Text

Forwards Responses to SER Outstanding Issues 76,3,4,49 & Question 252.1 of SER Outstanding Issues 50 & 51
	ML20084E892
Person / Time
Site:	Beaver Valley
Issue date:	04/30/1984
From:	Woolever E; DUQUESNE LIGHT CO.
To:	Knighton G; Office of Nuclear Reactor Regulation
References
	2NRC-4-049, 2NRC-4-49, NUDOCS 8405030042
	Download: ML20084E892 (31)
	v • d • e

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2NRC-4-049 (412) 787-5141 (412)923 - 1960 Telecopy (412) 787-2629 Nuclear Construction Division April 30, 1984 Robinson Plaza, Building 2, Suite 210 Pittsburgh, PA 15205 United States Nuclear Regulatory Cottmission Washington, DC 20555 ATTENTION:

Mr. George W. Knighton, Chie f Licensing Branch 3 Of fice of Nuclear Reactor Regulation

SUBJECT:

Beaver Valley Power Station - Unit No. 2 Do cke t No. 50-412 Open Item / Question Response Gentlemen:

This letter forwards responses to the issues listed below.

Duque sne Light Company plans to incorporate the res pons es to the FSAR questions into FSAR Amendment 7.

The following items are attached:

At tachment 1:

Res pons e to Out s t anding Issue 76 of the Beaver Valley Power Station Unit No. 2 Draf t Safety Evaluation Report. :

Response to Out s tanding Issue 3 of the Beaver Valley Power Station Unit No. 2 Draf t Safety Evaluation Report. :. Response to Outstanding Issue 4 of the Beaver Valley Power Station Unit No. 2 Draf t safety Evaluation Report.

Res po nse to Out standing Is sue 49 of the Beaver Valley Power Station Unit No. 2 Draf t safety Evaluation Report.

At tachment - 5 : Res ponse to Question 252.1 (Out s tanding Issues 50 and 51 of the Beaver Valley Power Station Unit No. 2 Draf t Safety Evalu-ation Report.

DUQUESNE LIGHT COMPANY y.u%G C roe

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O E. J. Woolever

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Vice President D -

SUBSCRIBED AND SWO TO BEFORE ME.THIS

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Notacy Public 6)f -'

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T ANITA ELAINE REITER, NOTARY PUBUC li#K.;y,;g,ddry nA

- ROBINSON TOWNSHIP, ALLEGHENY COUNTY l:

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MY COMMISSION EXPIRES OCT0EER 20,1986 '

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United Stctes Nuclear Rsgulcto'ry Conuniacion Mr. Gaorge W. Knighton, Chief

.Page 2 KAT/wjs At tachment s cc:

Mr. H. R. Denton, Director NRR (w/a)

.Mr. D. Eisenhut, Director Division of Licensing (w/a)

Mr. G. Walton, NRC Resident Inspector (w/a)

Mr.

M.' Lacitra, Project Manager (w/a) s v

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United Stcteo Nuclear Regulatory Commission Mr. George W. Knighton, Chief Page 3 COMMONWEALTH OF PENNSYLVANIA )

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COUNTY OF ALLEGHENY

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On this d(M.

day of

/9//, before me, a Notary Public in and for said Commonwealth and County, personally appeared R. J.

Washabaugh who being duly sworn deposed and said that (1) he is duly authorized to execute and file the foregoing Submittal on behalf of E. J.

Woolever, Vice President of Duquesne Light, (2) he is duly authorized to execute and file the foregoing Submittal on behalf of said Company, and (3) the statements set forth in the Submittal are true and correct to the best of his knowledge, s

Notary Public ANITA EUjNE REITER, NOTARY PUBLIC ROBINSON TOWNSHIP, ALLEGHENY COUNTY MY COMMISSION EXPIRES OCTOBER 20,1986

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ATTACHMENT 1 Posponse to Outstanding Issue 76 of the beaver Valley Power Station Unit No. 2 Draf t Safety Evaluation Report Draf t SER Sect ion 7.7.2.2 : High Energy Line Breaks and Consequential Control System Failures a

IE Information Notice 79-22, issued Spetember 19, 1979, raises a concern that ce rt ain nonsafety grade or control equipment, if subj ect ed to the adverse en r ironme nt of a high-energy line break, could malfunct ion and c au se plant conditions to be more severe than those analyzed in FS AR chapter 15.

The applicant was requested to perform a review to determine what, if any, design ch anges or ope rato r act ions would be neces sary to ens ure that high-energy line breaks will not cause control sys tem failures to complicate the event beyond the Chapter 15 analyses.

