ML19338C779

Forwards Addl Info on Cycle 4 Reload Re thermal-hydraulics, Transient & Accident Analyses, & Reactor Physics & Fuels,In Response to NRC 800806 Ltr

ML19338C779
Person / Time
Site:	Millstone
Issue date:	08/25/1980
From:	Counsil W NORTHEAST NUCLEAR ENERGY CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
A01155, A1155, TAC-11348, TAC-11561, TAC-12505, TAC-42846, NUDOCS 8009050195
Download: ML19338C779 (26)

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August 25, 1980 Docket No. 50-336 A01155 Director of Nuclear Reactor Regulation l Attn: Mr. M bert A. Clark, Chief l Ope ~ ing Reactor Branch #3 l U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l

References:

(1) W. G. Counsil letter to R. Reid, dated March 6, 1980.

(2 T. M. Novak letter to W. G. Coonsil, dated August 6,1980.

(3 W. G. Counsil letter to R. A. Clark, dated August 7, 1980.

(4 W. G. Counsil letter to R. A. Clark, dated August 7,1980.

j6) W. G. Counsli letter to R. Reid, dated March27, 5) 1979.

W. G. Counsil lette (7) W. G. Counsil letter to R. A. Clark, dated June 2, 1980. l Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Additional Information on Cycle 4 Reload In Reference (1), Northeast Nuclear Energy Company (NNECO) docketed the Basic Safety Report in support of Cycle 4 operation of Millstone Unit No. 2.

Reference (2) requested that NNEC0 provide the NRC Staff Nith additional information to complete the "eview of the thermal-hydraulics and transient and accident analyses sectioris of Reference (1). In addition, additional information was requested to complete the review of the reactor physics l and fuels sections in the Reload Safety Analysis and the small and large I break LOCA/ECCS perfonnance results. l NNEC0 provided the response to Enclosure 1 of Reference (2) in References l (3) and (4). l In response to Enclosure 2 of Reference (2), NNEC0 provides Attachment i

1. I l

We trust you find this information satsifactory to resolve all questions received to date regarding Cycle 4 operation at Millstone Unit No. 2.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY j

RVM s W.'G. Counsil~ '

Senior Vice President l Attachment 8000050195 2

Docket No. 50-336 Attachment 1 i

Millstone' Nuclear Power Station, Unit No. 2 i Additional Information on Cycle 4 Reload f

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1. Provide a list of physics tests to be performed during Cycle 4 testing, including the acceptance criteria for each test as well as the actions to be taken if the acceptance criteria are not met.

RESPONSE

In References (5) and (6), NNEC0 provided the Staff with a description of the start-up test program for Cycle 3. The Cycle 4 start-up test program will be identical to the program conducted for Cycle 3 with the exception of the power coefficient measurement.

Power coefficient measurement difficulties during the Cycle 3 start-up test program required that the test procedure be revised.

, The procedure is currently being revised for possible use during the Cycle 4 start-up test program however, the degree of readiness of the revised procedure will determine whether or not the power coefficient test is performed during Cycle 4 start-up testing.

' The Power Coefficient test is not mandatory and NNECO has performed the test.for informational purposes only.

Proposed changes to the Cycle 3 acceptance criteria for Cycle 4 are:

(a) The measured sum of all control banks should be equal to or greater than 90% of the predicted sum. The review criteria for individual CEA worth should be the greater of + 15% or 100 pcm from the predicted value.

(b) For power distributions, the acceptance criteria of + 10% on RPD should be changed to review criteria with the foTlowing values:

(1) i 10% (M-P) for RPD > 0.9.

(2) i 15% (M-P) for RPD < 0.9.

(c) The equivalent reactivity difference between measured and predicted boron concentrations should be less than i 1% ak/k.

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2. Previous cycles have used an augmentation factor to account for the power density spikes due to axial gaps caused by fuel densification.

These previous cycle augmentation factors were included in the determination of F How are densification spikes accounted for in Cycle 47 xY.

