ML18220A509

Consists of Responses to Outstanding Items of NRC Report to ACRS Dated 11/4/1977

ML18220A509
Person / Time
Site:	Cook
Issue date:	11/18/1977
From:	Judkis M American Electric Power, Westinghouse Electric Corp
To:	Case E Office of Nuclear Reactor Regulation
References
NS-PLC-4546, S.O. AMP-460
Download: ML18220A509 (11)

Text

NRC FORM 195

(? 70I UAe NUCLEAR REGULATORY COM I

N NRC DISTRlBUTlON FoR PART 50 DOCKET MATBR1AL DOCKET NVMEER i6 FII,E NVMEER PSAR/FSAR Ai%)T DIST.

l TO:

i Mr. Edson G. Case FROM

'westinghouse Elect Corp.

Pittsburgh, Pa.

15230 M. H. Judkis OATE OF OOCUMENT OATS R 8C El V 8 O 11/21/77 MLETTER l

CIORIQINAL Wcopv QNOTORIZEO JihlNC LASS IF IEO PROP INPUT FORM NUMEER OF COPIES RECEIVEO

, OESCRIPTION I

i Consists of responses to Outstandi tems of NRC report to ACRS dtd 11/04/7 n the matter of the Donald C.

Cook Nuc Plant Unit g2...w/att supporting inform ENCI,OSV RE ear tlOn ~ ~ ~ ~ ~

~

3p '+ 5p PLANT iNAbK:

DONALD C COOK UNIT g 2 I

jcm 11/22/77 I

FOR ACTiON/INFORMATION I

ASSIGNED AD:

LTR BRANCH CHIEF:

PROJECT iKQfAGER S

L

@sic ic PDR LAINAS IPPOLITO INTERNALOISTRI BUTION ICE P.

COLLINS OUSTON HELTEiiES CASE (LTR)

NIPC LTR)

KNIGHT LTR BOSNAK SIHWEIL PAWLICKI ROSS LTR NOVAK ROSZTOCZY CHECK TEDESCO LTR BENAROYA LPDR:

TIC NSIC ACRS 16 CYS SENT CA F.

ROSA GAMMILL 2)

VOLLiKR (LTR)

BUNCH J.

COLLINS KREGER CcJCAQN EXTERNALOISTRISUTION CONTROL NUMBER 773260149

Westinghouse Electric Corporation AEW-7036 PhvR $)$ tglT5 oil'I<.Jl BOX 355 Pi1 L~a~lfgtl PKZiS'>IVM'3 1523Q November 18, 1977

'INI grpgg)P l g/jg~f flLP Ii~'ower'@Sems Company Nr.

Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue

Bethesda, Maryland 20014

Dear Nr. Case:

AMERICAN ELECTRIC POWER NS-PLC-4546 S.O.

At1P-460 q,

jfg/(p 0)

Oping

)g~~

P~ia~'orat

~"JI 8qrgq" Donald C.

Cook Huclear Plant Unit 2 Res onses to Outstandin Items of I'tRC Re ort to ACRS The NRC report to the Advisory Committee on Reactor Safeguards (dated November 4, 1977) in the matter of the Donald C.

Cook Nuclear Plant Unit 2 has been re-viewed with respect to the Outstanding Items listed in Section 1.8.

This letter'ddresses those outstanding items in the Westinghouse area of cognizance.

A separate letter from the American Electric Power Service Corporation will for-ward the authorization for this submittal on -Docket Number 50-316.

The responses to these outstanding items are given below in the same sequence that the concern appears in the report.

1.

The additional information requested with regard to the new thermal design basis THING-IV uncertainties has been provided in a Westinghouse letter, NS-CE-1583, October 25, 1977, from tlr. C. Eicheldinger to Hr. J.

F. Stolz, HRC.

2; The additional information requested with regard to the WRB-1 heat transfer correlation has been provided in a Westinghouse letter, ftS-CE-1581, October 24, 1977, from Nr. C. Eicheldinger to Hr. J.

F. Stolz, NRC.

3.

The additional information requested relative to the treatment of the radial pressure

gradient, THINC-IV code has been provided in a Westinghouse letter, NS-CE-1591, November 2, 1977, from Mr. C. Eicheldinger to Hr. J.

F. Stolz, HRC.

4.

The steam generator subcompartment pressure response analysis in response to guestion 22.3 will be submitted to you by January 3, 1978.

The preliminary analysis was presented to the NRC in a meeting on November 11,

1977, and are provided in a separate letter.

The preliminary results presented in the meeting show that the postulated accident does not endanger public health and safety.

773260149 I

Edson G.

Case November 18, 1977 The additional documentation requested with regard to the electrical equip-ment environmental qualification tests is provided.in a separate Westinghouse

letter, AEW-7035, dated November 18, 1977, addressed to Hr. Edson G.

Case.

The revision to the Westinghouse fuel rod performance

code, PAD 3.3, described in MCAP-8720, "Improved Analytical Methods Used in Westinghouse Fuel Rod Design Calculations" has no impact on loss of coolant accident (LOCA) analysis results performed for Donald C.

Cook Unit 2.