The staf f has reviewed the applicant's response, contained in FSAR Amend-ment 4, and finds it needs further clarification in the following areas:

1.

PT 444 and 445, used for the pres surizer PORV control ar e not quali-fied.

The ap plicant's response. indicates all equipment as so ci at ed with this control system is Category I.

2.

The intent of NRC Ques tion.420.4 - was to require the ap plicant to review all possible control system malfunctions due to a high-energy line break inside o'r out s ide of cont ainment.

It appe ars that the applicant only. reviewed the four scenarios described in IE Informa-

- tion Notice 79-22 and limited that review to inside containment.

This item is open pending st'af f review of further clarification from the applicant.

Response

The' response to these issues has been incorporated into -the revised response

. to FSAR Question 420.4.

The revised ' response will be incorporated into-the next FSAR amendment and is included below:

Scope:

.On September 18,1979, Westinghouse presented to ; the Staf f a summary of the investigation that had been conducted which' led to' the identification of fo ur ' potential interaction scen'arios'. where ' the ef fect. of adverse -

environment s, resulting from high energy' line. breaks',' on - control sys tems could ~ lead toi consequeaces more limiting than the resuit a presented in

- the Final. Safety Analysis. Report (FSAR).

The four - pot ent ial. interact ion scenarios are:

1. - ' Steam generator power-operated relief valve control system,

2. ' Pressurizer. power-operated relief valve control sys tem,-

3.

Main feedwater control system, and _

4.. Automatic rod ; control syatem..

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Was t inghoure has not since ident ified any ot he r pot ent ial int eract ion scenarios with similar consequences.

DIE has rev i ewed the specific non-safety grade sys tems listed in IE Information Not ice 79-2 2 fo r pot ent ial int eract ions that could constitute a substantial safety hazard.

DIE has not been able to identify such an interaction. While variations from the FSAR licensing bases have be en ide nt ified, the basic conclusion of the FSAR (that these events do not constitute an undue risk to the health and safety of the public) remains untouch ed.

DIE has also not be en ab le to identify in its review of IE Information Notice 79-22 any other nonsafety grade equipment whose pe rfo rmance, when subj ect ed to an adverse environ-me nt, could impact the prot ect ive funct io ns pe rformed by safety grade equipment.

Implicit in the four pot ent ial int e ract io n sc enarios ident ified by Wes t inghou se are wors t case as sumpt ions conce rning the break s ize and loca t io n, and the type and extent of consequential failures in control systems induced by the adverse environnent. These assumptions are there-fore in addition to the already conservative set of as sumpt io ns asc ribed to the analysis of the des ign basis event s report ed in the FSAR.

It follows that these scenarios repr es ent a significantly le s s pr obab le subset of the design basis events that are dependent on the occurrence of additional event s, each having an as so ciatd uncert ainty of occur ring.

While no quantitative analysis has been conducted concerning the impr ot>-

ability of ove rall sc enarios, the following define, fo r two of the

- scenarios ident ified ab ove, the cons e rv at ive as sumpt ions already con-t ained in the design basis event analys is repo rt ed in the FSAR and the additional cons e rvat ive as sumpt ions to be made to de rive the pos tulated interaction scenario.

With regard to the probability of any single design basis event init i-

. ating, via the adverse environnent, failures in several control systems, it again can be noted from the following information that the probability of all the additional set of conditions occurring simultaneously for more than one scenario is of an even lower order of magnitude than for each individual scenario.

Furthermore, iliplementation of the proposed so lu-t ions fo r the individual scenarios will, as a cons eque nce, addres s any concern for multiple int eract ion from a single initiating Design Basis Event.

Due to the implement at ion in the de s ign of the electrical separat ion requirements between control and protection sys tems specified in IEEE-279, the only interaction mechanisms ident ified in the above sc enarios result from conservatively. as suming an adverse env ironnent at'the loca-tion of the control sys tems and the cons eque nt ial equipment failur e in the worst direct ion.

As a conseque nce, it can be ant icipa ted that any interaction scenarios yet to be ide nt ified, in as' yet unreviewed control systems, will be no more probable than the particular scenarios described' by Westinghouse.

1.

Steam Generator Power-Operated Relief Valve Control System IE No t ic e - 7 9-2 2 is not applicable to the BVPS-2 's t ean generato r pc.

Operated relief valves (PORV's_) 2 SVS*PCV101 A, 101B, 101C, and.