RESPONSE

Power peaking augn.entation factors shown in attached Figure 4.2-1 will be used for Cycle 4. They were included in the determination of Fnfor all accident analyses performed for Cycle 4. The Techni-cal 5pecification limits on local power density (Figure 2.2-2),

LOCA peak linea'e heat rate (Figure 3.2-1), and LOCA allowable power level (Figure 3.2-2) also account for the augmentation factors.

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3. A partial list of physics characteristics for Cycles 2 and 3 and preliminary Cycle 4 data was presented in the BSR. Provide a list of final Cycle 4 physics characteristics and . comparisons with previous cycle values including the maximum radial p';wer peaks j

expected to occur (Fr and F xy with uncertainties and biases).

RESPONSE

A comparison of Cycle 3 and final Cycle 4 physics characteristics is shown in Table 1. In Table 2 comparisons of Fr and F"Y with uncertainties and biases are given.

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TABLE 1

SUMMARY

OF CORE PHYSICS CHARACTERISTICS Cycle 3(l) sycle 4 Boron Concentration (ppm)

HZP-BOL, No Xe, Peak Sm, ARO s 1205 1339 HZP-BOL, No Xe, Peak Em,. Bank 7 In 1271 ,

HFP-BOL, No Xe, Peak Sm, ARO 1248 HFP, Eq. Xe at 150 MWD /MTU, ARO 830 1000 HZP-BOL, ARI, X<0.99 675 l Refueling C , ARI, K<0.90 [68'F] >2000 B,

Inverse Boron Worth (ppm /%ae)  :

i HZP-BOL 91 (2) 94

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9FP-BOL 93 98 HFP-E0L 82 82 ,

Control Rod Worths (-%so)

HZP-BOL, Bank 7 In  : 0.64 0.73 HZP'-BOL, ARI 8.18 HFP-150 MWD /MTU, Bank 7 In 0.66 0.75 HZP-EOL, ARI (w/HFP Eq. Xe) 9.08 c

Moderator Temperature Coefficient (pcm/*F)

HZP-BOL, AR0 5.4 (2) 3.8 HFP-BOL, ARO, Eq. Xe -2.0 -4.2 HFP-E0L, AR0 -18.0 -23.6 Doppler Coefficient (ptm/*F)

HZP-BOL, ARO -1.44 -1.80 HFP-BOL, ARO -1.13 -1.20 HFP-E0L, AR0 -1.22 -1.31 e

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SUMMARY

OF CORE PHYSICS CHARACTERISTICS Cycle 3 Cycle 4 Total Delayed heutron Fraction. T,ff HZP-BOL, ARO 0.00624 0.00584 HZP-EOL, ARO 0.00524 0.00508 ,

Neut on cene ation Time, f.* (usec)

HZP-BOL, ARO 27.2 18.1 HZP-EOL, AR0 31.8 19.7 Nuclear Enthalpy Rise Hot Channel Factor, F g HZP-BOL, ARD, No Xe . 1.64 HFP-BOL, AR0, No Xe 1.46 HFP-BOL, ARD, Eq. Xe 1.41 HFP-BOL, Bank 7 In, Eq. Xe 1.59 ,

HFP-EOL, ARO 1 35 HFP-EOL, Bank 7 In 1.53 l

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TABLE 2 Total Planar Radial Peaking Factors (Including Bias and Uncertainty)

Values of F r Cycle 3(I) Cycle 4 Unrodded Region 1.60 1.59 Bank 7 Inserted 1.81 1.74 Values of F xy ,

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4 REFERENCES

1. Letter, Counsil to Reid, Millstone Nuclear Power Station, Unit No. 2, Proposed License Amendment, Power Uprating, Docket No.

50336, February 12, 1979.

2. Letter, Counsil to Grier, Millstone Nuclear Power Station Unit No. 2, Startup Testing Report, Docket No. 50-336, September 7, 1979.

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4. Discuss the effects of using a different DNBR correlation for Cycle 4 transient analysis than was used for Cycle 3. .