The maximum stored energy occurs early in life when the PAD 3.3 and the approved PAD 3.1 codes give essentially equal fuel temperatures.

In fact, equal peak average temperatures were assumed in the LOCA analyses.

Calculations were also made-for higher burnups where PAD 3.1 and 3.3 give different values of temperature and rod pressure.

However, the early in life conditions were found to be most limiting.

PAD version 3.1 gives higher values of fission gas release than PAD 3.3 during typical first and second cycles of operation.

It is only for the third cycle and beyond that PAD 3.3 results in higher fission gas release rates than PAD 3.1.

The lead rod will not be calculated to exceed system pressure during the first cycle of operation with either PAD version The modified design criterion for fuel rod internal pressure is:

"The internal pressure of the lead rod in the reactor will be limited to a value below that which could cause (1) the diametral gap to increase due to outward cladding creep during steady state operation; and (2) extensive DNB propagation to occur."

In addition, we have reviewed the fuel damage assumptions used in the site dose evaluation for this plant, and we find that with the revised design criterion the dose consequences of the accidents remain essentially unchanged and well below the consequences due to the LOCA accident as reported in the FSAR.

The conclusion for an analysis of a postulated loss-of-coolant accident with the accumulators isolated with system pressure at 1000 psia and system tempera-ture at 425'F has been presented in FSAR Amendment 78 in response to guestion 212.33.

Additionally, the peak clad temperature would be expected to occur at the core midplane, where maximum power is available.

The clad temperature at the midplane will turn around shortly after BOC (bottom of core recovery) as it does in the SAR ECCS analysis, an additional increase of ('4.7 x 6) 28'F is anticipated based on adiabatic heatup at the core midplane, giving a peak clad temperature of 1746'F

~

A calculation for postulated gas blanketing on the tube side of the steam gen-erators for breaks less than two inches has been made in addition to the re-sponse to guestion 212.34 (FSAR Amendment 78).

The maximum volume of hydrogen that could be available to accumulate in the steam generators is 12Ft3 at 1200 psia-and 567'F, which are the conditions at which the RCS would stabilize when the steam generator safety valves are actuated.

This volume of non-conden-sible gas distributed among the four steam generators results in approximately a 0.27 foot (or ~0.1 psi) reduction in the available elevation head in each steam generator.

For a long time period the core and hot leg side of the steam

Hr. Edson G.

Case November 18, 1977 generators are voided so that a significant driving head exists.

When the RCS becomes water solid -the density difference between the hot side of the RCS and the cold side of the RCS is sufficient to drive flow around the loop.

The effect of non-condensibles accumulating in the steam generator would be to simply delay the time at which the system becomes water solid.

Adequate core cooling will at all times be provided.

9..

The behavior of the primary system pressure after a postulated steamline break has been evaluated.

This evaluation is presented in the atta'ched Appendix B to the response to Question 212.34.

10.

Westinghouse is performing a feedwater line break analysis, which will be presented for NRC review in November 1977.

11.

Analysis of containment temperature and pressure long term response to a postulated steamline break will be presented, approximately five months following the NRC approval of the Westinghouse LOTIC-3 code.

This commit-ment was made in the response to Question 22.9 presented in Appendix Q of the FSAR.

If there are any questi,ons on the above, please contact this office.

Very truly yours, I. C. Ratsep/lk Attachment Forty (40) cc:

R.

W. Jurgensen, 1L, 5A R. S. Hunter, 1L R. F. Hering, 1L S.

H. Horowitz, 1L P.

W. Daley, 1L S. J. Hilioti, 1L J.

G. Feinstein, 1L

)

lh.

H. J dkis, jariaZer-'meric n Elec ric Power Project

APPENDIX B The following additi'onal information is provided as requested in regard to guestion 212.34 (Amendment 78}.

Regarding the concern that press urizer thi ck metal coul d cause flashing of water entering the pressurizer during its refill phase following a hypothetical steamline break the following conservative calculation was made.

Assumptions:

1.

The energy contained initially in the pressurizer wall area in con-tact with the refilling water was transferred to the water until the pressure would no longer permit flashing.

2.

Only the portion of the pressurizer wall in contact with the fluid was assumed to transfer its stored heat.

3.. Pressurizer heat loss through its insulation during the transient was not considered.

4.

The water temperature used to calculate flashing was the highest entering the transient during the refill phase.

5.

All heat was assumed to be transferred by the time of peak repressur-ization (600 seconds) following the break.

Physical Parameters:

Pressurizer shell inside diameter

= 84 in.

Pressurizer shell thickness

= 4 in.

Pees'surizer lower head inside radius

= 43 in.

Pressurizer lower head thickness

= 2.8 in.

Pressurizer metal density

= 489 ibm/ft.

Pressurizer metal heat coapacity

= 0.13 Btu/ibm'F Pressurizer fluid enthalpy

= 250 Btu/ibm (Peak duting refill)

Initial wall temperature

= 652.7'F (Tsat at 2250 PSIA)

The peak pressurizer water volume during the transient was 677.8 Ft.

With a pressurizer lower head volume of 88 Ft3 the water level would thus rise to 15.33 ft. of the cylindrical shell height.