2 SV S*llCV104.

These -valves are Category I motor-operated modulating valves (refer to Table 3.11-1).

All equipment in.the' control' system-is Category I.

Power to the valves is from Class IE power sources.

Accident analysis. for this system.is provided in Section 15.1.4.

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2.

Pressurizer Powir-Opsrated Relief Valve Control System Summary of Postulated Scenario s

Following a feedline rupture inside containment, the pres surizer PORV control system malfunctions in such a manner that the power operated relief valves fail in the open pos it ion.

Thus, in ad di t ion to a feedline rupture between the steam generagor nozzle and the cont ain-ment penetration, a breach of the reactor coolant system boundary has occurred in the pressurizer vapor space.

Probability Assumptions Affecting Event Probability and Consequences a.

St andard Safety Analysis Repo rt Assumpt ions Concerning Feedline Break i

1) conservative ir.itial assumptions a). Appendix K decay heat model b) Engineered safeguards power plus calorimetric error c) Programmed RCS t enpe rature plus control de ad band and instrument errors d) initial conservative S/G inventory e) conservative core physics

2) conservative accident assumptions a) break (all sizes) in Safety Class 2 feedline piping 4

b) maximum adve rse 'env ironment al errors for prot ect ive instrumentation c) wors t single act ive failure (loss of any one - auxiliary-feed pump) d) operator action time b.

Additional Assumptions Required for this Scenario

1) Break ' must occur ins ide the containment ' between the steam generator nozzle and the containment penetrat ion. A break at other locations' invalidates this scenario.

2) Double ' ended break leads t'o limiting consequences.

Smalle r breaks permit longer operator action times. '

3): Adverse environment. result ing ' from the break can impact. the pressurizer power operated relief valve control system.

The PORV's'.are air. to open valve 'with.' the f air - supply be ing -

controlled by ' solenoid ' operated valves. - Thel PORV's ani SOV's are.~in cont ainment. - The PORV's are designed to - fail closed on loss of air and power. 'In the plant systen,. prior to each L

of' the.three (3) PORV's ' are three (3) motor operated valves..

The MOV's.and : associated. control systems are qualified Efor:

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postiaccident environmental conditions.

~ 4) : Due : to the adverse - Tenvironment, the pressurizer-PORV control-

.. sys tem ' -initiates. a :. spurious signal to - ope n the PORV's.

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Should the control system continue to operate within spcifi-cation or initate a spurious signal to close the PORV's, the scenario is invalidated.

5) Should the PORV's fail to the safe pos ition (i.e.. closed),

the scenario is invalidated.

Accident Consequences Sect ion 4.2 of WCAP-9600, "Repo rt on Small Break Ac cide nt s fo r Westinghouse NSSS Systems," describes transient analyses fo r a pos tu-lated loss of all main and auxiliary fe edwa ter (no pipe ru ptur e).

The res ult s indicate that, in the event the operator cannot res to r e auxiliary feedwater to the steam generator, the operator is requirai to open the pressurizer PORV 's within 2,500 second s to maint ain adequate core coolant inventory.

The int eract io n scenario pos tulated ab ove is similar to that pr e-sented in Section 4.2 of WCAP-9600.

The additional as sumptions made are the following:

a.

A feedline rupture is assumed to occur between the s t eam ge ner-ator nozzle and the containment penetrat ion, b.

Auxiliary fe ed water is inj ect ed into the int act stean generator following the feedlin:' rupture.

Conservatively assuming that all liquid inventory in the steam gener-ator as so ciated with the ruptured feed line is lor, t via the ruptur e without removing any heat ( i.e., liquid blowdown), the loss of heat sink due to the liquid inventory blowdown of the ruptur ed stean ge nerato r is more than counterbalanced by the auxiliary fe edwater being inj ect ed into the int set steam ge nerato rs following reacto r trip.

Therefore, the results of the analysis pr es ent in WCAP-9600, ect ion 4.2, which illustrates that the ope rato r is not required to take corrective action for at least 2,500 seconds following the loss of feedwate r also applies to this sc enario.

No Safety Analysis Reports assume gre ater than 30 minute ope rato r act ion following a feedline rupture.

Response

FSAR 2.4.11.1 is being revised as shown on the following pages.

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2.4.11 Low Water Considerations

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2.4.11.1 Low Flow in Streams The New Cumberland Lock and Dam maintains the New Cumberland Pool at el 664.5 feet.

Records indicate that this elevation can be maintained at flows up to 20,000 cfs as shown on Figure 2.4-15.