Response

For the Cycle 3 analysis which tses the CE-1 correlation. DNB is not predicted to occur if a DNBR of 1.19 is met. For Cycle 4 analysis which uses the W-3 correlation, DNB is not predicted to occur if a DNBR of 1.30 is met. Since two different DNB correlations (both approved by the NRC) have been used, a direct comparison of the absolute DNBR values is not valid. The Cycle 4 analyses has shown that the effects (if any) of using a different DNB ,

l correlation for Cycle 4 than was used in Cycle 3 are negligible. That is, the conclusions drawn for the Cycle 4 analyses (e.g. DNB will not occur for Condition II transients) are the sane as that determined for Cycle 3.

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5. For the CEA ejection accident at both HFP and HZP, how many fuel rods go into DNB and what is the maximum RCS pressure attained?

RESPONSE

Since the CEA ejection transient is a very short power spike event, the fuel limits are best defined in terms of peak fuel enthalpy, rather than DNB ratio. This is consistent with the criterion set forth.in Regulatory Guide 1.77. The CEA ejection analysis results presented in the Basic Safety Report and in the subsequent Reload Safety Evaluation Report for Cycle 4 indicate that the fuel limit for the transient is not exceeded. In fact, these results are less limiting than the results reported for Cycles 2 and 3. Therefore, the number of rods in DNB would be expected to be less in Cycle 4 than in previous cycles.

The RCS pressure spike resulting from the rod ejection is of no concern unless hot spot energy depositions in excess of 400 cal /gm are calculated, above which a pressure pulse could be postulated.

These conclusions are the results of extensive TREAT and SPERT experiments. Therefore, the RCS pressure was not determined since the maximum hot spot heat deposition for this event was calculated to be less than or equal to 172 cal /gm as reported in Reference (1).

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6. Previous cycle (Cycle 3) parameters assumed in the CEA drop analysis are identical to those assumed for Cycle 4 except for the more negative moderator temperature coefficient in Cycle 4. The minimum DNBR attained in the pre-vious cycle analysis using the CE-1 correlation was 1.21. Since the max-imum negative moderator temperature coefficient results in the mintiiium transient DNBR, why is the minimum DNBR obtained in the Cycle 4 analysis higher than that obtained in the Cycle 3 analysis? Also, since the EOC moderator temperature coefficient is much more negative than the BOC coefficient, why is it not used in the CEA drop analysis?

Response: .

All DNB ratios reported by Westinghouse are based upon the W-3 correlation, and are not directly comparable to any Cycle 3 DNS ratios, which are based upon the CE-1 correlation. Further discussion on this is given in the response to question 4.

The minimum DNB ratio attained during the CEA drop accident is not vary sensitive to the moderator temperature coefficient. A more negative moderator temperature coefficient would tend to return the core to full power with a smaller reduction in core inlet temperature. Since only manual rod control is available at Millstone,' there would be no power overshoot due to automatic rod motion in response to the dropping of a CEA. The DNB ratio would not fall below its value at initial operating conditions with the dropped rod.

Since the E0C moderator temperature coefficient is more negative than the BOC value, the EOC moderator temperature coefficient was assumed for the CEA drop analysis in the BSR.

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7. The PALADON computer code has not been approved by the staff for three-dimensional calculations. Provide a description of the types of calculations performed by PALAD0N for the Cycle 4 analysis.

RE,SPONSE PALAD0N two-dimensional calculations were used for the following Cycle 4 analyses:

(a) Cold shutdown and refueling boron concentrations.

(b) Dropped rod power distribution.

These types of applications of PALAD0N have been approved by the staff per the " Safety Evaluation of WCAP-9485", J. F. Stolz to T. M. Anderson, da ted September 12, 1979.

