended guillotine rupture of the recirculation lines.

A rulemaking proceeding on PWR piping is in progress; the

rule is expected to be issued in the near future. It has

even been suggested that components the NRC has required for

in PWRs actually may decrease the reliability of those

plants (NUREG-1061, Vol 5, p. 13) . A rule for BWRs may be

issued as early as 1986. Research to support such a rule is

in progress. The NRC's consultant has already stated that

the research is applicable to BWR Mark II facilities. It is

prudent to await the rulemaking outcome before adding

structures that may be found to be unnecessary or to detract

from safety.This ongoing research and its results were not contem-

plated at the time the NRC promulgated the General Design

Criteria controlling the design of the NMP Unit 2'I

containment. It makes no sense to require the addition off

bracing that would be unneeded under, the new criteria now

being considered for possible application to BWRs. Inasmuch

as all other BWR Mark IIs have installed bracing, NMP Unit 2

is seemingly unique in being able to take advantage of this

developing research and new rulemaking.

For these reasons, special circumstances are present

supporting the issuance of the requested exemption.

21

IV. The Other Requirements For IssuanceOf An Exem tion Are Met.

The requested exemption is authorized by applicable

law, including the Atomic Energy Act and National

Environmental Policy Act. With regard to the "common

defense and security" standard, the grant of the requested

exemption is consistent with the common defense and security

of the United States. The Commission's Statements of

Consideration in support of the exemption rule note with

approval the explanation of this standard as set forth in

Lon Island Li htin Com an (Shoreham Nuclear Power

Station, Unit 1), LBP-84-45, 20 NRC 1343, 1400 (1984).

Thus, the term "common defense and security" refers

principally to the safeguarding of special nuclear material,

the absence of foreign control over the applicant, the

protection of Restricted Data, and the availability of

special nuclear material for defense needs. The granting of

the requested exemption will not affect any of these

concerns and is therefore consistent with the common defense

and security.The proposed exemption has been analyzed and determined

not to involve additional construction or operational activ-

ities which may significantly affect the environment. Itwill not result in a significant increase in any adverse

environmental impact previously evaluated in the Final

Environmental Impact Statement-Operating License Stage, a

significant change in effluents or power levels or a matter

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not previously reviewed by the NRC which may have a signifi-cant adverse environmental impact.

V. Conclusion

For the foregoing reasons, Niagara Mohawk Power Corpo-

ration has demonstrated, that it meets the requirements

contained in 10 C.F.R. 550.12(a) for the issuance of an

exemption. Therefore, the requested exemption to permit

operation of Nine Mile Point Unit 2 during the time that

confirmatory analyses to verify the ability of the

downcomers to perform their intended function over the

design life of the station are being performed, and to allow

any changes to the facility required as a result of the

confirmatory evaluation to be completed prior to operation

following the first refueling outage should be granted.

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APPENDIX

HISTORY OF NRCDOWNCOMER REVIEW

January 31, 1983

February 1983

FSAR tendered

First Case Load Forecast visitUnbraced downcomers specificallyidentified to NRC as a unique reviewitem.

April 12, 1983

December 1983

May 1984

FSAR docketed.

Structural Engineering Branch Audit

Draft, SER issued. Downcomers identifiedas a confirmatory issue.

June 1984 Mechanical Equipment Branch AuditsMEB audit specifically discusseddowncomers

February 1985

February/March1985

SER issued listing downcomersas a confirmatory issue with noindication of problem with NRC lookingfor additional details of analysis.

ACRS meetings. NMPC made specificpresentations covering unbraceddowncomers. NRC Staff indicated it wasconfirmatory issue.

June 25, 1985 NMPC met with NRC Technical Reviewers todiscuss downcomer information submittedprior to meeting.

July 1985 Followup meeting with reviewers todiscuss downcomers.

September 30, 1985 Formal submittal of downcomerinformation.

October 18, 1985 Conference call with reviewer whoindicates he needs outside consultant toassist. with review.

October 29, 1985 Conference call with NRC providingadditional information.

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November 18, 1985 Schedule meeting with NRC management;downcomers identified as an issue thatneeded specific attention.

November 25, 1985 Conference call with NRC.

November 26, 1985 Downcomer calculations submittedinformally.

December 6, 1985 Meeting with R. Bernero at NMP-2regarding need to resolve issue.,

December 16, 1985

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December 20, 1985

Phone calls from NRC requesting NMPCsenior management at 12/20 meetingbecause of problems with downcomers.

Meeting with NRC Technical StaffFirst. indication that downcomer designmay not be acceptable.

December 31, 1985 Formal submission of downcomercalculations.

January 3, 1986 Phone call with NRC with reservationsindicated.

January 8, 1986

January 15, 1986

Letter from NRC with notes of 12/20meeting and Draft, SER for downcomer.

t

Meeting with NRC regarding revisedapproach.

January 20, 1986 Revised calculations informallyprovided.

January 23, 1986 -Letter transmitting revisedcalculations.

January 24, 1986 Letter transmitting report byApplicant's consultant StevensonAssociates.

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UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSION

In the Matter of

Niagara Mohawk Power Corporation

(Nine Mile Point Unit 2)

Docket No. 50-410

AFFIDAVIT

C. V. Mangan, being duly sworn, states that he is Senior Vice President ofNiagara Mohawk Power Corporation; that he is authorized on the part of saidCorporation to sign and file with the Nuclear Regulatory Commission the"Application For Schedular Exemption Related To Further Analysis Of AndPossible Modification To Containment Downcomers"; and that such document istrue and correct to the best of *his knowledge, information and belief.

Subscribed and sworn to before me, a Lotary Public in and for the State ofNew York and County of Onondaga, this ~/ — day of February, 1986.

Notary Public in and for OnondagaCounty, New York

My Commission expires:CNISTWE AUSI1N

N~a Pubhc in the Stateot Ncw YQu&fg!nQnondaga Co. NL4N+

grab 30, LR

MORA3HNESH)EN e:8 to olde ett ni iHdu'I ertqHKNQT5&4I sgsbnwO al h'dhu0Wl+ AM!4as~ a."meno3 yfP.