A low-flow frequency curve for the Ohio River at Shippingport is shown on Figure 2.4-16.

This curve represents the lowest continuous 7-day mean flows that would occur. It is based on a statistical analysis of historical flows for the past 44 years (1929-1973) modified by the present reservoir system. (U.S.

Army Corps of Engineers, Pittsburgh District:1970).

An. instantaneous flow could be lower, but with the large impoundments behind the storage dams, the 7-day flow could be provided continuously by temporarily drawing on the river storage when needed.

y Computerized models developed ' by ' the U.S. Army Corps of Engineers were used to simulate regulated stream flows in the Ohio River.

Results of the analysis show that a minimum flow of 4,000 cfs would have occurred at the site during the record drought of 1930 with the contemporary reservoir system.

JL complete failure of the. nearest downstream dam (the New Cumberland Dam) during minimum flow would result in a minimum water surface elevation at the site of 648.6 feet msl (U.S. Army Corps of Engineers, Pittsburgh District 1969, 1973).

This is discussed in more detail in Section 2.3.4 of the BVPS-2 PSAR.

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The USNRC, in its review lof the BVPS-2 PSAR, indicated that, by extrapolating.an unregulated low-flow frequency for drought conditions which may be characterized as,the: most severe reasonably possible at the plant site, an instantaneous low flow of 800 cfs

.INbfdI O could occur.

A Technical Specification, as described in Section 2.4.14, will be

.I established for low-flow / low level in the Ohio River so that safe shutdown will be accomplished >-while an adequate water supply is available.

2.4.11.2 Low Water Resulting from Surges, Seiches, or Tsunami Because _the site is not located onLthe coast or on a1lakeshore, this section is.not applicable to BVPS-2.

2.4.11.3 Historical Low Water-f-

The lowest' flow of recordLJoccurred during the extreme drought of 1930. A minimum of 1,250 cfs flowed past Shippingport in August of that. year.

Since that.timep eight reservoirs. with ~1ow. flow augmentation capabilities have been -constructed.

The lowest. flow that would have occurred :in E1930 with.the contemporary reservoir'

. system is.4,000 cfs.

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October 1983 r

Amendment 3 12.4-15 t.

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Insert A A limiting condition for operation of 654 ft. ms1. is the minimum oper-ating level of the Auxiliary Intake Structure.

The backup service water pumps located in the auxiliary intake structure will not have su f fic ient water available at the pump suct ions at rive r levels below 654 f t, ms1.

Below elevation 654' mal. the se rvice water pumps loca ted in the main int ake structure will be the only sour ce availab le to supply se rvice water to the s t at io n.

These service water pumps are designed to supply water to the station at river levels down to 648.6 f t. mal. The minimum design water level of 648.6 f t. mst. is the river level which would occur only if the nearest downstream dam failed concurrent with a river flow of 4,000 cfs, equal to draught of record.

Howeve r, the Corps of Engineers has maintained that the nearest downs tream dam is considered safe agains t e art hquakes (U.S. Army Co rp s of Engineers 1968, 1969).

The refo re, the river level will not drop below the suction level of the service water pumps.

References:

U.S. Army Corp of Engineers 1968, letter from Wayne S. Nicho ls,

Colonel, Pittsburgh District Engineer, to R. McAllister, Duquesne Light Company, dated December 16, 1968.

U.S. Army Corps of Engineers 1969, letter from Wayne S. Nichols, Colonel, Pittsburgh District Engineer, to R. Kitchell, Stone and Webster Engineering Corporation, dated August' 26, 1969.

ATTACHMENT 3 Response to Outstanding Issue 4 of the Beaver Valley Power Station Unit No. 2 Draft Safety Evaluation Report Draft SER Section 2.4.11.2:

Emergency Water Supply (excerpt)

Because the intake. structure is located on the bank of the Ohio River, it is expected that suspe nded sediment in the river will accumulate in the lower part of the structure. The applicant states that silt ac cumula t ion in the int ake structure will be monito red semiannually and that silt exceeding a 15-inch allowable limit will be removed by a pumping opera-t io n.

The applicant has not provided the design basis for an allowable silt ac cumula t ion level of 15 inches nor as surances that a semiannual inspection interval is freque nt enough to preclude sediment from accumu-lating to a level that could af feet the capability of the service water pumps - in the int ake structure to obt ain an adequate amount of cooling water for safety-related purposes.

The staf f has submitted questions. to the ap plica nt and will de t ermine the adequacy of the silt monito ring program when responses are received from the applicant.