8. Please submit values for the following variables that were not provided in the Millstone 2 small-break LOCA ECCS performance results.

a. Hot rod (1) differential pressure at time of rupture (2) temperature at time of rupture (3) axial distribution of circumferential strain

b. Hot assembly (1) time of blockage (2) differential pressure at time of blockage (3) temperature at time of blockage (4) axial distribution of reduction-in-flow area

RESPONSE

In discussion with the NRC Staff at a meeting at your offices in Bethesda

' on March 18, 1980 and as documented in Reference (7), it was NNEC0's understanding that a detailed review of the Westinghouse small break LOCA model for Millstone Unit No. 2 would not be made prior to model changes required by the Staff as a result of the TMI-2 accident. If the above understanding remains correct, the relevance of this question to the Cycle 4 reload is unclear. The Cycle 3 small break LOCA results are expected to serve as the basis for the Staff's evaluation. If this understanding is incorrect, NNEC0 respectfully requests clarification in this regard.

Nonetheless, responses are provided as follows:

(a) 1. Differential pressure at time of rupture: 457 psi 0

2. Temperature at time of rupture: 1633 F

3. Axial distribution of circumferential strain:

AD Location

• Strain at Burst, ( 11) 6.5 and below <10-0 7.63 2 X 10-4 8.544 0.021 8.886 0.055 9.4 and above 0.1

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(b) 1. Time of rupture: 1035 seconds

2. Differential pressure at time of rupture: 501 psi U

3. Temperature at time of rupture: 1614 F

4. Axial distribution of reduction - in-ficw-area: As described in section 2.0 of Reference 1, the small break analysis is performed with a model which conservatively addressed flow in hat rod heat-up calculations by using the steam flow rate asso-ciated with an unblocked average rod. If a consideration of blockage effects were combined with use of the steam flows that encompass the hot rod, the increase in steam flow rate would result in a PCT reduction from the Millstone 2 Cycle 4 related small break ECCS analyses.

• Distance (feet) above bottom core Ref (1): Addendum to WCAP-9528, Oct.1979

9. Please submit values for the following variables that were not provided in the MP2 large break LOCA ECCS performance risults.

a. Hot rod (1) differential pressure at time of rupture ,

(2) temperature at time of rupture (3) axial aistribution of circumferential strain _

(4) time of peak cladding temperature

b. Hot assembly (1) time of blockage (2) differential pressure at time of blockage ,

(3) temperature at time of blockage (4), axial distribution of reduction-in-flow area

Response

Infonnation below is provided for the limiting break ECCS analysis submitted in support of the cycle 4 reload for Millstone 2 (i.e. CD = 0.6 DECLG break);

(a) 1. Differential Pressure at time of rupture: 731 psi

2. Temperature at time of rupture: 16480F l

3. Axial distribution of circumferential strain:

Location

• Strain at Burst, (h) 2.848 and below <3 X 10-3 j 4.0 0.0473 j 4.5 0.1 Between 4.5 and 7.0 0.1 l 7.5 0.0483 8.0 0.0195 1 8.544 and above <5 X 10

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4. Time of peak cladding temperature: 162.6 seconds l l

(b) Burst is not predicted for the hot assembly rod in the Cycle 4 reload large break ECCS analysis for Millstone Unit 2.

• Distance (feet) above bottom of core

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10. The NRC staff has been generically evaluating three materials models that are used in ECCS evaluation models. Those models are claddiqg rupture temperature, cladding burst strain, and fuel assembly find blockage. Subsequent to Westinghouse subnittals and your applica-tion of WCAP-9528, "ECCS Evaluation Model for Westinghouse Fuel Reloads of Combustion Engineering NSSS," and its addendum, we have (a) met and discussed our review with Westinghouse and other industry representatives (b) publisised NUREG-0630, " Cladding Swelling and Rupture Models for LOCA Analysis, and (c) required fuel vendors and licensees to confirm that the plants would continue to be in conformance with the ECCS criteria of 10 CFR 50.46 if the materials models of NUREG-0630 were ubstituted for those models of their ECCS evaluation models and certain other compensatory model changes were allowed.