The amount of metal mass in contact with the fluid would be shell shell

head head 2

3 3

H (R i s'de2 -

R outside2) hshell

+ 3 H (R outside i s d

= 138.5 Ft3 APPENDIX Q UNIT 2 212.34-B1

MCp = 138.5 Ft

• 489 ibm/Ft

• 0.13 Btu/ibm - 'F

= 8804 Btu/'F The energy gained by the water boiled to make steam must be given up by the pressurizer metal wall.

When the pressurizer wall temperature decreases to the saturation temperature at the system pressure boiling will cease.

hE pressurizer wall = -aE Boiled Steam MCp aTwall

'steam (hfluid - hsteam By assuming a system pressure a

aTw<11 can be calculated and a solution for aMsteam can be found.

By assum>ng an isentropic compression of steam

'in the pressurizer the new system pressure can be found by calculating the fraction of the steam space occupied by the original steam.

initial steam Final Mi.t.

1 St

+ ~ steam

• Vf; 1 before mass addition Once Vfinal is calculated Pfinal can be calculated by knowing the pressure and steam volume at the time of',the beginning of pressurizer refill.

At the beginning of refill the steam volume is just the pressurizer and surge line volume.

At the begjnning of refill the pressure is 605.8PSIA and the steam volume 1843.7 ft.

1.4 1.4 final final Vo

)'4 final Vfinai The new Pfinal must be compared to the saturation pressure assumed in computing hT for the pressurizer metal.

The solution is iterative and convergance is reached when Pfinal and the saturation pressure assumed in the aT calculation are identical.

The final iteration of the calcu-lation is given below.

Assume Pf.

1

= 1520PSIA T t

= 598.0'F p ( initial sati

= 8804 Btu/'F*(652.7'F - 598.0'F)

= 481579 Btu sat steam "water

~ 1169.0 Btu/ibm - 250 Btu/ibm 9l9. Btu/ibm APPENDIX Q UNIT 2 212.34-B2

I

• g 481579 Btu 524 0 ibm steam ah tu m

At Po. = 605.8PSIA, Vo = 1843.7Ft 3 initial steam ll~)

)~

(1843.71b

- 677.81b final

'mm~

~ 1520PSIA The calculated Pfinal equals the assumed Pfina] so the final pressure has converged.

The final pressure calculated is approxi ately 500PSI higher than the peak pressure of 1020PSIA given in Figure 212.34-1 of'he response to question 212.34 f'r the limiting case with respect to reactor vessel integrity.

An evaluation of the reactor vessel integ-rity indicates that vessel integrity will be assured for the 40 year design life of the plant assuming the increased pressure.

The calculation presented is very conservative.

No credit is taken for pressurizer metal heat loss to the containment through the pressur-izer insulation.

Neither is credit taken for 'the recondensation of steam bubbles fromed at the wall as they rise through the subcooled water in the pressurized.

The calculation also assumes that boiling heat transfer is the mechanism for all heat removal from the wall.

During later portions of'he heat transfer the wall superheat will not support boiling.

Therefore, the actual impact on peak pressure of the pressurizer wall metal is expected to be much less than the calculation presented.

Figure 212.34-21 presents the hot leg temperature in the intact loops of the reactor coolant system for a hypothetical large steam line break transient.

At 200 seconds (the beginning of pressurizer refill) the hot leg temperature is approximately 280'F.

This serves as the basis for the pressurizer fluid enthalpy of 250 Btu/ibm used in a conservative

~

calculation of the pressurizer thick metal heat effect on coolant sys-tem repressurization following 9 steam line break.

Figure 212.34-21 corresponds to the case with full forced convection reactor coolant flow, presented in figures 212.34 212.34-4 (Amendment 78).

A similar figure is not provided for the case with free convection reactor coolant flow.

The evaluation of reactor vessel integrity per-fomed for the response to question 212.34 indicated that a similar size initial flow depth of less-than approximately 1/3 wall thickness would assure reactor vessel integrity for the design life of the plant.

The additional pressure due to fluid flashing in the pressurizer is much greater for the case with forced convection in the reactor coolant APPENDIX Q UNIT 2 212.34-B3

system than for the case with free convection.

The reason for the larger pressure increase is that the pressurizer refill is much greater for the former case (677.8 ft3) than the latter (278,4 ft3).

The larger pressurizer refill allows more thick metal to come into contact with the fluid, thus causing more flashing.

An evaluation of the reactor vessel integrity was performed using the conservatively calculated reactor coolant system pressure increase assuming pressurizer fluid flashing.

The fracture analysis utilized linear elastic fracture mechanics methods in the evaluation.

The frac-ture mechanics analysis results in the determination of the minimum depth flows that might propagate during the large steam line break.

The results of the fracture mechanics analysis are that a flow having a depth less than 1/4 of the reactor vessel wall thickness will not propagate during the large steam line break at end of plant life fluence levels.

This flow depth is within the range of flow depths that would not be missed during manufacturing and in-service inspections of the reactor vessel.

Therefore, no flow propagation will occur during the large steam line break.

APPENDIX g UNIT 2 21 2.34-B4

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