Response

The requested additional information on silt accumulation and removal is con'tained in the response to FSAR Question 240.09 (Amendment 6).

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ATTACHMENT 4 Response to Outstanding Issue 49 of the Beaver Valley Power Station Unit No. 2 Draf t Safety Evaluation Report Draft SER Sect ions 5.3.1 and 5.3.3:

Reactor Vessel Materials ard Reactor Vessel Integrity (excerpt)

The applicant has not reported the Charpy V-notch energy and mils lateral expansion da ta ve rs us tempe rature fo r - each reacto r vessel beltline material.

- Response:

The Charpy V-notch energy and mils lateral expansion data versus tempera-ture for each reac to r beltline material is pr ovided in the following tables.

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l BEAVER VALLEY UNIT 2 REACTOR VESSEL CORE BELTLINE REGION TOUGHNESS PROPERTIES Inter. and Lower Shell Longitudinal Weld Seams and Girth Seam Weld Code No. G1.42 Temperature Energy Lateral Expansion Shear

(*F)

(ft-lb)

(mils)

(%)

-80 7

1 0

-80 5

1 0

-80 7

2 0

-40 21 12 5

-40 17 8

0

-40 28 19 10 0

-28

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'.10 0

46 32 20 0

30' 22 10 30 93 62~

50 30 82' 49 40 30-110 74-60 100 146 87-100 100 117 78 80 100 138 84 90 160 145 85 100 160

-134 83' 100 160 127 81

.95 212 140 83 100 212

.148 82

-100.

212 146 83 100

.T

= -30*F NDT

-RTNOT = -30'FJ i

i BEAVER VALLEY UNIT 2 REACTOR VESSEL CORE BELTLINE REGION T0LIGlWESS_fR0PIIEIES Lower Shell Course P1 ate 89005-1 P1 ate B9005-2 Temp.

Energy Lat. Exp.

Shear Temp.

Energy Lat. Exp.

Shear

( F)

(ft-lb)

(mils)

(%)

(*F)

(ft-lb)

(mils)

JL

-40 10 6

0

-40 8

4 0

-40 11 7

0

-40 8

4 0

-40 15 10 0

-40 7

3 0

10 18 14 10 10 22 18

.l 10 10 28 21 15 10 24 20 10 10 21 17 10 10 28 23 15 40 29 23 15 40 34 26 20 40 38 28 20 40 33 27 20 40 30 26 15 40 37 29 20 74 54 42 40 74 40 33 30 74 41 32 30 74 48 39 35 74 49 37 35 74 52 41 40 s

100 65 51 60 100 60 49 60 100 76 57 70 100 59 47 60 100 57 45 60 100 55 45 60 160 85 65 100 160 77 61 100 160 82

.63 100 160 75 62 100 160 80 60 100 160 81 65 100 T

= -50 F T

= -40 F NDT NDT RT

= 28 F RT

= 33 F NDT NDT

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BEAVER VALLEY UNIT 2 REACTOR VESSEL CORE BELTLINE REGION TOUGHNESS PROPERTIES Intermediate Shell Course Plate B9004-l' Plate B9004-2

. Temp.

Energy.

-Lat. Exp.

Shear Temp.

Energy Lat. Exp.

Shear

']"F) 1]ft-lb);

(mils)

_f%)

("F) jft-lb)

(mils)

(%)

-40 5

5 0

-40 8

5 0

.-40

'6

'8 0

-40 10 7

0

-40 7

7 0

-40 10 6

0

.10 24 19 10 10 15 13 5

.10 20 18 10.

10 16 15 5

f 10 19 16 10 10 18 16 5

l 40 22 21 20 40 24 19 10 l

40 26 23 20 40 26 20 10 40-24' 20 20 40 23 17 10 100 43 37 30 75 39 29 25 100 51 42 30 75 38-30 25 100 45 40 30-75 40 32 25

'llo 48 39 25?

100 50 40 35 110' 50 40 30 100 51 40 35 110 48-38' 25 100 54 43 35 120 59 46 40 160 69 57 90 120 53 40 30 160 68 55 90 120 62 50 40 160 70 60 95 160 73 61 80 212 70 55 100

'160-70 58 80 212 78 59 100 160 71' 59 80 212 79 61 100 1212 82 64 100

'212

'85 67 100 1717 -

83 65 100 l

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ATTACHMENT 5 Response to Question 252.1 and Outstanding Issues 50 and 51 of the Beaver Valley Power Station Unit No. 2 Draf t Safety Evaluation Report Draft SER Sect ions 5.3.1 and 5.3.3 (OI 50):

Reactor Vessel Materials and Reactor Vessel Integrity (excerpt)

The applicant has not identified the azimuthal location and le ad fact o rs for each surveillance capsule.