The Westinghouse materials that are described in WCAP-9528 are virtually the same as those used in prior Westinghouse ECCS evaluation models, and they were evaluated in NUREG-0630. Small differences are attributable to modifications that were made to reflect the geometrical differences in fuel designs for the Millstone 2 plant. Therefore, until we have completed our materials moc'el review, we will require plant analyses performed with the ECCS evaluation model as described in WCAP-9528 to be accompanied by supplemental analyses to be performed with the materials models of NUREG-0630. Therefore we request that NNEC0 submit a sample calculation as described above.

RESPONSE

The possible penalties for fuel rod models proposed by the NRC Staff in NUREG-0630 has been considered and the following information is provided.

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A. Evaluation of the potential impact of using fuci rod models pre-sented in draf t iiUREG-CG30 on the Loss of G..Cooiant h // Accident (LCCA) ana1ysis f or /1nt :rotje- .'d st. t. M D's .

This e/aluation is based on the limiting break LOCA analysis identi-fied as follows: .

BREAK TYPE - DOUBLE Ei;DED C01.0 LEG GUILLOTIllE BREAKDISCl!ARGECOEFFICIEilT[D N . b . 4 6 "

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ilESTIllGliOUSE ECCS EVALUATIOlt f(CDEL VERSI0tl AIN- _

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• CLAlf STRAI;l CURIllG BLOWDOW AT THIS ELEVATI0tl h./3Percent Percent -

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• I4ACI!4U'4' Cl).D STRAI t AT THIS ELEVATIO!! - 4.65

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ms.Aurn' 11aximum temperature for this x nede cccurs when the core reflood rate is (GREATER) than 1.0. inch per second and reflood heat transfer i

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_- - The maximum potential impact on the ruptured clad node is expressed in letter !!S-TMA-2174 in terms of the change in the

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peaking factor limit (FQ) required to maintain a peak clad tem- l perature (PCT) of ?.2000F and.in terms of a change in PCT at a l constant FQ. Since the clad-water reaction rate increases sig-l nificantly at tcmperatures 2cve 2200.CF, individual effects l; (such as APCT 'due to changes in several fuel rod m.odels) '

indicated here may not accurately apply ovcr. large ranges,

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but a simultaneous cnange n.

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in the neighborhood of 2200.0F justifies use of this evalua-

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.From US-mA-2174: .

For the Burst liode of the clad: .

- 0.01 AFQ + < 150 F BURST fiODE APCT

- Use of the NRC burst model and tho ravised pastirchouse f'

. burst model could require an F0 reduction of 0.027 The maximun estimated impact of using the fiRC strain l

model is a required FQ reduction of 0.03.

..s l '

Therefore, the maximum penalty for the llot Rod burst node is: -

i ~

APCT1 = (0.027 + .03)' (1500F/.01) = 8550F liargin to the* 22000F limit is:

b 0F

~

f APCT2 = 2200.0F - PCTB= ,

~ '

• The FQ reduction required to maintain 'the 22000F clad l tempera- ,

ture limit is: .

~

. AFQ 3

= (APCTy^- 6PC.T 2 ) I'01 o AF0) -

150 F . .

.E('c69'-4Q)(h)

- =.D8bY (but not.less than zero).

. . ~ 24 fl0ft-BURST tiODE_ .

The maximum tempera.ture calculated for d~non-burst section of h- clad typically occurs at an elevation above the core mid-plane

.- during the core reflood phase of the LOCA transient. The potan-

- tial impact on that maximum clad temperature of using the fiRC

> " , fuel rod models can be estimated.by examining two aspects of the

, analyses. The first aspect is the change in pellet-clad gap conductance resulting frca a difference in clad strain Note at that the

• T non-burst maximum clad temperature node elevaticn.

"' clad strain all along the fuel rod stops af ter clad burst occurs and use of a different clad burst model can change the time at

.. Three sets of LOCA analysis results

'i which burst is calculated.

were studied to establish an acceptable sensitivity to. apply generically in this evaluation. The possible PCT increase I

resulting from a change in strain (in the Hot Rod) is +20.0F

[

• per percent decrease in strain at the maximum clad temperature 1 .