Response

The azimuthal location and lead factors for each surveillance capsule is

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provided in the res pons e to Ques tion 252.1, Table 1,

"S urve illa nce Capsule Removal Schedule."

Draft SER Section 5.3.2 (OI 51):

Pressure Temperature Limits (excerpt)

The applicant has not supplied pres sure-temperature limit curves for the reacto r pr es sure ves sel, which comply - with the beltline and c los ure flange requirements of Appendix G, 10CFR50.

The applicant has not repo rt ed the amount of nickel for each beltline material.

- Response:

-The pres sure temperature limit curves for. hydros tatic pr es sur e and ' le ak tests, heatup, cooldown,.. and core operations is provided in the response to Question 252.1, Item C-1.

The amount of. nickel for : each.-beltline, material -is provided in the -

. res ponse to' - Que e t ion 252.1, Table 3, " Fracture Toughnes s Properties of 2the Reactor Vessel."

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Request for Additional Information Beaver Valley Unit 2 Materials Application Section Materials Engineering Branch 252.1 Appendices G and H, 10 CFR Part 50 were revised in the Federal Register on May 27, 1983 and became effective on July 25, 1983.

a.

Identify ferritic reactor coolant pressure boundary materials that do not comply with the fracture toughness requirements of Section 50.55a and Appendices G and H of 10 CFR Part 50.

b.

For materials which cannot meet the fracture toughness requirements of Section 50.55a and Appendices G and H of 10 CFR Part 50, provide alternative fracture toughness data and analyses to demonstrate their equivalence to the requirements of 10 CFR Part 50.

c..

To demonstrate conformance to Appendices G and H, 10 CFR Part 50:

(1) Provide pressure temperature limit curves for hydrostatic pressure and leak tests, heat-up, cooldown and core operations.

(2)

Identify the withdrawal schedule, lead -factor, test samples and materials in the Reactor Vessel Materials Surveillance Program.

(3)

Indicata the reference temperature, RTNDT' # #

materials in the reactor vessel closure flange region and the beltline regions.

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a r-'q) 2-(4)

Indicate the chemical composition (copper, nickel and phosphorus), unirradiated upper-shelf energy, and projected end-of-life RT and upper-shelf NDT energy for all beltline materiais.

RTNDTprojec-tions are to be estimated using the "Guthrie Formula" in Ccmmission Report SECY-82-465.

Upper-shelf energy projects are to be estimated using Regulatory Guide 1.99, Rev.1.

These projects are to be for the end-of-life neutron fluence at the 1/4T and ID r.eactor vessel locations.

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Resocnse 252.l(a)

All the Beaver Valley Unit 2 ferritic reactor coolant pressure boundary materials meet the July 26, 1983 effective revision of 10 CFR 50 Appendices G and H.

Specifically, the 10 CF' 50 ruling states that all reactor vessel beltline materials must have an initial Charpy upper shelf energy of 75 ft-lbs and must maintain upper shelf energy throughout the life of the vessel of no less than 50 ft-lbs. Since the ferritic material of the reactor vessel pressure boundaries lowest initial upper shelf energy is 75.5 ft-lbs, in the intermediate shell plate 89004-2, and the lowest end-of-life upper shelf energy of the beltline is. predicted to be 54 ft-lbs (Table 4 ), the fracture. toughness requirements of 10 CFR 50 are met.

252.l(b)

All ferritic reactor coolant pressure boundary materials meet the fracture toughness requirements of Section 50.55a and Appendices G and H of 10 CFR 50.

252.l(C-1)

Beaver Valley Unit 2 heatup'and cooldown curves are not

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impacted by the new 10 CFR 50 rule; which addresses the metal temperature of the. closure head flange.

Specifically, the-10 CFR 50 rule states the minimum metal temperature of the closure head flange should be RT 'DT +120'F for pressure above h

621 psig which-is 20 per_ cent of the preservice hydrotest pressure of 3105 psig. This minimum temperature for the closure head

-is'110*F since the RT is -10'F.

As a result, the closure head NDT flange region is less limiting than the-heatup and cooldown curves which are based on the beltline region. The original-heatup limitations (Figure 1) and cooldown limitations:(Figure 2) curves still apply for the Beaver Valley. Unit 2 reactor.

252.-1 ( C-2) -

The Beaver Valley-Unit-2 surveillance capsule removal schedule is listed in Table 1.

The' lead-factors-have changed due to an updated analysis'of 1ead factors'that-is moreLadvanced and'true to life.

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TAELE 1 SURVEILLANCE CAPSULE RE.YOVAL SCHEDL'LE Orientation Ex;ec:ec Identification of Lead Removal Capsule Capsules (g)

Factor (3)

Tine Fluence (n/cm-1 Ib U

343 3.12 1st 6.83 x 10 Refueling (C)

V 107" 3.12 6 EFPYs 3.16 x 10 9(d) 14 110 2.7 12 EFPYs 5.47 x 10 I9 X

287 3.12 18 EFPYs 9.48 x 10 Y

290 2.7 Standby Z

340 2.7 Standby (a). Reference Irradiation Capsule Assembly Drawing, Figure 2-5.

(From WCAP-9615.)

(b) The factor by which the capsule fluences leads the vessels maximum inner wall fluence.

(c) Approximate Fluence at 1/4 wall l thickness at' end-of-life.

(c) Approximate fluence at' vessel inner. wall at end-of-life.

(e) Not required by 10 CFR Part 50 Appendix H or ASTM E185-52 but recommended by Westinghouse.

h

The plate material used for surveillance was the intermeciate shell plate 390C4-2. The weld material was taken frca the joining of the surveillance plate and the adjoining icwer shell plate 390C5-2. All heat affected 20ne specimens were taken from the weld heat affected zone cf the intermediate shell plate B5004-2.

The test soecimens in each capsule are listed in Table 2.

TABLE 2 TYPE AND NUMBER OF SPECIMENS IN THE BEAVER VALLEY UNIT 2 SURVEILLANCE TEST CAPSULES Material Number of Soecimens of Indicated Tyos Charpy Tensile CT Bend Bar Plate B9004-2 Longitudinal 15 3

4 Transverse 15 3

4 1

Weld Metal 15 3

4 HAZ 15 252.l(C-3)

The reference temperature RTNDT, for materials in the reactor vessel closure flange region and beltline region are listed in Table 3.

252.l(C-4)

Chemical comoositions for all beltline material is listed in Table 3.

Unirradiated upper shelf energy (U.S.E.), projecced end-of-life USE's and projected end-of-life.

RT

. are N D,i indicated in Tabies 4 and 5.

RT projections were estimated NDT using the "Gutnrie Formula" from Commission Report SECY-82 465.

EOL uoper shelf energy predictions were estimated using Regulatory Guide 1.99, Revision 1 (Figure 3). Although crojections #cr bcth 1/a7 (Table 5) and ID reactor vessel locations (Table a) were calculated, the ID location is more limiting due to the higher fluence level.

t

_'-aW T;.BLE ~

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< N

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c

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N u-t e r Spec. No.

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r 'i o c u r a -o.

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u,.. '. e p r,c 9 w,

2 u. v a-A.m scmp e,

l v

c.

i

=

Qutlet Nozzle 39012-3 A508, Cl. 2 0.002 0.58

-10 ! -10 l'2 I

4 Nozzle Snell B9003-1 A5338, Cl.1 0.13 0.008 0.61

-10

-1C j

98 j

p Nozzle Shell

.89003-2 A5336, C1. 1 0.12 0.009 0.58 0

60 l 79.5 i

i Nozzle Shell B9003-3 A5338, C1. 1 0.13 0.008 0.51

-10 50 97.5

~"J '

nter. Shel' B9004-1 A5338, C1. 1 0.07 0.010 0.53 0

60 E3 Inter. Shell 89004-2 A5336, Cl. 1 0.07 0.007 0.59

-10 1 40 1 75.5 I

I i

Lower Shall B9005-1 A533B, Cl. 1 0.08 0.009 0.59

-50l 28 E2 I

l

=

l Lower S.

i

,9005-2 Ao330,.Cl. 1 0.-/

u.uC9 n,

nei.

c u

!-.e e.:

i Be++mm Head ~.^ us

.a. c. '. _1 A.:.3.u, r. i. i

.n. i a

n.. c r 7

..ac.

.n e

c

.w 3

botto Haad,ome ervur,

see o, l.,

t.-

u,. e n,

,a.::

-;c

-c:

. c.

ut, m

u j

i Weld (Inter. &

l I

Lower he,i.

onc 3_

i a

Scams & Girt-q a

Sean)*

0.08 0.008 : 0.09

-30

2.5 HAZ (Plate 59004-2

-e3

-20 75

'1 rwn i

i M

c j

( 2. i. A a c o tr e-. w !. 6 d u e.s c w -.n u.