2

local suii>. ,..u. e..... .

coolant systen blowdown phase of the acciuentthe use o strain

' blowdown" indicated above.

that must be considered here is

.'; Therefore: -

[ RAlli)

={0%.01 strain ) (MAX STRAllt - BLO' D0ilii ST APCT -

1 3 .

I'

=(j)(d.13#). - . /3_) xifA ,

pa Y -

' The second aspect ofSince the analysis the greatest thatvalue can increase PCT is of blockage flow blockage calculated.

indicated by the !!?.C blockage model is 75 percent,'the maximum PCT increase cSn be estimated by assuming that the current leve of blockage in the analysis (indicated above) is raisedl to 75

. percent and'then applying an appropriate sensitiyity formu a

- shown in itS-TMA-2174. ..

5 Therefore, '

APCT4 = 1.250F + 2.360F (50 - PERCEtiT CURREtiT SLOCKA (75-50)

= 1.25 (50 - ___) + 2.'36 (75-50) op . ,

o ' .

flood rate 'is greater than 1.0 .

. /If PCTr4 occurs when the core reThe total potential PCT increase -

V incli per second APCT4 = 0. .

for the non-burst node is then

, APCTS = APCT3 bCT4 eY. .

~

'liargin to the 22000F limit is APCT 6=22000F-PCTgg*fi* ,

- 1 -

The.FQ reduction required to maintain this 22000F clad tem-4 pcrature limit is (frca ilS-TMA-2174) - -

=.. .

q4 - g4 .

) *

' i -

AFQ; f = (6 PCT S - 4.CT ) 10 6 b *F'hPCT '

4FQ;; =

~~ OW but not less than zero.

. e 3

. e

- - - , _ , , vw w

? * ,

The peaking factor reduction required to maintain the 2200 andFAFQg ,

clad temperature limit is therefore the greater of oFQB or; = 0.0269 a FQPEt!ALTY D. The effect on ECCS analysis results of using improved, more representative data has been assessed in relation to the ECCS analysis performed and submitted for the cycle 4 reload of the f4illstone 2 plant. It has been  ;'

determined that the margin involved in the conservatism of input parameters is more than adequate to offset potential burst-blockage model impacts.

l Specifically, design value fuel pellet temperatures were assumed for the Millstone 2 ECCS analysis involving Hestinghouse fuel. Fuel parameters specific for cycle 4 confirm the existence of additional margin (330F) compared to the values utilized in the analysis.

Previous licensing credits applied to the W evaluation model analysis

~

have resulted in a mininum FQ increment of 0.07 for each 85oF reduction in pellet temperature. Therefore, incorporating the cycle-4 specific fuel information would result in a cycle 4 margin of 0.0271 in Fg for the 33of margin in the pellet temperature parameter for the cycle 4 tii11 stone 2 fuel. Hence, consideration of pellet temperature-related input confirms that adequate margin exists in the ECCS analysis submittal to preclude any Fg or peak kw/ft adjustments associated with burst-blockage considerations.

C. The peaking factor limit adjustment required to justify plant operation for this burst-blockage issue is determined as the appropriate aFQ credit identified in section (B) above, minus the AFQ calculated in section (A) above (but not greater than zero):PEllALTY F ADJUSTMEllT = 0.0271 - 0.0269 E0 ,

q O

e

This evaluation demonstrates that a conservative assessment of those penalties is compensated for by available improvements in the ECCS analysic already provided to the Staff. The procedure utilized to perform the analysis is deemed appropriate and suitably conservative and provided adequate supplementary material until final resolution of the overall fuel rod model concern is achieved.

The format is similar to evaluations already provided to the NRC to support licensing of Westinghouse-NSSS operating plants. Credits specific to the Cycle 4 Millstone 2 reload have been developed.

2

.