S.,3..,..., 7. % ',

3 q

NOTE:

Sa-'e neat of wire and lot of flux used in al' seams inc'udin:

weldment.

. survei'ian:e A

-g

_-3s e

B

TABLE a PREDICTED E0L RT AND UPPER SHELF ENERGY ET OF BELTLINE MATERIAL FOR ID LOCATION Se9."

o'-

Fluen e Avg.

Guide ECL USE 1.99 USE Component Code No.

n/cm

"'NDT

a NDT

' NOT"'

ft-lb UUSE ft-lb s

1 I9 Inter. Shell'

.B9004-1 5.4x10 60 57 117 83

-28%

60 I9 B9004-2 5.4x10 40 59 99 75.5

-28';

54 10 i

Lcwer-Shell B9005-1 5.4x10 '

28 70 98 82

-28%

59 10 B9005-2 5.4x10 -

33 59 92 77.5

-28%

56 I9 Inter. +

5.4x10

-30 47 17 144.5

-36%

92

- Lower Shell

.Long Weld' l

Seam.

Inter.

.to' Lower-

- Shell Girth q

i

~

i Weld ***-

j.

t

. j --

aRTND was calculated using the Guthrie method..

n.

b 0(Cu) f 50 E {u x P. W R aRTNDT( )

(-

10 a Upper shelf enbrgy_ was determined from the Regulatory Guide 1.99 method

~

-(Attached Figure - 3.) -

- - - - -Same' heat of wire and' lot of fluxTused in -all. seams -' including the-

-surveillance' weld.

N r

i m..h

};l

~

p.

g M.

TABLE 5 PRED!CTE3 EOL 'lT -. AND UPPER SHELF ENERGY NUi 0F BELTLINE. MATERIAL FOR iT LOCATION op

' Reg.**

Avg.

Guide EOL Fluenge RT USE 1.99 USE

-m Component Code No.

n/cm D0'.

a"'NDT

"'NCT ft-lbiSaOSE ft-lb i Initial i

i i

19 4

i Inter. Shell B9004-1 3.2x10 60 49 109 S3

-25 52 i

I lo B9004-2 3.2x10 -

40 51 91 75.5

-25 57 l'

19 Lower Shell 39005-1:

3.2x10 28 60 88 82 52 19 39005-2~

3.2x10 33 51 84 77.5

-25 55 j

Inter.'&

-30 41 11 144.5

-32 9S-Lower Shell' Long Weld Seams.

Inter.

- to Lower

-Shell-Girth-Weld ***'

i'

[

i LaRT was calculated using:the Guthrie method.

NDT 3RTNDT(PF). = -(-10 + 470' (Cu-)- +1350 (% Cu :x - % Ni))

(

g)

a'JppershelfenergyLwasdeterminedfromtheRegulatoryGuidehl.99.:ethod.

~

- - " * = ' -

e

' ~ '

- '(Attached Figure;32)

u

- - -[Samefheat,of. wire and.I ot of llux~used in' al,1/seamsfincludingltne-

~

~ l

~

- surveillance weld.

a

(

1

f ^-

p

?

~

r E

m

MATERIAL PROPERTY BASIS C0fiTROLLIfiG MATERIAL PLATE METAL r

COPPER C0t1TEf1T C0f15ERVATIVELY ASSUMED TO BE 0.10 ',lTP.

tk I

h RT AFTER 10 EFPY 1/4T, 139*F t10T 3/4T, 114 F CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 10 EFPY AtlD C0tiTAltiS t%RGIf15 0F 10*F Afl0 60 PSIG FOR POSSIBLE IriSTRUMElli ERRORS w..

i I

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HYDR 02TATIC TEST TEMPERATURE-(279'F) l FOR THE SERVICE PERIOD UP TO'10 EFPY l

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INDIC ATED TEMPER ATU AE ( DEG.F 1 FIGURE 1 BEAVER VALLEY UllIT 2 REACTOR C00LAriT SYSTEM HEATUP LIMITATI0ftS APPLICABLE.UP TO 10 EFPY 6

MATERIAL P90PERTY BAS!5

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CC iTROLLItiG CTERIAL PLATE METAL CCPPER C0tiTEtiT C0!;5ER'/AT:'!ELY ASSUMED TO BE 0.10

',T PHOSPPORUS C0!!TE!1T

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ML20084E892

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